Bharatiya Nabhikiya Vidyut Nigam Ltd

Kalpakkam, India

Bharatiya Nabhikiya Vidyut Nigam Ltd

Kalpakkam, India

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Prasad T.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Rana K.P.S.,Netaji Subhas Institute of Technology | Kumar V.,Netaji Subhas Institute of Technology
Measurement: Journal of the International Measurement Confederation | Year: 2017

In a recent work, Kumar et al. have presented design and modeling of an intelligent temperature to frequency converter (TFC) for thermistor in the input temperature range of 0–100 °C [1]. In this work, a 555-timer based a signal conditioning circuit (SCC) was employed as TFC. Linearity, between the input temperature and the output frequency, was further enhanced with the application of artificial neural network. It is pointed out that this work developed erroneous frequency expression, waveforms and SCC circuit. In this comment, based on detailed mathematical formulations supported by Multisim™ simulation results, errors of the reported intelligent TFC are pointed out and the required corrections are proposed. © 2017 Elsevier Ltd


Rana K.P.S.,Netaji Subhas Institute of Technology | Kumar V.,Netaji Subhas Institute of Technology | Prasad T.,Bharatiya Nabhikiya Vidyut Nigam Ltd.
IEEE Sensors Journal | Year: 2017

Recently, Kumar et al. [1] have reported an Artificial Neural Network (ANN) based linearization technique for Voltage Controlled Oscillator (VCO) thermistor circuit. In this work ANN was used to linearize the frequency output of VCO, driven by the voltage across thermistor, for an input temperature range of 0 °C-100 °C. Performance of the developed technique was claimed to be experimentally verified for sensitivity and linearity of 5 kHz/°C and ±0.2%, respectively, for all the investigated three different thermistors (M/S Phillips make). In this comment, using the data provided in [1] and based on the relevant theoretical analysis it is shown the obtainable theoretical sensitivities are 136.918 Hz/°C, 108.919 Hz/°C and 267.938 Hz/°C for the used thermistor-1, thermistor-2 and thermistor-3, respectively, instead of the claimed 5 kHz/°C. © 2016 IEEE.


Sivasailanathan V.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Kumar P.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Manoharan N.,Research and Development Division University
Biosciences Biotechnology Research Asia | Year: 2014

Radiation has become inseparable part of the living environment contributing as cosmic rays, terrestrial radiation, fall-out from earlier nuclear accidents and testing, the increased use of diagnostic radiology etc., Since early ages of time, the homosapiens have lived with natural radiation and his system has been adapted for the surrounding radiation and its effects. But still, radiation and its effects gain attraction in the minds of the general public because of the baseless panic registered through passage of time regarding the uncertainties associated with the consequences of radiation and the knowledge inadequacy on handling of the same. Operating nuclear power plants are preferably considered as the elemental portal to elucidate the effects of radiationas a fairly large quantity of radioactive materials are being used as fuel in the reactors. The main reason of the so called menace is because of the strong reason that radiation is nonsensory. Only instruments can detect the presence of radioactivity and measure the level of the radiation field. The dose reduction techniques in the occupational environment involve three major concepts: 1. Time 2. Distance and 3. Shielding. Apart from these control measures and though there are various other protective measures to protect from contamination by using protective clothing, respirators, etc., the dose reduction scheme has to be built in the design of the nuclear power reactor. The intensive practice and strict compliance to the radiation protection aims at bringing the exposure level As Low As Reasonably Achievable (ALARA). The paper brings out the features of ventilation methodology adopted in the prototype fast breeder reactor and the effective identification of radiological zoning to contain the contamination. The paper also tries to justify that the appropriate design strategyhelps in effective dose reduction in operating nuclear power plants.


Chellapandi P.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Rao P.R.V.,Indira Gandhi Center for Atomic Research | Kumar P.,Bharatiya Nabhikiya Vidyut Nigam Ltd
Pramana - Journal of Physics | Year: 2015

Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper. © Indian Academy of Sciences.


Aithal S.,Indira Gandhi Center for Atomic Research | Babu V.R.,Indira Gandhi Center for Atomic Research | Balasubramaniyan V.,Indira Gandhi Center for Atomic Research | Velusamy K.,Indira Gandhi Center for Atomic Research | Chellapandi P.,Bharatiya Nabhikiya Vidyut Nigam Ltd
Nuclear Engineering and Design | Year: 2016

An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼φ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR. © 2016 Elsevier B.V. All rights reserved.


Raj B.,National Institute of Advanced Studies | Chellapandi P.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Mudali U.K.,Indira Gandhi Center for Atomic Research
Procedia CIRP | Year: 2015

The increasing demands on reliability, safety and economics of nuclear systems translate to challenges in realization of high-performance components for operation at steady state, transient and severe accident conditions. The authors, based on their four decades of research, development and deployment experiences, present a review of findings relating to life cycle management of critical structural components in Indian thermal, fast reactors and reprocessing facilities. The challenges relating to specific structural components are described with highlights of materials to improve life for prolonged service with safety and economics. © 2015 The Authors. Published by Elsevier B.V.


Choudhry N.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Riyas A.,Indira Gandhi Center for Atomic Research | Devan K.,Indira Gandhi Center for Atomic Research | Mohanakrishnan P.,Indira Gandhi Center for Atomic Research
Nuclear Engineering and Design | Year: 2013

Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu-U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs) is made along with a design change in CSRs to increase the reactivity worth of absorber rods for ensuring safe reactor operation. This new enhanced burnup core can be used in the future with advanced structural materials for clad and hexcan, which can withstand higher neutron fluence and gamma radiation. © 2012 Elsevier B.V.


