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.
Stephen N.H.,Indira Gandhi Center for Atomic Research |
Sathiyasheela T.,Indira Gandhi Center for Atomic Research |
Paul D.,Indira Gandhi Center for Atomic Research |
Devan K.,Indira Gandhi Center for Atomic Research |
And 2 more authors.
Annals of Nuclear Energy | Year: 2014
This study focuses on computing static reactivity coefficients and analyzing Unprotected Loss of Flow Accident in a Th-Pu fuelled metal reactor. An attempt is also done to compare the static and dynamic performance of the fresh core with characteristics of U-Pu-6Zr fuelled 500 MWe metal reactor. Isothermal temperature coefficient and power coefficients are evaluated in the steady state and found to be negative. The excess reactivity and control rod worth requirements of Th-Pu metal core are assumed similar to that of U-Pu-6Zr metal core. In the Unprotected Loss of Flow Accident (ULOFA) analysis, with flow coast down initiated by station black out, it is found that power to flow ratio is increasing initially up to 53 s and then starts to reduce continuously. Power to flow ratio is found to be less than 2 at all times thus ensuring the absence of coolant boiling in the entire core. Sodium voiding starts around 886 s in the upper axial blanket and provide negative reactivity. Also it will not propagate to the core center ensuring the probability for core disruptive accident a remote one. Net reactivity feedback is negative and the major contribution is from core radial expansion. Within 12 min, the power drops to 32 MWt, making it possible for Safety Grade Decay Heat Removal (SGDHR) system to start heat removal from core ensuring safe shutdown of reactor. Sensitivity analysis by considering an uncertainty margin of ±10% in thermo physical properties of fuel composition shows that feedback reactivity of the Th-Pu system is insensitive and the conclusion on the safe shutdown remains unaltered. From this study it is found that inherent safety of Th-Pu metal fuel core is better than that of reactor core fuelled with U-Pu-6Zr metal type under ULOFA condition. © 2014 Elsevier Ltd. All rights reserved.
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.