The IPHWR (Indian Pressurized Heavy Water Reactor) is a class of Indian pressurized heavy-water reactors designed by the Bhabha Atomic Research Centre.[1] The baseline 220 MWe design was developed from the CANDU based RAPS-1 and RAPS-2 reactors built at Rawatbhata, Rajasthan. Later the design was based on VVER technology which was scaled to 540 MW and 700 MW designs. Currently there are 17 units of various types operational at various locations in India.
The first PHWR units built in India (RAPS-1 and RAPS-2) are of Canadian CANDU design similar to the first full-scale Canadian reactor built at Douglas point, Ontario. The reactors were set up in collaboration with Government of Canada. Starting in 1963, 100 MWe RAPS-1 was mostly built with equipment and technology supplied by AECL, Canada. RAPS-1 was commissioned in 1973 but the cessation of Canadian cooperation in light of successful development of nuclear weapons by India as part of Operation Smiling Buddha the RAPS-2 commissioning could only be completed by 1981. India took help of Soviet Union whose VVER (Pressurised Water Reactor type) technology was used as a design for indigenization by Bhabha Atomic Research Centre in partnership with Indian manufacturers Larsen & Toubro and Bharat Heavy Electricals Limited. Successively, a totally Indian design of 220 MWe power capacity was designed and two units were built at KalpakkaminTamil Nadu state christened MAPS-1 and MAPS-2. MAPS-1&2 design was evolved from RAPS-1&2, with modifications carried out to suit the coastal location and also introduction of suppression pool to limit containment peak pressure under loss of coolant accident (LOCA) in lieu of dousing tanks in RAPS-1&2. In addition, MAPS-1&2 have partial double containment. This design was further improved and all subsequent PHWR units in India have double containment.[2]
With experience of design and operation of earlier units and indigenous R&D efforts, major modifications were introduced in NAPS-1&2. These units are the basis of standardized Indian PHWR units later designated as IPHWR-220.
The design of subsequent units i.e. KGS-1, KGS-2, RAPS-3, RAPS-4, RAPS-5, RAPS-6, KGS-3 and KGS-4 is of standard Indian PHWR design. The major improvements in these designs include valve-less primary heat transport system and a unitized control room concept. In addition, the design of these units included improvements in Control and Instrumentation system and incorporation of computer based systems to match with the advancement in technology.
Upon completion of the design of IPHWR-220, a larger 540 MWe design was started c. 1984 under the aegis of BARC in partnership with NPCIL.[3] Two reactors of this design were built in Tarapur, Maharashtra starting in the year 2000 and the first was commissioned on 12 September 2005.
The IPHWR-540 design was later upgraded to a 700 MWe with the main objective to improve fuel efficiency and develop a standardized design to be installed at many locations across India as a fleet-mode effort. The design was also upgraded to incorporate Generation III+ features.
The 700 MWe PHWR design includes some features, which are introduced for the first time in Indian PHWRs which include partial boiling at the coolant channel outlet, interleaving of primary heat transport system feeders, passive decay heat removal system, regional over power protection, containment spray system, mobile fuel transfer machine, and a steel liner on the inner containment wall.[4]
By 2031–2032, NPCIL plans to construct 18 more nuclear power reactors, which together have the potential to produce 13,800 MWe of electricity. This will bring the total amount of atomic power in the energy mix to 22,480 MWe.[5]
Specifications
IPHWR-220[2]
IPHWR-700[4]
Thermal output, MWth
754.5
1730
2166
Active power, MWe
220
540
700
Efficiency, net %
27.8
28.08
29.08
Coolant temperature, °C:
core coolant inlet
249
266
266
core coolant outlet
293.4
310
310
Primary coolant material
Heavy Water
Secondary coolant material
Light Water
Moderator material
Heavy Water
Reactor operating pressure, kg/cm2 (g)
87
100
100
Active core height, cm
508.5
594
594
Equivalent core diameter, cm
451
-
638.4
Average fuel power density
9.24 KW/KgU
-
235 MW/m3
Average core power density, MW/m3
10.13
-
12.1
Fuel
Sintered Natural UO2 pellets
Cladding tube material
Zircaloy-2
Zircaloy-4
Fuel assemblies
3672
5096
4704 fuel bundles in 392 channels
Number of fuel rods in assembly
19 elements in 3 rings
37
37 elements in 4 rings
Enrichment of reload fuel
0.7% U-235
Fuel cycle length, Months
24
12
12
Average fuel burnup, MW · day / ton
6700
7500
7050
Control rods
SS/Co
Cadmium/SS
Neutron absorber
Boric Anhydride
Boron
Residual heat removal system
Active: Shutdown cooling system
Passive: Natural circulation through steam generators
Active: Shutdown cooling system
Passive: Natural circulation through steam generators
and Passive Decay heat removal system
Safety injection system
Emergency core cooling system
Uranium resources
Nuclear reactor technologies
Proposed
Nuclear Power Plants
Research and testing
Organisations
Other topics
None
(fast-neutron)
Others
Nuclear technology Nuclear accidents