Anoop B.,Bharatiya Nabhikiya Vidyut Nigam Ltd. | Balaji C.,Indian Institute of Technology Madras | Velusamy K.,Indira Gandhi Center for Atomic Research
Annals of Nuclear Energy | Year: 2015

Conjugate heat transfer from serrated fins on the outside of the tubes of a sodium to air tubular heat exchanger of sodium cooled fast breeder reactors, has been investigated by combined experimental and computational approaches. For the latter approach, the RNG k-ε model, which is applicable for a wide range of Reynolds numbers, was used for turbulence closure. The numerical model employed was validated by conducting in-house heat transfer experiments on a single serrated finned tube. A detailed parametric study has been carried out to investigate the effect of serration depth, fin pitch, fin height and fin thickness. In addition to pure cross flow, the effect of angle of attack of the flow on the heat transfer also has been studied. A correlation for determining the Nusselt number over a serrated finned tube has been proposed taking into account the serration parameters. This is expected to be useful in the design of sodium to air heat exchangers of fast breeder reactors. © 2015 Elsevier Ltd. All rights reserved.


Rangasamy R.G.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Kumar P.,Bharatiya Nabhikiya Vidyut Nigam Ltd
Advanced Materials Research | Year: 2013

Austenitic stainless steels are the major material of construction for the fast breeder reactors in view of their adequate high temperature mechanical properties, compatibility with liquid sodium coolant, good weldability, availability of design data and above all the fairly vast and satisfactory experience in the use of these steels for high temperature service. All the Nuclear Steam Supply System (NSSS) components of FBR are thin walled structure and require manufacture to very close tolerances under nuclear clean conditions. As a result of high temperature operation and thin wall construction, the acceptance criteria are stringent as compared to ASME Section III. The material of construction is Austenitic stainless steel 316 LN and 304 LN with controlled Chemistry and calls for additional tests and requirements as compared to ASTM standards. Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type, 500 MWe reactor which is at advanced stage of construction at Kalpakkam, Tamilnadu, India. In PFBR, the normal heat transport is mainly through two secondary loops and in their absence; the decay heat removal is through four passive and independent safety grade decay heat removal loops (SGDHR). The secondary sodium circuit and the SGHDR circuit consist of sodium tanks for various applications such as storage, transfer, pressure mitigation and to take care of volumetric expansion. The sodium tanks are thin walled cylindrical vertical vessels with predominantly torispherical dished heads at the top and bottom. These tanks are provided with pull-out nozzles which were successfully made by cold forming. Surface thermocouples and heaters, wire type leak detectors are provided on these tanks. These tanks are insulated with bonded mineral wool and with aluminum cladding. All the butt welds in pressure parts were subjected to 100% Radiographic examination. These tanks were subjected to hydrotest, pneumatic test and helium leak test under vacuum. The principal material of construction being stainless steel for the sodium tanks shall be handled with care following best engineering practices coupled with stringent QA requirements to avoid stress corrosion cracking in the highly brackish environment. Intergranular stress corrosion cracking and hot cracking are additional factors to be addressed for the welding of stainless steel components. Pickling and passivation, Testing with chemistry controlled demineralised water are salient steps in manufacturing. Corrosion protection and preservation during fabrication, erection and post erection is a mandatory stipulation in the QA programme. Enhanced reliability of welded components can be achieved mainly through quality control and quality assurance procedures in addition to design and metallurgy. The diverse and redundant inspections in terms of both operator and technique are required for components where zero failure is desired & claimed. This paper highlights the step by step quality management methodologies adopted during the manufacturing of high temperature thin walled austenitic stainless steel sodium tanks of PFBR. © (2013) Trans Tech Publications, Switzerland.


Rangasamy R.G.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Kumar P.,Bharatiya Nabhikiya Vidyut Nigam Ltd | Manoharan N.,AMET University
Biosciences Biotechnology Research Asia | Year: 2014

The term "Quality management" has a specific meaning within many business sectors. Quality has metamorphosed from the synonyms of "Customer satisfaction" to "Customer delight", thriving for excellence in every sphere of business with continuous improvements. In today's business scenario, more often quality is perceived as "Fitness for purpose "focusing on the customer. But there is horizon beyond this as, "quality" is always intertwined with "safety and reliability" if the nature of the business is perilous. The prime focus for any nuclear industry is about the safety and reliability which can be accomplished only through the inherent quality. The quality embarks right from the construction phase of the nuclear power plant till the decommissioning and the four main components i.e. quality planning, quality control, quality assurance and quality improvement trek alongside. The high temperature low pressure system of fast breeder reactors using sodium as the coolant demands very high reliability and high degree of quality during each and every stage of construction of all the individual components for trouble free operation of the reactor for the committed years. The "Quality Assurance (QA)" plan of Prototype Fast Breeder Reactor is unique. The reactor being first of its kind in India, quality assurance starts right from the raw material procurement and extends through all the stages of plant till commissioning. The principal material of construction being stainless steel for the reactor components shall be handled with care following best engineering practices coupled with stringent QA requirements to avoid stress corrosion cracking in the highly brackish environment. Integranular stress corrosion cracking and hot cracking are additional factors to be addressed for the welding of stainless steel components. The low alloy ferritic steel like 9Cr-1Mo (mod) has been extensively deployed and the fabrication requires structured inspection, testing and QA plan. Corrosion protection and preservation during fabrication, erection and post erection is mandatory be it reinforcement bar or a reactor vessel.

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