Prototype Fast Breeder Reactor

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PFBR
GenerationGeneration III+ reactor
Reactor conceptPlutonium Fast breeder reactor
Reactor lineIFBR (Indian fast-breeder Reactor)
Designed byIGCAR
Manufactured byBHAVINI
StatusUnder development
Main parameters of the reactor core
Fuel (fissile material)235U/239Pu (NEU/239Pu/MOX)
Fuel stateSolid
Neutron energy spectrumFast
Primary control methodcontrol rods
Primary coolantLiquid Sodium
Reactor usage
Primary useBreeding of 233U for AHWR-300 and Generation of electricity
Power (thermal)1253
Power (electric)500
Prototype Fast Breeder Reactor
CountryIndia
LocationMadras
Coordinates12°33′11″N 80°10′24″E / 12.55306°N 80.17333°E / 12.55306; 80.17333Coordinates: 12°33′11″N 80°10′24″E / 12.55306°N 80.17333°E / 12.55306; 80.17333
StatusUnder construction
Construction began2004
Commission dateOctober 2022 (planned)[1]
Construction cost5,850 crore (equivalent to 170 billion or US$2.43 billion in 2019)[1]
Owner(s)BHAVINI
Operator(s)BHAVINI
Nuclear power station
Reactor typefast breeder
Cooling source
Power generation
Nameplate capacity500 MW

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe fast breeder nuclear reactor presently being constructed at the Madras Atomic Power Station in Kalpakkam, India.[2] The Indira Gandhi Centre for Atomic Research (IGCAR) is responsible for the design of this reactor. The facility builds on the decades of experience gained from operating the lower power Fast Breeder Test Reactor (FBTR). Originally planned to be commissioned in 2012, the construction of the reactor suffered from multiple delays. As of August 2020, criticality is planned to be achieved in 2021.[3]

History[edit]

In 2007 the reactor was planned to begin operating in 2010, but as of 2019 it is expected to reach first criticality in 2020.[3] The Kalpakkam PFBR is designed to use uranium-238 to breed plutonium in a sodium-cooled fast reactor design. The power island of this project was engineered by Bharat Heavy Electricals Limited, largest power equipment utility of India.[citation needed]

The surplus plutonium (or uranium-233 for thorium reactors) from each fast reactor can be used to set up more such reactors and grow the nuclear capacity in tune with India's needs for power. The PFBR is part of the three-stage nuclear power program.

India has the capability to use thorium cycle based processes to extract nuclear fuel. This is of special significance to the Indian nuclear power generation strategy as India has one of the world's largest reserves of thorium, which could provide power for more than 10,000 years,[4] and perhaps as long as 60,000 years.[5][6]

The design of this reactor was started in the 1980s, as a prototype for a 600 MW FBR. Construction of the first two FBR are planned at Kalpakkam, after a year of successful operation of the PFBR. Other four FBR are planned to follow beyond 2030, at sites to be defined.[7]

In July 2017, it was reported that the reactor is in final preparation to go critical.[8] However in August 2020, it was reported that the reactor might go critical only in December 2021.[9]

As of February 2021, around 5,850 crore (equivalent to 170 billion or US$2.43 billion in 2019) have been spent in the construction and commissioning of the reactor. The reactor is now expected to be operational by October 2022.[1]

Technical details[edit]

Schematic diagram showing the difference between the Loop and Pool designs of a liquid metal fast breeder reactor. The pool-type has greater thermal inertia to changes in temperature, which therefore gives more time to shut down/SCRAM during a loss of coolant accident situation.

The reactor is a pool type LMFBR with 1,750 tonnes of sodium as coolant. Designed to generate 500 MWe of electrical power, with an operational life of 40 years, it will burn a mixed uranium-plutonium MOX fuel, a mixture of PuO
2
and UO
2
. A fuel burnup of 100 GWd/t is expected. The Advanced Fuel Fabrication Facility (AFFF), under the direction of BARC, Tarapur, is responsible for the fuel rods manufacturing. AFFF comes under " Nuclear Recycle Board" of Bhabha Atomic Research Center. AFFF has been responsible for fuel rod manufacturing of various types in the past.

Safety considerations[edit]

The prototype fast breeder reactor has a negative void coefficient, thus ensuring a high level of passive nuclear safety. This means that when the reactor overheats (below the boiling point of sodium) the speed of the fission chain reaction decreases, lowering the power level and the temperature.[10] Similarly, before such a potential positive void condition may form from a complete loss of coolant accident, sufficient coolant flow rates are made possible by the use of conventional pump inertia, alongside multiple inlet-perforations, to prevent the possible accident scenario of a single blockage halting coolant flow.[11] The active-safety reactor decay heat removal system consists of four independent coolant circuits of 8MWt capacity each.[12] Further active defenses against the positive feedback possibility include two independent SCRAM shutdown systems, designed to shut the fission reactions down effectively within a second, with the remaining decay heat then needing to be cooled for a number of hours by the 4 independent circuits.

