| IPHWR-700 Reactor Class | |
| Concept: | pressurized heavy-water reactor |
| Generation: | Generation III reactor |
| Reactor Line: | IPHWR |
| Design: | NPCIL |
| Maker: | NPCIL |
| Status: |
|
| Fuel Type: | NU_SEU_LEU |
| Fuel State: | Solid |
| Spectrum: | THERMAL |
| Control: | Control rods |
| Coolant: | Heavy water |
| Moderator: | Heavy water |
| Electric: | 700 MWe |
| Thermal: | 2166 MWth |
| Use: | Generation of electricity |
The IPHWR-700 (Indian Pressurized Heavy Water Reactor-700) is an Indian pressurized heavy-water reactor designed by the NPCIL.[1] It is a Generation III reactor developed from earlier CANDU based 220 MW and 540 MW designs. It can generate 700 MW of electricity. Currently there is two unit operational, 6 units under construction and 8 more units planned, at a cost of .
PHWR technology was introduced in India in the late 1960s with the construction of RAPS-1, a CANDU reactor in Rajasthan. All the main components for the first unit were supplied by Canada. India did the construction, installation and commissioning. In 1974, after India conducted Smiling Buddha, its first nuclear weapons test, Canada stopped their support of the project. This delayed the commissioning of RAPS-2 until 1981.[2]
After Canada withdrew from the project, research, design and development work in the Bhabha Atomic Research Centre and Nuclear Power Corporation of India (NPCIL) enabled India to proceed without assistance. India took help of Soviet Union whose VVER(Pressurised Water Reactor type) technology was used as a design for indigenization. Some industry partners did manufacturing and construction work. Over four decades, fifteen 220-MW reactors of indigenous design were built. Improvements were made in the original VVER design to reduce construction time and cost. New safety systems were incorporated. Reliability was enhanced, bringing better capacity factors and lower costs.
To get economies of scale, NPCIL developed a 540 MW design. Two of these were constructed at the Tarapur Atomic Power Station.
After a redesign to utilise excess thermal margins, the 540 MW PHWR design achieved a 700 MW capacity without many design changes. Almost 100% of the parts of these indigenously designed reactors are manufactured by Indian industry.[3]
Like other pressurized heavy-water reactors, IPHWR-700 uses heavy water (deuterium oxide, D2O) as its coolant and neutron moderator. The design retains the features of other standardized Indian PHWR units, which include:[4]
It also has some new features as well, including:
The reactor has less excess reactivity. Therefore, it does not need neutron poison inside the fuel or moderator. These designs handle the case of a loss of coolant accident such as occurred in the Fukushima Daiichi nuclear disaster.[5]
The reactor fuel uses natural uranium fuel with Zircaloy-4 cladding. The core produces 2166 MW of heat which is converted into 700 MW of electricity at a thermal efficiency of 32%. Because there is less excess reactivity inside the reactor, it needs to be refuelled continually during operation. The reactor is designed for an estimated life of 40 years.[6]
Unit 3 of Kakrapar Atomic Power Station was connected to the grid on 10 January 2021.[7]
| In Operation | ||||||
|---|---|---|---|---|---|---|
| KAPS-3 | Kakrapar, Gujarat | 700 x 2 | 700 | 2021[8] | ||
| 2023[9] | ||||||
| Under Construction | ||||||
| RAPS-7 | Rawatbhata, Rajasthan | NPCIL | 700 x 2 | 1400 | 2026[10] | |
| RAPS-8 | ||||||
| Gorakhpur, Haryana | 700 x 2 | 1400 | 2032 | |||
| GHAVP-2 | ||||||
| KGS-5 | Kaiga, Karnataka | 700 x 2 | 1400 | |||
| KGS-6 | ||||||
| Planned [11] | ||||||
| Banswara, Rajasthan | NPCIL | 700 x 4 | 2800 | rowspan="8" | ||
| Chutka 1 | Chutka, Madhya Pradesh | 700 x 2 | 1400 | rowspan="2" | ||
| Chutka 2 | ||||||
| Gorakhpur, Haryana | 700 x 2 | 1400 | rowspan="2" | |||
| GHAVP-4 | ||||||
| Specifications | IPHWR-220[12] | IPHWR-540[13] [14] [15] [16] | IPHWR-700[17] | |
|---|---|---|---|---|
| Thermal output, MWth | 754.5 | 1730 | 2166 | |
| Active power, MWe | 220 | 540 | 700 | |
| Efficiency, net % | 27.8 | 28.08 | 29.00 | |
| 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 | |||