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Prudhvi Raju P.V.S.N.,Nuclear Fuel Complex | Mandal D.,Bhabha Atomic Research Center
Journal of Nuclear Materials | Year: 2015

The mean particle size and size distribution of Ammonium Di-Uranate (ADU) particles, precipitated during the precipitation reaction of Uranyl Nitrate Pure Solution (UNPS) with ammonia play an important role on the sintered density of UO2 pellets. The quality of precipitated ADU depends on number of process parameters viz., pH of UNPS, concentration of uranium in UNPS, flow rate of ammonium hydroxide, temperature etc. However, the effects of the presence of free acid and entrained Tri-Butyl-Phosphate (TBP) in UNPS on the quality of ADU powder were not studied till date. Experiments were conducted to study the effect of free acidity and the presence of entrained TBP on the quality of precipitated ADU particles. It was found that as the concentration of free acid as well as the concentration of entrained TBP in UNPS increases, the particle size of precipitated ADU decreases. Based on the experimental results two correlations were developed to determine the mean particle size of ADU; one is based on the free acid content of UNPS and the other is based on the content of entrained TBP in UNPS, which is used for the precipitation. It was found that the correlated values are well fitted with the experimental data within ±3% errors for both the cases. Both these correlations are applicable when other process parameters remain constant. The experimental details and results are discussed in this paper. © 2015 Published by Elsevier B.V.

Sudhakar Rao G.,Banaras Hindu University | Chakravartty J.K.,Bhabha Atomic Research Center | Nudurupati S.,Nuclear Fuel Complex | Mahobia G.S.,Banaras Hindu University | And 3 more authors.
Journal of Nuclear Materials | Year: 2013

Fuel cladding and pressure tubes of Zircaloy-2 in pressurized light and heavy water nuclear reactors experience plastic strain cycles due to power fluctuations in the reactor, such strain cycles cause low cycle fatigue (LCF) and could be life limiting factor for them. Factors like strain rate, strain amplitude and temperature are known to have marked influence on LCF behavior. The effect of strain rate from 10-2 to 10-4 s-1 on LCF behavior of Zircaloy-2 was studied, at different strain amplitudes between ±0.50% and ±1.25% at room temperature. Fatigue life was decreased with lowering of strain rate from 10-2 to 10-4 s-1 at all the strain amplitudes studied. While there was cyclic softening at lower strain amplitudes (Δεt/2 ≤ ±0.60%) cyclic hardening was exhibited at higher strain amplitudes (Δεt/2 ≥ ±1.00%) at all the strain rates. Further, there was secondary cyclic hardening during the later stage of cycling at all the strain amplitudes and the strain rates. Cyclic stress-strain hysteresis loops at the lowest strain rate of 10-4 s-1 were found to be heavily serrated, resulting from dynamic strain aging (DSA). There was significant effect of strain rate on dislocation substructure. The results are discussed in terms of high concentration of point defects generated during cyclic straining and their role in enhancing interaction between solutes and dislocations. © 2013 Elsevier B.V. All rights reserved.

Mani Krishna K.V.,Materials Science Division | Srivastava D.,Materials Science Division | Dey G.K.,Materials Science Division | Hiwarkar V.,IITB | And 2 more authors.
Journal of Nuclear Materials | Year: 2011

Various methods of Kearns "f" parameter evaluation were compared for their consistency and dependency on measurement cross section of the sample and variation in the microstructure across different cross sections. The study showed that, EBSD (Electron Back Scattered Diffraction) method is more consistent in comparison to X-ray based techniques for the "f" parameter determination especially in case of recrystallized microstructures. © 2011 Elsevier B.V. All rights reserved.

Limbadri K.,MREC | Gangadhar J.,Griet | Maruti Ram A.,Nuclear Fuel Complex | Singh S.K.,Griet
Materials Today: Proceedings | Year: 2015

Sheet metal forming is a very important step in manufacturing a variety of components used in automobile, nuclear industries and so on. Usually the formability of any material is a property which primarily depends upon work hardening and the type of texture in the material. The type of texture affects the material properties and specially the formability of the material but these aspects are not studied by any researcher. In the present article a review has been carried out to study the formability of material with respect to texture and how it is related to the formability of material. © 2015.

