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.
See main article: IPHWR-220.
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 Kalpakkam in Tamil 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.
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 under the aegis of BARC in partnership with NPCIL. Two reactors of this design were built in Tarapur, Maharashtra starting in the year 2000 and the first was commissioned on 12 September 2005.
See main article: IPHWR-700.
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.
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.[2]
| 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 systemPassive: Natural circulation through steam generators | Active: Shutdown cooling systemPassive: Natural circulation through steam generators and Passive Decay heat removal system | ||
| Safety injection system | Emergency core cooling system | |||