[en] Zirconium alloy pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs) undergo irradiation enhanced creep, growth and hydriding. The calandria tubes also undergo axial creep and growth under the influence of neutron irradiation. The irradiation induced creep-growth manifests. as axial elongation, diametrical expansion and sagging of the pressure tube. Another degradation mechanism, hydriding, can lead to failure of pressure tubes through hydride blister formation at a cold spot, hydrogen embrittlement and delayed hydride cracking (DHC). Considering these degradation mechanisms a set of analytical methodologies and associated computer codes along with various inspection and diagnostic technologies have been developed to facilitate safe management of ageing of coolant channels of Indian PHWRs. This paper dwells briefly upon the aspect of development and application of the above mentioned degradation models and, tools and technologies for diagnosis, inspection and life extension of pertaining to ageing management of coolant channels of Indian PHWRs. (author)

[en] This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

[en] Zirconium alloy coolant channels are central to the design of Indian Pressurised Heavy Water Reactors (PHWRs) and form the individual pressure boundaries. These coolant channels consist of horizontal pressure tubes made of zirconium alloys, which are separated from cold calandria tubes using garter spring spacers. High temperature heavy water coolant flows through the pressure tube which supports the fuel bundles. A typical coolant channel in a PHWR is shown. These pressure tubes are subjected to several life limiting degradation mechanisms like creep and growth, hydrogen pick-up, reduction in fracture toughness and delayed hydride cracking phenomena because of their operation under high temperature, high stress and high fast neutron flux environment. Considering the early onset of these degradation mechanisms in Zircaloy-2 pressure tubes used in the early generation of Indian PHWRs, the life management of these coolant channels becomes a challenging task, involving multidisciplinary R and D efforts in areas like analytical modelling of degradation mechanisms, evolution of methodologies for assessment of fitness for service and, tools and techniques for remote on line monitoring of integrity, maintenance and replacement. The degradation mechanisms have been modelled and incorporated into specially developed computer codes, such as SCAPCA for irradiation induced creep and growth deformation modelling, HYCON for hydrogen pick-up modelling, BLIST for hydrogen diffusion, blister nucleation and growth modelling and CEAL for assessment of leak before break behaviour. These codes have been validated with respect to the results of in-service inspection and post irradiation examination. Development of analytical models actually paved the way for the evolution of more refined methodologies for assessing the safe residual life of coolant channel. Information gathered from various experiments simulating the degradation mechanisms, results of post-irradiation examination of the coolant channels and various research publications in international journals formed the bases for evolution of the above safety methodologies. Today, the analytical models together with the safety evaluation methodologies have become an important tool for assessing fitness for service of individual pressure tube in different reactors. Apart from developing the above analytical tools and methodologies for safety evaluation, several systems and tools have been designed and developed to cater for the activities like remote inspection, online integrity monitoring and life extension etc., of our life management programme of coolant channel. Amongst the systems developed for remote inspection of the coolant channel such as Dry Visual Inspection System (DRYVIS), BARC Inspection System(BARCIS) and Hydraulic Remote Inside Diameter Measurement System (HYRIM) etc., BARCIS is more versatile in terms of its capability to carry out different type of measurements. Inspection of the coolant channel by BARCIS has been incorporated as an important activity in our life management programme of the coolant channel. A non destructive diagnostic technique based on the principle of vibration measurement has been developed to identify the contact between the pressure tube and calandria tube. This technique called 'Non Intrusive Vibration Diagnostic Technique (NIVDT)' is being used as a tool to screen the vulnerable pressure tubes so as reduce the reduce the inspection load. Tools and techniques developed for online integrity monitoring includes Sliver Sample Scraping Tool (SSST). At present, scraping operation can be carried out in both the dry and the wet channel of operating reactor. The development of SSST is intended for taking nondestructively sliver samples from the operating coolant channel. These sliver samples are then analysed to get hydrogen content in them. Measured hydrogen content helps not only in predicting the operating life of a coolant channel but also in improving the predicting ability of the analytical model. Extension of service life of a coolant channel is the end objective of the programme for the life management of coolant channel of Indian PHWRs. In the fresh reactor of earlier design, loose fit garter springs shift from their design locations during hot commissioning. Since the shift of garter spring from the design location is undesirable from the bending creep point of view, they need to be relocated to the extent possible. Mechanical Flexing Tool (MFT) has been developed to carry out the above task. To accomplish the similar task in an operating channel, Integrated Garter Spring Repositioning System (INGRES) has been developed. With all these analytical models and methodologies on one hand and tools and techniques for inspection, online integrity monitoring and life extension on the other, the programme for life management of the coolant channels of Indian PHWRs has been successfully implemented This paper gives the highlights of the above mentioned different R and D activities carried out for life management of coolant channels of PHWRs which have progressed hand-in-hand with the operation and inspection experience