Tonpe S.,Nuclear Fuel Complex | Mudali U.K.,Indira Gandhi Center for Atomic Research
Materials and Manufacturing Processes | Year: 2016

This paper presents the methodology used for manufacturing the Zircaloy-4 tubes required for the Zircaloy-4 mock-up dissolver assembly. The evolution of the microstructure at different stages of production of Zircaloy-4 tubes was characterized using scanning electron microscopy (SEM). Microstructural evolution follows the sequence as dendritic structure (as-cast ingot) → Widmanstätten structure (β-quenched) → bimodal grain size (hot extruded) → heterogeneous deformed structure (pilgered) → partially recrystallized structure (final annealed). High strength and low ductility were obtained for the Zircaloy-4 tubes at pilgered condition due to grain size refinement and work hardening during the pilgering process. Compared with the pilgering stage, the Zircaloy-4 tubes in the final annealed condition exhibited moderate strength and high ductility due to the partially recrystallized microstructure. Autoclaving was carried out to improve the corrosion properties of the pilgered Zircaloy-4 tubes in boiling 11.5 M nitric acid. When exposed to boiling 11.5 M nitric acid for 1000 h, a lowest corrosion rate of 0.003 mpy was obtained for autoclaved Zircaloy-4 tubes. Laser Raman Spectroscopy (LRS) analysis confirmed that the origin of passivity of pilgered and autoclaved Zircaloy-4 tubes was due to the presence of a protective passive film composed of ZrO2. 2017 Copyright © Taylor & Francis Group, LLC

Tonpe S.,Nuclear Fuel Complex | Saibaba N.,Nuclear Fuel Complex | Jayaraj R.N.,Nuclear Fuel Complex | Ravi Shankar A.,Indira Gandhi Center for Atomic Research | And 2 more authors.
Energy Procedia | Year: 2011

Spent fuel reprocessing for Fast Breeder Reactor (FBR) requires a dissolver made of a material which is resistant to corrosion as the process involves nitric acid as the process medium. The focus was on the use of advanced materials with high performance in high concentrations (greater than 8 N) and high temperatures of nitric acid employed in the dissolver. The commercially available materials like SS316l, titanium, Ti - 5% Ta, and Ti - 5% Ta - 1.8% Nb were tried and the corrosion behavior of these materials was studied in detail. It was also decided to try out Zircaloy - 4 as the material of construction due to its excellent corrosion resistance in nitric acid environment. The specifications were stringent and the geometrical configurations of the assembly were very intricate. On accepting the challenge of fabricating the dissolver, NFC made different fixtures for electron beam (EB) welding and TIG welding. Trials were carried out for optimization of various operating parameter for EB welding & TIG welding processes, which were qualified by radiography, liquid dye penetrant testing, and metallographic, mechanical and chemical analys es. Finally the dissolver was subjected to helium leak test. The tests conducted at IGCAR with Zircaloy dissolver assembly revealed that Zircaloy-4 exhibited the superior corrosion resistance over all other materials tried out so far for this purpose. The results clearly indicates that Zircaloy - 4 is a near-zero corrosion material in comparison to the other materials. It is thus demonstrated that the dissolver assembly for spent fuel reprocessing for Fast Breeder Reactor (FBR) can be manufactured from Zircaloy - 4 on an industrial scale, with long life expectancy with near zero corrosion, employing EB & TIG welding processes. © 2011 Published by Elsevier Ltd.

Santra S.,Nuclear Fuel Complex | Ramana Rao S.V.,Nuclear Fuel Complex | Kapoor K.,Nuclear Fuel Complex
Materials Performance and Characterization | Year: 2016

Alloy 690 in the thermally treated (TT) condition is an advanced steam generator tubing material. The effect of thermal ageing treatments on the intergranular corrosion (IGC) behavior of the alloy was probed. Isothermal ageing was carried out at 600, 700, and 800 ô C with varying ageing time. Morphology of the carbides was studied as a function of thermal treatment temperature and duration. The carbides morphology was observed to affect the IGC resistance. Grain boundaries devoid of carbides (solutionized condition) or with coarse carbide particles (formed at 800°C) enhanced IGC resistance, whereas dendritic fine carbides formed at lower ageing temperature yield inferior IGC resistance. Copyright © 2016 by ASTM International.

