Research Reactors in BARC:Bhabha Atomic Research Centre(BARC), Department of Atomic Energy,Government of India

No.	Item	Description
1	Reactor Type	Swimming pool type/ Thermal Reactor
2	Date of Criticality	September 10, 2018
3	Reactor Power (Th)	2 MW (Maximum)
4	Fuel Material	Uranium Silicide (LEU)
5	Fuel Element	Plates
6	Fuel Cladding	Aluminium alloy
7	Core Size	662 mm x 662 mm x 140 mm; 64 lattice positions
8	Max Neutron Flux	6.1X10¹³n/cm²/sec
9	Moderator	Light water
10	Coolant	Light water
11	Reflector	Beryllium Oxide
12	Shut off rods / Control Rods	Hafnium
13	Uses	Basic research Production of isotopes Neutron activation analysis Neutron radiography Testing of neutron detectors Shielding experiments
Apsara-U (Apsara- Upgraded) was successfully commissioned and its First Approach to Criticality (FAC) was achieved on 10th September 2018. Indigenously developed Low Enriched Uranium (LEU) fuel in Uranium Silicide form is used in the reactor. Hot water layer concept at the top of pool, which is first of a kind in India, is employed to minimize radiation dose. By virtue of higher neutron flux, Apsara-U will enhance indigenous production of radioisotopes for various societal applications. The reactor will also be used extensively for research in nuclear physics, material science and radiation shielding.

No.	Item	Description
1	Reactor Type	Swimming pool type/ Thermal Reactor
2	Date of Criticality	August 4,1956
3	Reactor Power (Th)	1 MW (Maximum)
4	Fuel Material	Enriched uranium -aluminium alloy
5	Fuel Element	Plates
6	Fuel Cladding	Aluminium alloy
7	Total weight of Fuel	4.5 kg
8	Core Size	560 mm X 560 mm X 615 mm(H)
9	Max Neutron Flux	10¹³n/cm²/sec
10	Moderator	Light water
11	Coolant	Light water
12	Shut off Rods	Cadmium
13	Uses	Production of isotopes Basic research Shielding experiments Neutron activation analysis Neutron radiography Testing of neutron detectors
The Apsara Reactor was permanently shutdown in 2010.

No.	Item	Description
1	Reactor Type	Vertical Tank Type / Thermal
2	Date of Criticality	January 14,1961
3	Reactor Power (Th)	100 W (Nominal)
4	Fuel Material	Natural Uranium Metal
5	Fuel Element	Rods
6	Fuel Cladding	Aluminium
7	Total weight of Fuel	Variable (3.7 T in the first core)
8	Core Size	2.29m (D) X 4.35m(H)
9	Max Neutron Flux	0.5x10⁸n/cm²/sec
10	Moderator	Heavy water
11	Coolant	Heavy water
12	Shut off Rods	Cadmium
13	Uses	Reactor Lattice studies
14	Year of decommissioning	1983

The CIRUS reactor was commissioned on 10th July 1960. It was built in collaboration with Canada under the leadership of Dr. Homi Jahangir Bhabha. This reactor was extensively used for condensed matter research, using neutron beams extracted from its core. The reactor was also extensively used for material irradiation, fuel testing, neutron activation analysis and production of radioisotopes for application in the fields of 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 decided to permanently shut-down the reactor in December 2010.

No.	Item	Description
1	Reactor Type	Vertical Tank Type / Thermal Reactor
2	Date of Criticality	July 10,1960
3	Reactor Power (Th)	40 MW (Maximum)
4	Fuel Material	Natural Uranium Metal
5	Fuel Element	Rods
6	Fuel Cladding	Aluminium
7	Total weight of Fuel	10.5 T
8	Core Size	2.67m (D) X 3.14m (H)
9	Max Neutron Flux	6.5x10¹³n/cm²/sec
10	Moderator	Heavy water
11	Coolant	Light water
12	Shut off Rods	Boron Cadmium
13	Uses	Production of isotopes R & D in reactor technology Manpower training Neutron beam research Neutron activation analysis Development and testing of fuel assemblies Testing of neutron detectors
The Reactor was permanently shutdown on 31.12.2010.