The fact that the PFBR is cooled by liquid sodium creates additional safety requirements to isolate the coolant from the environment, especially in a loss of coolant accident scenario, since sodium explodes if it comes into contact with water and burns when in contact with air. This latter event occurred in the Monju reactor in Japan in 1995. Another consideration with the use of sodium as a coolant is the absorption of neutrons to generate the radioactive isotope 24
Na
, which has a 15-hour half life.[13]

Technical Specifications[edit]

Specifications PFBR[14][15][16][17]
Thermal output, MW 1253
Active power, MW 500
Efficiency, net % 39.9
Coolant temperature, °C:
     core coolant inlet 397
     core coolant outlet 547
Primary coolant material Liquid Sodium
Active core height, cm 111
Equivalent core diameter, mm 1900
Average fuel power density, MW/m3 416
Average core power density, MW/m3 -
Fuel two enrichment zones of 20.7 and 27.7 wt% of PuO2 in the mixture of PuO2 and UO2
Cladding tube material 20% CW D9
Fuel assemblies 85 of 20.7 % PuO2

96 of 27.7% PuO2,

Number of pins in assembly 217
Enrichment of reload fuel
Fuel cycle length, Effective Full Power Days (EFPD) 250
Average fuel burnup, GW · day / t 134
Control rods B4C Boron Carbide
Neutron absorber B4C Boron Carbide

See also[edit]

References[edit]

  1. ^ a b c "Lok Sabha Unstarred Question No. 330, Budget Session 2021" (PDF). Department of Atomic Energy, Government of India. 3 February 2021. Retrieved 18 April 2021.
  2. ^ Baldev Raj, S.C. Chetal and P. Chellapandi (8 January 2010). "Great expectations". Nuclear Engineering International.
  3. ^ a b "Indian government takes steps to get nuclear back on track". world nuclear news. World Nuclear Association. 11 February 2019.
  4. ^ Chris Rhodes (26 February 2012). "Thorium can power civilization for over 3000 years". Retrieved 23 March 2012.
  5. ^ MacKay, David J. C. (20 February 2009). Sustainable Energy - Without the Hot Air. UIT Cambridge Ltd. p. 166. Retrieved 23 March 2012.
  6. ^ Rodricks, Dan (9 May 2011). "Thor's nuclear-powered hammer". The Baltimore Sun. Retrieved 23 March 2012.
  7. ^ "India plans to construct six more fast breeder reactors". The Economic Times. 1 December 2015. Retrieved 15 December 2015.
  8. ^ "Nuclear reactor at Kalpakkam: World's envy, India's pride". The Times of India. 26 November 2017. Retrieved 2 July 2017.
  9. ^ "India's First Prototype Fast Breeder Reactor Has a New Deadline. Should We Trust It?".
  10. ^ Raj, Baldev (30 October 2009). "Design Robustness and Safety Adequacy of India's Fast Breeder Reactor". Science & Global Security. 17 (2–3): 194–196. doi:10.1080/08929880903451397.
  11. ^ Raj, Baldev (30 October 2009). "Design Robustness and Safety Adequacy of India's Fast Breeder Reactor". Science & Global Security. 17 (2–3): 194–196. doi:10.1080/08929880903451397.
  12. ^ "Design of 500 MWe Prototype Fast Breeder Reactor" (PDF). Archived from the original (PDF) on 17 April 2012. Retrieved 17 April 2012.
  13. ^ "BARC Highlights,Reactor Technology & Engineering" (PDF). Bhabha Atomic Research Centre. Retrieved 21 March 2021.
  14. ^ Choudhry, Nakul; Riyas, A (February 2013). "3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup". Nuclear Engineering and Design. Volume 255: 359–367. doi:10.1016/j.nucengdes.2012.11.011. |volume= has extra text (help)
  15. ^ Devan, K (2008). "A new physics design of control safety rods for prototype fast breeder reactor". Annals of Nuclear Energy. 35: 1484–1491. doi:10.1016/j.anucene.2008.01.013.
  16. ^ Lee, S.M; Govindarajan, S; Indira, R (1996). "Conceptual design of PFBR core" (PDF). IAEA-TECDOC. International Atomic Energy Agency. 907.

External links[edit]