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Operating Experience of Fast Breeder Test Reactor

[+] Author Affiliations
K. V. Suresh Kumar, V. Ramanathan, G. Srinivasan

IGCAR - Indira Gandhi Centre for Atomic Research, Kalpakkam, India

Paper No. ICONE17-75592, pp. 893-898; 6 pages
  • 17th International Conference on Nuclear Engineering
  • Volume 1: Plant Operations, Maintenance, Engineering, Modifications and Life Cycle; Component Reliability and Materials Issues; Next Generation Systems
  • Brussels, Belgium, July 12–16, 2009
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4351-2 | eISBN: 978-0-7918-3852-5
  • Copyright © 2009 by ASME


Sodium cooled fast breeder reactors constitute the second stage of India’s three-stage nuclear energy programme, for effective utilization of the country’s limited reserves of natural uranium and exploitation of its large reserves of thorium. The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, is a sodium cooled, loop type fast reactor. Its main aim is to provide experience in fast reactor operation and large scale sodium handling and to serve as a test bed for irradiation of fast reactor fuels and materials. FBTR was built on the lines of the French Rapsodie-Fortissimo reactor, with modifications to make it a generating plant. FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. The steam water system mainly consists of a once-through steam generator, which produces super heated steam at a pressure of 125 bars and temperature of 480° C, feed water system and condensate system. The steam produced is supplied to a condensing turbine coupled to an alternator. The reactor achieved first criticality in Oct 85 with a small core of 22 fuel subassemblies (SA) having a unique carbide fuel rich in Pu. This fuel (called MK-I) was developed and made in India and has a composition of 70% PuC-30% UC. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. In the light of the excellent performance of the carbide fuel, which has endured a burn-up of 155 GWd/t without any clad failure, the core has been gradually expanded by the addition of mark II (55% PuC-45% UC) and MOX (44% PuO2 −56% UO2 ) fuel SA to compensate for the burn-up reactivity loss, and the reactor power has been progressively raised from 10.6 MWt to a maximum of 17.4 MWt. Over the years, several safety related experiments have been conducted. These include natural convection tests and experiments to validate the Failed Fuel Detection System. The challenges faced include a major fuel handling incident, primary sodium leak and reactivity transients. Two major modifications were carried out — one on the Steam Generator Leak Detection System and the other in the steam-water circuit. These helped in improving the campaign availability from less than 50% to more than 90%. The main component limiting the life of reactor is the grid plate supporting the core. The fast flux at the grid plate was measured using Np foils. The residual life of the grid plate has been estimated as 11 effective full power years. The paper presents operating experience of the reactor, performance of the carbide fuel, safety related experiments done in the reactor, various challenges faced, various modifications carried out to improve system reliability and availability and residual life assessment of the reactor.

Copyright © 2009 by ASME



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