[{"@context":"http:\/\/schema.org\/","@type":"BlogPosting","@id":"https:\/\/wiki.edu.vn\/en\/wiki21\/iphwr-700-wikipedia\/#BlogPosting","mainEntityOfPage":"https:\/\/wiki.edu.vn\/en\/wiki21\/iphwr-700-wikipedia\/","headline":"IPHWR-700 – Wikipedia","name":"IPHWR-700 – Wikipedia","description":"before-content-x4 From Wikipedia, the free encyclopedia after-content-x4 Indian nuclear reactor design The IPHWR-700 (Indian Pressurized Heavy Water Reactor-700) is an","datePublished":"2019-04-06","dateModified":"2019-04-06","author":{"@type":"Person","@id":"https:\/\/wiki.edu.vn\/en\/wiki21\/author\/lordneo\/#Person","name":"lordneo","url":"https:\/\/wiki.edu.vn\/en\/wiki21\/author\/lordneo\/","image":{"@type":"ImageObject","@id":"https:\/\/secure.gravatar.com\/avatar\/c9645c498c9701c88b89b8537773dd7c?s=96&d=mm&r=g","url":"https:\/\/secure.gravatar.com\/avatar\/c9645c498c9701c88b89b8537773dd7c?s=96&d=mm&r=g","height":96,"width":96}},"publisher":{"@type":"Organization","name":"Enzyklop\u00e4die","logo":{"@type":"ImageObject","@id":"https:\/\/wiki.edu.vn\/wiki4\/wp-content\/uploads\/2023\/08\/download.jpg","url":"https:\/\/wiki.edu.vn\/wiki4\/wp-content\/uploads\/2023\/08\/download.jpg","width":600,"height":60}},"image":{"@type":"ImageObject","@id":"https:\/\/upload.wikimedia.org\/wikipedia\/commons\/thumb\/6\/67\/I-PHWR700_Model.png\/220px-I-PHWR700_Model.png","url":"https:\/\/upload.wikimedia.org\/wikipedia\/commons\/thumb\/6\/67\/I-PHWR700_Model.png\/220px-I-PHWR700_Model.png","height":"158","width":"220"},"url":"https:\/\/wiki.edu.vn\/en\/wiki21\/iphwr-700-wikipedia\/","wordCount":3684,"articleBody":"     (adsbygoogle = window.adsbygoogle || []).push({});before-content-x4From Wikipedia, the free encyclopedia       (adsbygoogle = window.adsbygoogle || []).push({});after-content-x4Indian nuclear reactor designThe IPHWR-700 (Indian Pressurized Heavy Water Reactor-700) is an Indian pressurized heavy-water reactor designed by the Bhabha Atomic Research Centre.[1] It is a Generation III reactor developed from earlier CANDU based 220\u00a0MW and 540\u00a0MW designs. It can generate 700\u00a0MW of electricity. Currently there are 6 units under construction and 10 more units planned, at a cost of \u20b91.05 lakh crore (US$13\u00a0billion).       (adsbygoogle = window.adsbygoogle || []).push({});after-content-x4Table of ContentsDevelopment[edit]Operation[edit]Reactor fleet[edit]Technical specifications[edit]See also[edit]References[edit]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. 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 CANDU design to reduce construction time and cost. New safety systems were incorporated. Reliability was enhanced, bringing better capacity factors and lower costs.       (adsbygoogle = window.adsbygoogle || []).push({});after-content-x4To 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]  I-PHWR700 Model  installed in GCNEP Office, HaryanaLike 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 systemsDouble containment of reactor buildingA water filled calandria vaultAn integral calandria \u2013 end shield assemblyZr-2.5% Nb pressure tubes separated from respective calandria tubesA calandria tube filled with carbon dioxide (which is recirculated) to monitor pressure tube leakIt also has some new features as well, including:Partial boiling at the coolant channel outletInterleaving of primary heat transport system feedersA system to remove passive decay heatRegional protection from over powerA containment spray systemA mobile fuel transfer machineA steel lined containment wallThe 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]Technical specifications[edit]SpecificationsIPHWR-220[12]IPHWR-540[13][14][15][16]IPHWR-700[17]Thermal output, MWth754.517302166Active power, MWe220540700Efficiency, net\u00a0%27.828.0832.00Coolant temperature, \u00b0C:\u00a0 \u00a0 \u00a0core coolant inlet249266266\u00a0 \u00a0 \u00a0core coolant outlet293.4310310Primary coolant materialHeavy WaterSecondary coolant materialLight WaterModerator materialHeavy WaterReactor operating pressure, kg\/cm2 (g)87100100Active core height, cm508.5594594Equivalent core diameter, cm451\u2013638.4Average fuel power density9.24 KW\/KgU235 MW\/m3Average core power density, MW\/m310.1312.1FuelSintered Natural UO2 pelletsCladding tube materialZircaloy-2Zircaloy-4Fuel assemblies367250964704 fuel bundles in 392 channelsNumber of fuel rods in assembly19 elements in 3 rings3737 elements in 4 ringsEnrichment of reload fuel0.7% U-235Fuel cycle length, Months241212Average fuel burnup, MW \u00b7 day \/ ton670075007050Control rodsSS\/CoCadmium\/SSNeutron absorberBoric AnhydrideBoronResidual heat removal systemActive: Shutdown cooling systemPassive: Natural circulation through steam generatorsActive: Shutdown cooling systemPassive: Natural circulation through steam generatorsand Passive Decay heat removal systemSafety injection systemEmergency core cooling systemSee also[edit]References[edit]^ “ANU SHAKTI: Atomic Energy In India”. BARC.^ “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 \u2013 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.^ Varadhan, Sudarshan (31 May 2022). “Operation of nuclear power unit in India’s western Gujarat state delayed”. Reuters. Retrieved 15 October 2022.^ a b “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 \u2013 Indian 220 MWe PHWR (IPHWR-220)” (PDF). International Automic Energy Agency. 4 April 2011. Retrieved 21 March 2021.{{cite news}}:  CS1 maint: url-status (link)^ Soni, Rakesh; Prasad, PN. “Fuel technology evolution for Indian PHWRs” (PDF). International Atomic Energy Agency. S. Vijayakumar, A.G. Chhatre, K.P.Dwivedi.{{cite news}}:  CS1 maint: url-status (link)^ 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\u20138): 701\u2013722. 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.^ “Status report 105 \u2013 Indian 700 MWe PHWR (IPHWR-700)” (PDF). International Atomic Energy Agency. 1 August 2011. Retrieved 20 March 2021.{{cite news}}:  CS1 maint: url-status (link)       (adsbygoogle = window.adsbygoogle || []).push({});after-content-x4"},{"@context":"http:\/\/schema.org\/","@type":"BreadcrumbList","itemListElement":[{"@type":"ListItem","position":1,"item":{"@id":"https:\/\/wiki.edu.vn\/en\/wiki21\/#breadcrumbitem","name":"Enzyklop\u00e4die"}},{"@type":"ListItem","position":2,"item":{"@id":"https:\/\/wiki.edu.vn\/en\/wiki21\/iphwr-700-wikipedia\/#breadcrumbitem","name":"IPHWR-700 – Wikipedia"}}]}]