FAST BREEDER TEST REACTOR
FULFILLS ALL TECHNOLOGY OBJECTIVES WITH SMALL CARBIDE CORE 
In April 1999, with the attainment of 50,000 MWd/t ( ~ 5.5 atom %) burnup of indigenously developed Pu-U monocarbide Mark-I fuel (70% PuC – 30% UC) without any fuel clad failure, FBTR has fulfilled all technology objectives set with small carbide core. FBTR is a 40 MWt ( 13 MWe) loop type, liquid sodium cooled, mixed carbide fuelled reactor (Fig 1) presently being operated with a small carbide core (Fig 2). Its main objectives are to generate experience in operation of large sodium systems, to demonstrate generation of electricity from this technology and to serve as an irradiation facility for developing fuel and structural materials for future fast reactors. Since first criticality in October 1985, the reactor has completed seven irradiation campaigns with the maximum operating power of 12 MWt at a maximum linear heat rating of 320 W/cm. After the sodium heated steam generator was put in service for the first time in January 1993, all the set objectives were systematically and progressively achieved during the past six years ( Table 1).
 

TABLE– 1
 MILESTONES ACHIEVED
 
S.No
Description
Date
1
 Steam generator put in service Jan 93
2
 Full power for small core achieved Dec 93
3
 Completion of high power physics and engineering tests Jun 94
4
 Target burnup of 25,000 MWd/t for Mark I fuel achieved Jun 96
5
 First fuel irradiation programme completed Apr 97
6
 TG synchronisation to grid Jul 97
7
 First material irradiation programme completed Apr 97
8
 Revised target burnup of 50,000 MWd/t for Mark I fuel achieved Apr 99
 
The primary and secondary sodium coolant circuits have been in service for the past 14 years at a maximum temperature of 420° C. Nuclear grade purity of sodium has been well maintained as indicated by the on-line plugging meters, periodic impurity analysis by sampling and mirror like sodium surface as viewed through the periscope. There has been no radioactive primary sodium leak so far. Sodium service components have worked excellently and in fact one of the four mechanical sodium pumps crossed 1,000,000 hours of trouble free operation in February 1999. One of the most critical sodium service component viz the sodium heated steam generator has worked satisfactorily for 7500 h without any steam / water tube leak. The procedures for handling fuel and other sodium components are well established following the cardinal principles of maintaining leaktightness during handling operation and adequate shielding to prevent radioactive exposure to personnel. Operating liquid sodium systems, having an inventory of 150 tonnes, for such a long period without any incident indicates that sodium technology has been well and truly mastered.

After achieving stabilised parameters at full power with small core in December 1993, it was mandatory to carry out safety related high power physics and engineering tests to validate the various assumptions made in safety report and to ensure plant safety under anticipated operational transients. This testing programme was completed in June 1994 including confirmation of conduciveness of primary and secondary sodium circuits for establishment of natural convection for safe decay heat removal. Tests with various plant incidental situations were as per design predictions and helped in validation and improvement of plant dynamic codes developed in house.

With the completion of balance construction and precommissioning works and attaining superheated steam parameters ( 400° C temperature and 120 Kg/cm2 pressure), turbine rolling trials were started during 1996 and after satisfactory completion of all commissioning activities on turbine and generator, the TG was synchronised to the southern grid on 11th July 1997. This was the first time that electricity from a fast reactor in India was fed to the grid. Total time of synchronisation was about 20 days at a maximum power of 1.2 MWe. This historic event ushered in the second phase of India’s nuclear power programme, essential for energy security in the next millennium.

Because of nonavailability of enriched uranium, a bold decision was taken in early eighties to develop plutonium rich-uranium monocarbide fuel with the indigenously available resources for use as driver fuel for FBTR. As this was a unique fuel without any available in-pile irradiation data, the reactor itself served as an irradiation facility for ascertaining the fuel performance. Hence a cautious approach was taken to increase the reactor power in stages after assessing the fuel performance through plant observations and post irradiation examination (PIE) route. A fuel irradiation programme was therefore put in place as early as July 1994, involving irradiation of Mark I and Mark II ( 55% PuC – 45% UC) composition fuel pins to various burnups for PIE. Also the central fuel SA irradiated to 25,000 MWd/t was discharged for examination (Fig-3). This programme was completed in April 1997 and has been a great success. The PIE results have been excellent and morale boosting as regards fuel performance in terms of both enhancement in burnup and linear heat rating. It is now planned to carry out PIE of one of the fuel SA that has reached a burnup of 50,000 MWd/t to assess the case for pushing the fuel burnup further.

The first material irradiation programme was initiated in June 1998 and involved loading of experimental SA having pressurised capsules of Zircalloy and Zr-Nb alloy for measurement of irradiation creep for PHWR programme along with structural material specimens of D9 and 316 LN (candidate materials for PFBR) for evaluating change in mechanical properties of irradiated material. It may be noted that even short time irradiation in FBTR provides dpa levels comparable to several years of irradiation in PHWR. The experimental SA were sequentially discharged and this irradiation programme was completed in April 1999.

Having successfully accomplished all the technology objectives and having ascertained excellent performance of the indigenous fuel, it is now proposed to progressively expand the FBTR core for higher power operation to utilise this facility more effectively. The operational experience feed back has been appropriately incorporated in the design of 500 MWe PFBR.
 
                                                                                                                                (R.P. Kapoor)