[en] India is the only country in the world, which has the experience of operating large number (seven) of pressurised heavy water reactors (PHWRs) with zircaloy-2 pressure tubes for the maximum number of hot operating years (HOYs). With the re-tubing (expected) of KAPS unit 1 with Zr-2.5%Nb pressure tubes in the later half of the year 2008, the era of zircaloy-2 pressure tube started nearly three decades ago will come to an end. Nearly twelve years of experience of operating PHWRs with zircaloy-2 pressure tubes in safe and integrated manner spread over 3 decades, has put India in a position to share its vast experiences with regard to this material behaviour in the radiation field. One such experience is the material characteristics of picking up hydrogen during the course of reactor operation. It is well known that the progressive structural degradation of zircaloy-2 pressure tube associated with the accelerated rate of hydrogen pick-up in the reactor environment is the sole reason its replacement much earlier than the stipulated design life. Assurance of safety of these pressure tubes during the period they remain in service has been a challenging task. This has been accomplished by monitoring the hydrogen content in a set of representative pressure tubes on regular basis. This paper summarises the Indian observation on in-reactor hydrogen pick-up behaviour of zircaloy-2 pressure tubes. (author)

[en] Advanced Heavy Water Reactor (AHWR) is a pressure tube type of reactor. Heat generated in the fuel assembly is removed by boiling light water natural circulation loop. Design criteria of coolant channels is given

[en] Material samples can be obtained from a tubular component by a sliver sample scraping technique without affecting integrity and residual service life of an operating system. These samples may be used for chemical analysis and determination of material properties, dimensional deviations, corrosion and microstructure. This information could provide an indication of remaining useful life of a component. This paper highlights the developmental goals, qualification of sliver sample scraping technique, typical profiles of sliver samples, determination of operating parameters, and evaluation of scraped regions and sliver samples. (author)

[en] The sludge accumulation in secondary side of mushroom type steam generators of Indian Pressurised Heavy Water Reactors (PHWRs) may lead to loss of thermal efficiency and corrosion. Sludge removal is required to minimise such effects for safe and enhanced operating life of the steam generators. A sludge lancing system has been developed for sludge removal from the secondary side of the steam generators. Jet Manipulator is one of the various modules of the sludge lancing system. The JM consists of three modules namely walker, elevator and nozzle heads. Each module is designed to pass through hand hole, having 180 mm diameter and 100 mm wide gap between steam generator shell and shroud. These three modules are connected to each other by quick connecting type joints and are having their specific functions. The walker crawls by step of single pitch of the tube along the central no-tube lane of the steam generator by taking lateral supports on the nearest tubes. The elevator is capable of lifting the nozzle head to a suitable height required for lancing operation of entire tube sheet of the steam generator. The nozzle head directs the multiple jets along the narrow inter tube lanes having 3 mm width, on both sides of the central no-tube lane. The nozzle can be set to move at different elevations such that the multiple jets will graze along the narrow tube lane to create the sludge lancing action. The provision exists for movement of JM in both directions, i.e. forward and reverse. This paper highlights the objective, design and development, selection of nozzles, qualification and performance evaluation of JM. The manipulator is remotely operable by compressed air in the forward and reverse direction in the central no-tube lane to position the nozzle head in the horizontal direction. (author)

