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The Nuclear Power Corporation of India Limited is a government-owned corporation of India based in Mumbai.It is wholly owned by the Central Government and is responsible for the generation of nuclear power for electricity. NPCIL is administered by the Department of Atomic Energy, Govt. of India . NPCIL was created in September 1987 as under the Govt. Act 1956, "with the objective of undertaking the design, construction, operation and maintenance of the atomic power stations for generation of electricity in pursuance of the schemes and programmes of the Government of India under the provision of the Atomic Energy Act 1962." All nuclear power plants operated by the company are certified for ISO-14001 . NPCIL was the sole body responsible for constructing and operating India's commercial nuclear power plants till setting up of BHAVINI. As of 10 August 2012 the company had 21 nuclear reactors in operation at seven locations, a total installed capacity of 5780 MWe. Subsequent to the government's decision to allow private companies to provide nuclear power, the company has experienced problems with private enterprises "poaching" its employees. Wikipedia.

Mishra S.,Nuclear Power Corporation of India | Modak R.S.,Bhabha Atomic Research Center | Ganesan S.,Bhabha Atomic Research Center
Nuclear Science and Engineering | Year: 2012

Large-sized pressurized heavy water reactors (PHWRs) are neutronically loosely coupled and hence are prone to significant changes in flux shape during operation. As a result, they need a sophisticated regulation procedure based on an online flux mapping system (OFMS). During the reactor operation, neutron flux is continuously measured at certain predetermined in-core locations. The purpose of OFMS is to compute a detailed flux map at all points in the reactor, after every 2 min, by making use of the measured fluxes. The knowledge of detailed flux distribution is then used for an appropriate regulating action. The choice of computational method used by OFMS is of crucial importance because the method is expected to be both efficient and accurate and should work for a range of reactor configurations occurring during the operation. In this paper, three different methods, namely, flux synthesis, internal boundary condition, and combined least squares (CLSQ), are analyzed for their prospective use in the forthcoming 700-MW(electric) Indian PHWR. The CLSQ method is found to be most accurate, although it needs significant computation. A hybrid method that combines certain features of other methods is also studied and seems to give good accuracy with moderate computational effort. Source

Chaudhry V.,Indian Institute of Science | Chaudhry V.,Nuclear Power Corporation of India | Kailas S.V.,Indian Institute of Science
Wear | Year: 2013

In a practical situation, it is difficult to model exact contact conditions due to challenges involved in the estimation of contact forces, and relative displacements between the contacting bodies. Sliding and seizure conditions were simulated on first-of-a-kind displacement controlled system. Self-mated stainless steels have been investigated in detail. Categorization of contact conditions prevailing at the contact interface has been carried out based on the variation of coefficient of friction with number of cycles, and three-dimensional fretting loops. Surface and subsurface micro-cracks have been observed, and their characteristic shows strong dependence on loading conditions. Existence of shear bands in the subsurface region has been observed for high strain and low strain rate loading conditions. Studies also include the influence of initial surface roughness on the damage under two extreme contact conditions. © 2013 Elsevier B.V. Source

Rao G.N.,Nuclear Power Corporation of India
Procedia Engineering | Year: 2014

In nuclear power plants, continuous efforts are made to ensure that plants are operated in safe, reliable and economical manner. While utmost care is taken during design, construction and commissioning of the structures, systems & components (SSCs), continued healthiness has to be ensured during operation phase in accordance with the design intent. This is primarily achieved through the establishment of a comprehensive life management programme of surveillance, condition monitoring, periodic Inservice inspections (ISI) and maintenance, the purpose of which is to ensure that required safety margins are maintained for all important SSCs throughout plant service life. In any industry, the important mechanisms which lead to SSCs degradation are general corrosion, Flow accelerated corrosion, erosion, thermal effects, fatigue and mechanical wear & fretting. In a Nuclear Power Plant, in addition to these, degradation also occurs due to irradiation and creep. Therefore SSCs need to be designed with due consideration of all such degradation mechanisms. During operation phase, the limiting conditions for operation and requirements of surveillance & condition monitoring for all components important to safety are documented in Technical Specification. A plant specific ISI manual defines and elaborates the ISI internal inspection methods and acceptance criteria. The ISI requirements are decided taking into account best industry practices and operational experience and meet the requirements of existing codes and standards. Both these documents are approved by AERB. In pressurized heavy water reactors, major components covered in ISI programme are coolant channels, feeder pipes, steam generators, heavy water heat exchangers, pressure boundary components, relief valves etc. The inspection methods adopted include use of UT, ECT, DPT and visual inspection. In addition to functional checks for some of the components, ferrography, vibration monitoring and thermography etc are also utilized for rotating equipment. The SSCs where unacceptable indications are revealed by ISI are repaired, replaced or isolated. The base line data generated during Pre-Service inspection is used for comparing and trending of observations. The observations made during ISI are subjected to thorough review and analysis by qualified experts to obtain assurance that unacceptable degradation in component quality is not occurring and it remains fit for service. The containment system of a NPP plays a crucial role in minimizing dose to the public in case of an accident situation. The integrity of Primary and Secondary Containment is assessed by conducting integrated leak rate tests in every Biannual Shut Down. The concrete structure is also subjected to NDT checks on a specified frequency to ensure continued healthiness. Indian nuclear power plants have accumulated an operating experience of more than 380 reactor years of operation. Over the years, the programmes for surveillance, condition monitoring and in-service inspections have been improved significantly and match with best in the industry. As a result, there has been no major age related failure in any of the important SSCs and thus continued safe and reliable operation of NPPs is assured. This paper brings out the ISI & health monitoring methodology adopted at Indian NPPs, to ensure safe and reliable operation. © 2014 Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license.. Source

