HISTORY OF INDIA’S FAST REACTOR PROGRAMME

Nuclear energy is an inevitable source to meet the fast-growing energy demands of India and to provide better quality of life to all the citizens. Its need arises as India has poor coal and oil resources. Presently known reserves of natural uranium in the country can support an installed capacity of about 10 GWe for about 30 years, based on its use in the pressurized heavy water reactors.(See India’s Three Stage Nuclear Programme, Swatantra Mag) Natural uranium contains 0.7% of U235 which is fissionable and 99.3% of U238 which is not fissionable but gets converted to Pu 239 after absorption of neutrons. The used fuel from the heavy water reactors, contain unused U235, U238 and some Pu239. Reprocessing is carried out to get back the unused U235 and Pu 239 for use in the future reactors. It has been demonstrated in USA, Russia, France, and Germany that the conversion ratio of U238 to Pu 239 is much higher in fast reactors, leading to effective utilization of U238 in the natural uranium. As indicated in the earlier article (India’s three stage programme) our Uranium reserves can generate 500 GWe for 100 years (GWe = 106 MWe). Thus, we can effectively utilise the U238 resources by construction of fast reactors that breed fuel. The other material which is available in India in Th232. 55% of world’s thorium reserves are in India. This can be converted into U233, which is again a fissile material that can be used. Since thorium technology is specific to India, the development of fuel fabrication, reprocessing and recovery of U233 is being pursued indigenously. This has a potential to generate another 500 GWe for 350 years.

Dr.Vikram Sarabhai, who took over as Chairman AEC in 1966 took a bold and decisive step to speed up the introduction of the fast reactor technology in India, through effective collaboration with a country having design, construction and operating experience in FBRs. At that time, France, having focused fast reactor programme and an experimental reactor, Rapsodie-Fortissimo, offered to collaborate with India. As the offer was considered attractive, a bilateral agreement was signed in 1969. Under this collaboration agreement, a design team consisting of both French and Indian Engineers was constituted for preparing a preliminary design and a project report for an experimental test reactor (FBTR), to be constructed in India, that could be used as a test bed for materials (i.e. fuel, cladding, etc.) as well as human resource development.

pfbr-construction

The design of FBTR was largely based on RAPSODIE in its primary coolant (sodium) circuit. In RAPSODIE, the nuclear heat was dissipated to atmosphere through sodium to air heat exchanger which was modified in FBTR by adding 4 modules of steam generators (SG) and turbo generator (TG) in the secondary sodium circuit, like 250 MWe PHENIX reactor which was under construction then in France. The idea was to have not only a test reactor for development of advanced fuel and structural materials but also a power demonstration reactor as an integrated step, preceding a fully indigenous effort to design and construct large fast breeders.

As per the agreement with CEA (CEA is the French Atomic Energy Commission, Commissariat à l’énergie atomique),they agreed to give the drawings and specifications of systems and equipment and also provide assistance in design and construction. However the responsibility of the project would rest with India. This was a radical departure from earlier occasions-Tarapur Boiling Water Reactor units 1,2 were turnkey projects by GE, USA and Rajastan Pressurised Heavy Water Reactor Units 1,2 were built with Canada taking the responsibility and Indian members working as part of the construction team. Agreements with France also provided for transfer of technology for manufacture of components in India. In fact all the components were manufactured in ndia with the support of the Indian industries.

FBTR was successfully commissioned in 1985. The criticality and related parameters were predicted very closely for the high plutonium content carbide, even with no experimental support. It may be noted that high Pu Carbide has never been used for a full reactor core anywhere in the world. This fuel has reached burn up of ~ 2,00,000 MWd/te.

FBTR has been a fountainhead of nurturing of competent personnel for the design, construction, and operation of subsequent FBR. Based on the experience in the design and construction of FBTR, DAE decided to launch the design of 500 MWe Prototype Fast Breeder Reactor (PFBR).In order to raise the confidence in the operation of mechanisms in sodium and also to understand the complex phenomena such as heat & mass transfer in the cover gas, a few dedicated test loops were commissioned; the largest one commissioned in 1994, called Large Component Test Rig (LCTR), and is being used effectively for the qualification of prototype components.

From the regulatory side, AERB first developed “Safety Criteria for design of the PFBR” and the guidelines given in this document are used for review of FBR design. Earlier, the conceptual design of PFBR was reviewed by Novatome of France and OKBM of Russia, which are fast reactor design organizations who have considerable experience with Fast Neutron Reactors. Comments from these organizations have been appropriately considered in the PFBR design. Based on preliminary review by AERB, the Kalpakkam site was accepted and clearance for excavation was given in July 2002. The fabrication, construction and erection were done without any foreign collaboration. Safety being the watchword at every step, the commissioning of systems and testing is proceeding gradually. The reactor is expected to go critical and go on power in a step by step manner.

In conclusion, India has made significant inroads into mastering the FBR technology and is in a position to design, build and operate fast reactors.

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