Nayak I.K.,Nuclear Fuel Complex | Rao S.V.R.,Nuclear Fuel Complex | Kapoor K.,Nuclear Fuel Complex
Materials Performance and Characterization | Year: 2016

The alloy 800 is being used for steam generator tubings for pressurized heavy-water reactors (PHWRs). This material is susceptible to degradation because of severe operating conditions, like high temperature, stress, and corrosive environment. These degradation mechanisms include primary water stress corrosion cracking (PWSCC), secondary side or outer diameter stress corrosion cracking (ODSCC), inter-granular corrosion (IGC), fretting, wear, denting, high cycle fatigue, corrosion fatigue, etc. The present study was conducted to analyze the effect of solution annealing temperatures, sensitization treatments, and surface conditions on the corrosion rate. The alloy 800 tubular samples were investigated by means of a conventional corrosion test (according to ASTM G28-02) and an electrochemical potentiokinetic reactivation (EPR) test. Susceptibility to inter-granular corrosion under various experimental parameters was examined by using both the test methods and the results are compared. The observed trends in corrosion rate obtained by using conventional method and EPR test were found to be similar. These results were used to obtain most optimum heat-treatment parameters and surface condition, which will yield best corrosion resistance. Copyright © 2016 by ASTM International.

Rao J.V.,Nuclear Fuel Complex | Chandraiah K.,Sri Venkateswara University
Indian Journal of Occupational and Environmental Medicine | Year: 2012

Backround: Experience of occupational stress is inevitably involved in the execution of any type of work. Stress has an adaptive value. It motivates the individual to attend to the task and get rid of the tension or demand the unattended task produced. Materials and Methods : The study was planned to investigate the differences between executives and shop floor workers on occupational stress, mental health, job satisfaction and coping. A random sample of 200 executives and shop floor employees collected from Nuclear Fuel Complex of Hyderabad City. A well developed sub-scales of Occupational Stress indicator like Mental Health, and Coping behavior were used in the present study. Results and Conclusion : The shop floor workers experiencing more job stress and lower mental health. But these two groups did not differ in their coping behaviour. The executives are better with work home balance.

Jayaraj R.N.,Nuclear Fuel Complex
Energy Procedia | Year: 2011

Presently the installed capacity of electricity generation in India is 160 GWe. The integrated energy policy aims at an installed capacity of 778 GWe by 2031-32 to achieve per capita electricity consumption of 2700 kWh/year as against the 700 kWh/year. The share of nuclear power in India is around 3% presently. The increase in installed capacity is possible by increasing the share of nuclear power. India has opted for a unique three-stage power programme based on closed nuclear fuel cycle, which provides a multiplier effect through breeding. Consequent to the 123 agreement and clearance from Nuclear Supplier's Group (NSG) for international cooperation in the field of nuclear energy, the reactor technology options are wide open. Some of the technologies which are being pursued are: Pressurized Heavy Water Reactors (PHWRs) fueled with domestic Natural Uranium (NU), PHWRs fueled with imported Natural Uranium/Slightly Enriched Uranium (SEU), Light Water Reactors (LWRs) procured from abroad using imported Lightly Enriched Uranium (LEU) fuel, PHWRs with Reprocessed Uranium (RU) obtained from reprocessing spent fuel of LWRs, indigenous Pressurized Water Reactors (PWRs) and Fast Breeder Reactors (FBRs) using MOX/metallic fuel. The increase in the share of nuclear power through above technologies require setting up of fuel fabrication facilities such as: new PHWR Fuel Fabrication with indigenous sources, joint collaboration with foreign countries for fabricating fuel for imported LWRs, setting up PHWR fuel fabrication facilities with SEU/RU under international safeguards, setting up of enrichment and fuel fabrication facilities for indigenous PWRs and a series of fast reactor fuel fabrication facilities for fabricating fuel for FBRs. In addition to fuel fabrication facilities, facilities for manufacturing zirconium alloy and stainless steel structurals, tubes and components are also required to be set up. The paper gives in detail the emerging nuclear fuel fabrication activities in India. © 2011 Published by Elsevier Ltd.

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