Chaudhry V.,Nuclear Power Corporation of India | Kailas S.V.,Indian Institute of Science
Wear | Year: 2015

Fretting is of a serious concern in many industrial components, specifically, in nuclear industry for the safe and reliable operation of various component and/or system. Under fretting condition small amplitude oscillations induce surface degradation in the form of surface cracks and/or surface wear. Comprehensive experimental studies have been carried out simulating different fretting regimes under ambient and vacuum (10-9MPa) conditions and, temperature up to 400°C. Studies have been carried out with stainless steel spheres on stainless steel flats, and stainless steel spheres against chromium carbide, with 25% nickel chrome binder coatings. Mechanical responses are correlated with the damage observed. It has been observed that adhesion plays a vital role in material degradation process, and its effectiveness depends on mechanical variables such as normal load, interfacial tangential displacement, characteristics of the contacting bodies and most importantly on the environment conditions. Material degradation mechanism for ductile materials involved severe plastic deformation, which results in the initiation or nucleation of cracks. Ratcheting has been observed as the governing damage mode for crack nucleation under cyclic tangential loading condition. Further, propagation of the cracks has been observed under fatigue and their orientation has been observed to be governed by the contact conditions prevailing at the contact interface. Coated surfaces show damage in the form of brittle fracture and spalling of the coatings. Existence of stick slip has been observed under high normal load and low displacement amplitude. It has also been observed that adhesion at the contact interface and instantaneous cohesive strength of the contacting bodies dictates the occurrence of material transfer. The paper discusses the mechanics and mechanisms involved in fretting damage under controlled environment conditions. © 2015 Elsevier B.V. Source

Chhatre A.G.,Nuclear Power Corporation of India
Exploration and Research for Atomic Minerals | Year: 2014

The Nuclear Power Plant (NPP) is designed for two levels of earthquake viz., Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE). The OBE (SI level ground motion) corresponds to the maximum level of ground motion, which can reasonably be experienced at the site once during the operating life of nuclear power plant with a return period of 100 years. The SSE (S2 level ground motion) represents the maximum level of ground motion to be used for design of safety related structures, systems and equipment (SS&E) of NPP and is based on the maximum earthquake potential of the region, with a return period of 10,000 years. For these two levels of earthquakes, it is required to determine Peak Ground Acceleration (PGA) and thereby, specify the Design Basis Ground Motion (DBGM). In order to determine the PGA, seismotectonic study is of utmost importance. A study of regional geology and seismotectonic features in the region of 300 km radius from site is conducted to delineate the faults/lineaments based on the study of satellite imageries, Seismotectonic Atlas of India and its Environs, and other published or unpublished maps/reports. Earthquake data (historical as well as recorded) along with Micro-Earthquake (MEQ) data in the region are also collected from various sources and a seismotectonic map is generated. This is followed by field check study which is carried out at three levels viz., regional (300 km radius), intermediate (50 km radius) and local (5 km radius). Based on the field check study, capable (active) faults are identified and a maximum earthquake potential (in terms of magnitude) and depth of focus are assigned to the identified capable faults. These two data clubbed with shortest distance of the capable fault is used to determine the PGA. The DBGM is determined deterministically by three methods viz., using recorded time histories from the sites having similar geological and seismological features, by generating synthetic ground motion and by attenuation correlation. In the time history based method, a normalized mean plus one sigma spectral shape is determined for the range of S2 level of earthquake which is multiplied by site PGA based on the controlling earthquake to get the DBGM. In case of the attenuation correlation based method, the normalized mean plus one sigma spectra is generated for the scenario based earthquake which is then anchored to the mean PGA of the controlling earthquake to get the DBGM. The aim of generating a normalized mean plus one sigma spectra for such a range of S2 level earthquake time histories is to generate a mean plus one standard deviation spectra i.e. to define Dynamic Amplification Factors (DAF) with 84% non exceedance probability at all the frequencies. The normalized mean plus one sigma spectral shape is then anchored to the mean PGA derived for the controlling earthquake which is maximized for the magnitude and brought to the minimum epicentral distance by moving this earthquake on to a fault, closest to the plant site. The present paper brings out the procedure for conducting field check study, determination of the ground motion and also a case study of field check carried out and generation of ground motion for KAPP-3&4 NPP site. © Director, AMD, DAE, Govt. of India. Source

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