[en] Pressure tube in Indian pressurized heavy water reactor is known to have sag due to dislocation of loose spacer wires (called garter spring spacers), irradiation creep and growth during operating life of the nuclear power plant. When the sag of the pressure tube is large enough, the pressure tube comes in contact with calandria tube, resulting in formation of hydride blisters over a period of time, which may eventually lead to failure of pressure tube. A system has been developed for precise detection of Garter spring spacers (SPGs), which play a vital role in relocation of Garter spring for extending the creep limited life of pressure tube. The system consists of eddy current based sensor, hardware conditioning module, PC-field Interface Unit, software conditioning module and ISA bus based add on cards to acquire the Garter spring data in conditioned form. The system also consists of precise displacement optical sensor and a quadrature pulse counter card for accurate position of the sensor head. The In-house developed automation software drive the tool with the help of motor driven linear mechanism for online recording of sensor signal along with the position information data along the length of pressure tube. All the input output signal between field and PC are optically and magnetically isolated to ohmic noise. The recorded data is passed through surge filters, digital filters, smoothing filter, interpolator, calibrators, slope finder to indicate the exact position of Garter spring spacers in automatic and semiautomatic mode. The analysis mode also takes care of data saving on storage disk, data security and audio-visual error messages while operating the system. All software and hardware modules are indigenously developed at RED/BARC and this system has been successfully used in numbers of Indian PHWRs. The same system with a minor modification in software and hardware can be used in inspection of tubular components in other industries also. (author)

[en] Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type reactor. A coolant channel consists of pressure tube, made of Zr-2.5 % Nb, which is separated from cold calandria tube using garter spring spacers. The principal function of pressure tube is to support and locate the fuel assembly and allows light water coolant through fuel assembly by natural circulation. Since AHWR is designed for life of 100 years, it necessitates the replacement of pressure tubes during service life. Easy replaceability of pressure tube, along with surveillance requirements, has major bearing on the design of coolant channel assembly. The several systems and tools have been conceptualised to cater the needs for easy and quick replacement of a pressure tube during reactor shut down. This paper gives the highlights of the innovative design features of coolant channel, preliminary design and pre-requisites for replacement, and experimental programme for demonstration of easy replaceability. (author)

[en] Pressure tubes made of zirconium alloys are a major core component of PHWRs, housing the fuel bundles and operating at high pressure-high temperature conditions. A number of life limiting issues are associated with pressure tubes. Estimates of blister growth rate have been made for a contacting Zr 2.5 wt% Nb pressure tube as a function of different parameters. In the absence of contact, the major reason that limits the life of an individual pressure tube is the presence of flaw. Based on the condition for Delayed Hydride Cracking (DHC), the allowable depth of flaw has been estimated as a function of stress and root radius. It is observed from the analysis that as the applied stress increases, the root radius beyond which it can be treated as blunt also increases. The results can be used directly for assessing the fitness for service of a pressure tube without categorizing the flaw as either sharp or blunt. In case hydrogen concentration is less than the solubility limit, estimates of allowable number of thermal cycles have been made as a function of depth of flaw to avoid gross fracture initiation as per ASME Section XI. In the absence of any flaw, the life of the pressure tube is governed by the period up to which leak before break can be assured. This assessment is based on a postulated crack and depends on the variation of fracture toughness and DHC velocity with temperature and irradiation. Based on the available information on material properties, an assessment of end of life of pressure tube has been carried out. Since, presently, the life management activities are concentrated on zircaloy-2, the results from the present study can be used to evolve guidelines for assessment of fitness for service of Zr 2.5 wt% Nb pressure tubes. (author)