During early 1970s a strong need was felt for building a research reactor with higher neutron flux to meet the growing demand of radioisotopes and advanced research in basic sciences. This led to the setting up of a research reactor at BARC which was named Dhruva by Dr. Giani Zail Singh, the then President of India. Construction of DHRUVA was an important milestone in the development and implementation of indigenous nuclear technology in India. The reactor incorporates several features catering to the requirements of a broad-based multidisciplinary user community as also in the production of radioisotopes of high specific activity. Dhruva has been declared as 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), the Board of Research in Nuclear Sciences (BRNS) and other agencies. At present there are about 40 active projects running under the UGC-DAE-CSR scheme.

No.	Item	Description
1	Reactor Type	Vertical Tank Type / Thermal Reactor
2	Date of Criticality	August 8,1985
3	Reactor Power (Th)	100 MW (Maximum)
4	Fuel Material	Natural Uranium Metal
5	Fuel Element	Clusters
6	Fuel Cladding	Aluminium
7	Total weight of Fuel	6.35 T
8	Core Size	3.72m(D) x 3.87m (H)
9	Max Neutron Flux	1.8x10¹⁴n/cm²/sec
10	Moderator	Heavy water
11	Coolant	Heavy water
12	Shut off Rods	Cadmium
13	Uses	Basic research Production of isotopes Manpower training Neutron activation analysis Testing of neutron detectors

No.	Item	Description
1	Reactor Type	Fast Reactor
2	Date of Criticality	May 18 , 1972
3	Reactor Power (Th)	1 W (Nominal)
4	Fuel Material	Plutonium oxide
5	Fuel Element	Plates
6	Fuel Cladding	Stainless Steel
7	Total weight of Fuel	21.6 kg
8	Core Size	180 mm(D) X 180 mm(H)
9	Max Neutron Flux	~10⁸n/cm²/sec
10	Moderator	None
11	Coolant	Air
12	Shut off Rods	Molybdenum
13	Uses	Fast reactor physics studies
14	Year of decommissioning	1973

No.	Item	Description
1	Reactor Type	Vertical Tank Type / Thermal Reactor
2	Date of Criticality	May 10,1984
3	Reactor Power (Th)	100 mW (Nominal)
4	Fuel Material	Uranium-233
5	Fuel Element	Uranyl Nitrate Solution
6	Fuel Cladding	None
7	Total weight of Fuel	400 g
8	Core Size	148 mm(D) X (250-400) cm(H)
9	Max Neutron Flux	10⁷n/cm²/sec
10	Moderator	Light water
11	Coolant	Light water
12	Control Rods	Boron carbide
13	Shut off Devices	Cadmium
14	Uses	U-233 fuel studies; futuristic reactor evaluation
15	Year of decommissioning	1986

No.	Item	Description
1	Reactor Type	Vertical Tank Type / Thermal Reactor
2	Date of Criticality	November 9,1990
3	Reactor Power (Th)	1 W (Maximum)
4	Fuel Material	Uranium-233-aluminium alloy
5	Fuel Element	Plates
6	Fuel Cladding	Aluminium
7	Total weight of Fuel	600 g
8	Core Size	204 mm X 204 mm X 275 mm (H)
9	Max Neutron Flux	~10⁸n/cm²/sec
10	Moderator	Light water
11	Coolant	Light water
12	Control Elements	Cadmium plates
13	Reflector material	Berillium oxide
14	Uses	Mockup studies for Kamini reactor
15	Year of decommissioning	1991

No.	Item	Description
1	Reactor Type	Vertical Tank Type/Thermal Reactor
2	Date of Criticality	April 7, 2008
3	Nominal Reactor Fission Power	100 watts
4	Fuel Material	Natural Uranium Metal
5	Fuel Element	Clusters
6	Fuel Cladding	Aluminium
7	Total Weight of Fuel	5.0 T
8	Core Size	3.3 m (D) x 5.0m (H)
9	Neutron Flux (Average)	~10⁸n/cm²/sec
10	Moderator	Heavy Water
11	Shut Off Rods	Cadmium
12	Uses	Validation of the reactor physics design of AHWR Neutron activation analysis Nuclear detector testing Assessment of negative reactivity of absorber material
13	Provision for short duration operation of the reactor at higher power to obtain higher neutron flux of around 10⁹ n/cm²/sec for the development & calibration of self powered neutron detectors.