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Abstract: A plain carbon steel feeder pipeline in the secondary circuit failed downstream of a flow measurement device (orifice meter) during operation at nuclear power plant. A detailed failure analysis done on the failed pipeline is described in this paper. The results established the fine surface pattern of ‘Horseshoe pits’ at the affected regions. X-ray diffraction analysis on the samples far from the failed regions showed presence of magnetite but on the sample from the failed region showed peaks due to base metal only, indicating dissolution of the oxide. Thickness profiling of the pipeline indicated reduction of thickness from the design 7.62mm to a minimum of 0.4–1.4mm at the location of the failure. These observations are characteristic of single phase flow accelerated corrosion. This paper details the extent of flow accelerated corrosion in various Indian power plants and the remedial measures for replacement and possible design and water chemistry changes to combat it. [Copyright &y& Elsevier]

This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils. Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V–4Cr–4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu–1 wt%Cr–0.1 wt%Zr – an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall – was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER). India has proposed to install and test a lead–lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li 2 TiO 3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb–Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box are briefly described. India has also initiated the development of fusion grade superconductors. Success achieved in the fabrication of Nb 3 Sn based multi-filamentary wires using the internal tin process and cable-in-conduit-conductors is also briefly presented. [ABSTRACT FROM AUTHOR]

Abstract: Hyperstoichiometric Plutonium rich (70%) mixed Uranium Plutonium carbide fuel was proposed as the driver fuel for Indian Fast Breeder Test Reactor (FBTR) after some deliberations. The issues of mixed carbide fuel and of plutonium rich (70%) carbide fuel, in particular, are: presence of ‘O’ impurity, clad carburization, lower solidus temperature, lower thermal conductivity and pyrophorocity. The paper highlights the extensive research and development carried out to address the above issues. The modification incorporated in the different process steps of fabrication flow sheet is also presented. The performance of the fuel has been assessed by periodic Post Irradiation Examination (PIE) of the fuel sub-assemblies for extension of life. Till date the fuel has seen a burn up of 155Gwd/t without any pin failure. [Copyright &y& Elsevier]

Abstract: Plutonium is poised to play a key role in the Indian atomic energy programme, as fast breeder reactors with Pu fuels will be a major source of nuclear energy production in India in the coming decades. Starting from separation of Pu from irradiated PHWR fuel, to fabrication of Pu containing fuels and their utilization in fast reactors, India has mastered various aspects of Pu based fuel cycles. This paper describes the development of technologies for the fuel cycles of oxide, carbide and alloy fuels containing plutonium being pursued in India, and highlights the broad based programme of basic research on Pu science that has provided the confidence for effective utilization of plutonium. [Copyright &y& Elsevier]

Abstract: Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90–95. Such tubes are conventionally produced by the ‘welded route’, which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the ‘seamless route’ which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties. [Copyright &y& Elsevier]

Abstract: The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16000MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO2 fuel bundles from Indian PHWRs, which included two high burnup (∼15000MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ∼11000MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29000MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO2, enriched UO2, MOX, and ThO2 fuels. [Copyright &y& Elsevier]

Abstract: A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10MPa and 573K), and fast neutron flux (typically 3×1017 n/m2 s, E >1MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr–2.5wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors. [Copyright &y& Elsevier]

Abstract: India’s nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th–Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle. [Copyright &y& Elsevier]

Abstract: Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30years in China, Europe, India, Japan, Russia and the USA for application in ITER test blanket modules (TBMs) and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical properties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation to different TBM and DEMO options. [Copyright &y& Elsevier]