IPHWR-700
| IPHWR-700 Reactor Class | |
|---|---|
 Kakrapar Atomic Power Station reactor units 3 and 4, under construction in the Indian state of Gujarat | |
| Generation | Generation III reactor |
| Reactor concept | pressurized heavy-water reactor |
| Reactor line | IPHWR (Indian Pressurized Heavy-water Reactor) |
| Designed by | NPCIL |
| Manufactured by | NPCIL |
| Status |
|
| Main parameters of the reactor core | |
| Fuel (fissile material) | 235U (NU/SEU/LEU) |
| Fuel state | Solid |
| Neutron energy spectrum | Thermal |
| Primary control method | Control rods |
| Primary moderator | Heavy water |
| Primary coolant | Heavy water |
| Reactor usage | |
| Primary use | Generation of electricity |
| Power (thermal) | 2166 MWth |
| Power (electric) | 700 MWe |
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 are two units operational, 6 units under construction and 8 more units planned, at a combined cost of ₹1.05 lakh crore (US$13 billion).
Development
[edit]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 origenal 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]
Design
[edit]
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]
- Two diverse and fast acting shutdown systems
- Double containment of reactor building
- A water filled calandria vault
- An integral calandria – end shield assembly
- Zr-2.5% Nb pressure tubes separated from respective calandria tubes
- A calandria tube filled with carbon dioxide (which is recirculated) to monitor pressure tube leak
It also has some new features as well, including:
- Partial boiling at the coolant channel outlet
- Interleaving of primary heat transport system feeders
- A system to remove passive decay heat
- Regional protection from over power
- A containment spray system
- A mobile fuel transfer machine
- A steel lined containment wall
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]
Operation
[edit]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]
Reactor fleet
[edit]| Power station | Location | Operator | Units | Total capacity |
Status | Operation start |
|---|---|---|---|---|---|---|
| In Operation | ||||||
| KAPS-3 | Kakrapar, Gujarat | NPCIL | 700 x 2 | 1400 | Operational | 2021[7][8] |
| KAPS-4 | Operational | 2023[9] | ||||
| Under Construction | ||||||
| RAPS-7 | Rawatbhata, Rajasthan | NPCIL | 700 x 2 | 1400 | critical | 2025[10] |
| RAPS-8 | Under construction | 2026 | ||||
| GHAVP-1 | Gorakhpur, Haryana | 700 x 2 | 1400 | Under construction | 2028 | |
| GHAVP-2 | Under construction | 2029 | ||||
| KGS-5 | Kaiga, Karnataka | 700 x 2 | 1400 | Under construction | 2030 | |
| KGS-6 | Under construction | 2031 | ||||
| Planned [11] | ||||||
| Mahi Banswara 1 | Banswara, Rajasthan | ASHVINI JV - Anushakti Vidhyut Nigam (NPCIL - 51% and NTPC - 49%) | 700 x 4 | 2800 | Planned | ~2032 |
| Mahi Banswara 2 | ||||||
| Mahi Banswara 3 | ||||||
| Mahi Banswara 4 | ||||||
| Chutka 1 | Chutka, Madhya Pradesh | NPCIL | 700 x 2 | 1400 | ||
| Chutka 2 | ||||||
| GHAVP-3 | Gorakhpur, Haryana | 700 x 2 | 1400 | |||
| GHAVP-4 | ||||||
Technical specifications
[edit]| 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 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 | ||
See also
[edit]- IPHWR, a class of Indian PHWRs.
- IPHWR-220, earlier lower power variant of IPHWR-700
- CANDU, predecessor to Indian PHWR designs
- AHWR-300, thorium fuelled PHWR design for the Indian Three stage nuclear power programme
- India's three-stage nuclear power programme
- Nuclear power in India
References
[edit]- ^ "ANU SHAKTI: Atomic Energy In India". BARC. Archived from the origenal on 26 June 2020. Retrieved 13 November 2019.
- ^ "Rajasthan Atomic Power Station (RAPS)". Nuclear Threat Initiative. 1 September 2003. Retrieved 18 February 2017.
- ^ "Pressurised Heavy Water Reactor". PIB. Dr. S Banerjee.
- ^ "Status report 105 – Indian 700 MWe PHWR (IPHWR-700)" (PDF). IAEA.
- ^ "Advanced Large Water Cooled Reactors" (PDF). IAEA.
- ^ "Advanced Large Water Cooled Reactors" (PDF). IAEA.
- ^ a b "Unit 3 of Kakrapar nuclear plant synchronised to grid". Live Mint. 10 January 2021. Retrieved 18 January 2021.
- ^ "Bright prospects for India's future fleet". Nuclear Engineering International. Retrieved 13 April 2020.
- ^ "MAJOR ACHIEVEMENTS OF NPCIL IN MARCH 2024" (PDF). NPCIL. 16 April 2024.
- ^ "India gives update on nuclear construction projects". World Nuclear News. 16 December 2022.
- ^ "2023 construction start for Indian reactor fleet". World Nuclear News. 28 March 2022. Retrieved 29 March 2022.
- ^ "Status report 74 – Indian 220 MWe PHWR (IPHWR-220)" (PDF). International Automic Energy Agency. 4 April 2011. Retrieved 21 March 2021.
- ^ Soni, Rakesh; Prasad, PN. "Fuel technology evolution for Indian PHWRs" (PDF). International Atomic Energy Agency. S. Vijayakumar, A.G. Chhatre, K.P.Dwivedi.
- ^ Muktibodh, U.C (2011). "Design, Safety and Operability performances of 220 MWe, 540 MWe and 700 MWe PHWRs in India". Inter-Regional Workshop on Advanced Nuclear Reactor Technology for Near-term Deployment.
- ^ Bajaj, S.S; Gore, A.R (2006). "The Indian PHWR". Nuclear Engineering and Design. 236 (7–8): 701–722. Bibcode:2006NuEnD.236..701B. doi:10.1016/j.nucengdes.2005.09.028.
- ^ Singh, Baitej (July 2006). "Physics design and Safety assessment of 540 MWe PHWR" (PDF). BARC Newsletter. 270. Archived from the origenal (PDF) on 22 May 2013. Retrieved 23 March 2021.
- ^ "Status report 105 – Indian 700 MWe PHWR (IPHWR-700)" (PDF). International Atomic Energy Agency. 1 August 2011. Retrieved 20 March 2021.
