Dr. Homi Jehangir Bhabha conceived the Nuclear Program in India. Dr Bhabha established the Tata Institute of Fundamental Research (TIFR) for carrying out nuclear science research in 1945. To intensify the effort to exploit nuclear energy for the benefit of the nation, Dr Bhabha established the Atomic Energy Establishment, Trombay (AEET) in January 1954 for multidisciplinary research program essential for the ambitious nuclear program of India. After the sad demise of Bhabha in 1966, AEET was renamed Bhabha Atomic Research Centre (BARC).
Dr. Bhabha established the BARC Training School to cater to the manpower needs of the expanding atomic energy research and development program. In Bhabha’s own words “When Nuclear Energy has been successfully applied for power production in, say a couple of decades from now, India will not have to look abroad for its experts but will find them ready at hand”. Dr Bhabha emphasized on self reliance in all the fields of nuclear science and engineering.
BARC is the mother of the R&D institutions such as IGCAR, RRCAT, VECC, etc., which carry out pioneering research on nuclear and accelerator technologies and industrial establishments such as NPCIL, NFC, ECIL, etc., spearheading nuclear power production, materials technology, electronics & instrumentation.
Atomic Energy has a key role in reducing the carbon intensity of the overall Power sector of India. Coal based thermal power contributes 186,293 MW (July 2016), 61% of the total installed power while renewables and nuclear contribute 44,237 MW (14.5%) and 5,780 MW (1.9%) respectively (www.powermin.nic.in). While renewable sources of energy are environment friendly, they are intermittent sources of power. Nuclear power, being a non-intermittent and concentrated source of power with negligible carbon footprint, is an essential component of the Indian power-mix to meet the International environmental commitments of India.
India has limited domestic uranium resources while we have abundant Thorium. To exploit Thorium, our planners have envisaged the Three Stage Nuclear Power Program.
Indigenously built Pressurised Heavy Water Reactors (PHWRs) form the backbone of the first stage Indian nuclear power program. PHWRs use domestic natural uranium (UO2) containing 0.7 % fissile U-235 and 99.3% U-238 as fuel and Heavy water as moderator and primary coolant.
Reprocessing of the spent fuel from PHWRs and waste management are important components of the three stage nuclear program. These technologies were developed with total indigenous efforts. Uranium and Plutonium are chemically separated and recycled, while the other radioactive fission products were separated and sorted according to their half lives and radioactivity and stored with minimal environmental impact.
Pu-239 extracted from the spent fuel serves as the fuel for the Fast Breeder Reactors (FBRs) – part of the second stage of the nuclear program. FBR fuel is so designed that a blanket of U-238 surrounds fuel core. U-238 undergoes transmutation to produce fresh Pu-239. Thus an FBR not only consumes Pu-239 but also breeds more Pu-239 than it consumes. But FBR technology is very complex and only advanced countries like USA, UK, France, Japan and USSR have mastered this technology.
India announced its entry into this exclusive club when the 40 MWth Fast Breeder Test Reactor (FBTR) went critical in the Indira Gandhi Centre for Atomic Research, Kalpakkam in October 1985. A unique feature the FBTR is the indigenously developed U-Pu carbide fuel rich in Pu. With the operational experience gained from FBTR, India embarked upon the construction of a 500 MWe Prototype Fast Breeder Reactor (PFBR) which is likely to be operational in 2017.
Th-232, which is abundant in India, is not a fissile material. However, by a neutron capture reaction, Th-232 transforms into U-233, which is a fissile material like U-235 and Pu-239. The strategy of the three stage program is to convert Th-232 into U-233 in the fast reactors. U-233 will be the fuel in the futuristic third stage of nuclear program. Further, it is proposed to use thorium along with a small feed of plutonium-based fuel in Advanced Heavy Water Reactors (AHWRs) which are expected to facilitate large-scale thorium utilization.
BARC has active groups for Research and Development in Reactor Technologies, Fuel reprocessing and waste management, Isotope Applications, Radiation Technologies and their application to health, agriculture and environment, Accelerator and Laser Technology, Electronics, instrumentation and reactor control and Materials Science. Strong emphasis on basic and applied research in a number of core disciplines of Science has made synergy between basic research and technology development possible.
APSARA is the first nuclear reactor in Asia to have achieved criticality on August 4, 1956. It is a pool type reactor of 1 MW power with a highly enriched uranium as fuel (4.5 kg) in the form of plates. Light water was used as both moderator and coolant. The maximum neutron flux was around 1013 neutrons/cm2/s. It was mainly used for production of isotopes, basic research, shielding experiments, neutron activation analysis, neutron radiography and for testing neutron detectors. Apsara was permanently shutdown in 2010. An upgraded version of 2 MWth with a light enriched uranium fuel is under construction in BARC.
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ZERLINA was a 100 Watt thermal reactor based on natural uranium metallic fuel and heavy water as both moderator and coolant. It achieved criticality in January 14, 1961. It was mainly utilized for reactor lattice studies and was decommissioned in 1983.
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CIRUS reactor (40 MWth) was built in collaboration with Canada and was commissioned on 10th July 1960. CIRUS was extensively used for condensed matter research using neutron beams, material irradiation, fuel testing, neutron activation analysis and production of radioisotopes for applications in medicine, agriculture and industry. CIRUS reactor proved to be an excellent platform for training of engineers and scientists and in understanding the intricacies of managing natural uranium, heavy water, reactor systems which eventually evolved into the Indian pressurized heavy water reactor programme. After 50 years of successful operation, it was permanently shut-down in December 2010.
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DHRUVA was conceived in the 1970s out of the need for a research reactor with higher neutron flux for research in basic sciences in addition to a growing demand of radioisotopes and advanced. Construction of DHRUVA was a major milestone in the development and implementation of indigenous nuclear technology in India. The reactor incorporates several features not only to satisfy multidisciplinary user community but also in the production of radioisotopes of high specific activity. Dhruva is a National Facility for Neutron Beam Research to cater to the needs of Indian scientific community where scientists from BARC, other units of the Department of Atomic Energy (DAE), universities and national laboratories work under collaborative projects. Many of the collaborations are supported by the University Grants Commission – DAE Consortium for Scientific Research (UGC-DAE-CSR) and the Board of Research in Nuclear Sciences (BRNS).
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PURNIMA-I was the first experimental fast reactor built in BARC with Plutonium Oxide as the fuel. This 1 Watt reactor was critical on 18th May 1972 and was used for studies of fast reactor physics. It was decommissioned in 1973.
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PURNIMA-II was 100 mW experimental thermal reactor in BARC achieved criticality on May 10, 1984. This reactor operated on U-233 in the form of Uranyl Nitrate Solution as fuel and light water as both moderator and coolant. PURNIMA-II had maximum neutron flux of 107 neutrons/cm2/s. The purpose of this reactor was U-233 fuel evaluation and futuristic reactor studies and was decommissioned in 1986.
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PURNIMA-III was another U-233 based 1 W thermal reactor built in BARC intended for mockup studies for KAMINI reactor. It achieved criticality on November 9, 1990. Like its predecessor, this reactor too operated on light water as both moderator and coolant. PURNIMA-III had maximum neutron flux of 108 neutrons/cm2/s. This reactor was decommissioned in 1991.
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KAMINIÂ is a 30 kW reactor in Indira Gandhi Centre for Atomic Research, Kalpakkam, achieved criticality in October 1996. It marks a small but significant step towards utilisation of our thorium reserves. It is the only operating reactor in the world using U-233 fuel. It is used for neutron radiography in IGCAR.