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Gas-cooled fast reactor

January 01, 1970

The gas-cooled fast reactor (GFR) system is a nuclear reactor design which is currently in development. Classed as a Generation IV reactor, it features a fast-neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reference reactor design is a helium-cooled system operating with an outlet temperature of 850 °C (1,560 °F) using a direct Brayton closed-cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being considered based on pin- or plate-based fuel assemblies or prismatic blocks, which allows for better coolant circulation than traditional fuel assemblies.

Gas-cooled fast reactor scheme

The reactors are intended for use in nuclear power plants to produce electricity, while at the same time producing (breeding) new nuclear fuel.

Reactor design edit

Fast reactors were originally designed to be primarily breeder reactors. This was because of a view at the time of their conception that there was an imminent shortage of uranium fuel for existing reactors. The projected increase in uranium price did not materialize, but if uranium demand increases in the future, then there may be renewed interest in fast reactors.

The GFR base design is a fast reactor, but in other ways similar to a high temperature gas-cooled reactor. It differs from the HTGR design in that the core has a higher fissile fuel content as well as a non-fissile, fertile, breeding component. There is no neutron moderator, as the chain reaction is sustained by fast neutrons. Due to the higher fissile fuel content, the design has a higher power density than the HTGR.

Fuel edit

In a GFR reactor design, the unit operates on fast neutrons; no moderator is needed to slow neutrons down. This means that, apart from nuclear fuel such as uranium, other fuels can be used. The most common is thorium, which absorbs a fast neutron and decays into Uranium 233. This means GFR designs have breeding properties—they can use fuel that is unsuitable in light water reactor designs and breed fuel. Because of these properties, once the initial loading of fuel has been applied into the reactor, the unit can go years without needing fuel. If these reactors are used for breeding, it is economical to remove the fuel and separate the generated fuel for future use.

Coolant edit

The gas used can be many different types, including carbon dioxide or helium. It must be composed of elements with low neutron capture cross sections to prevent positive void coefficient and induced radioactivity. The use of gas also removes the possibility of phase transition–induced explosions, such as when the water in a water-cooled reactor (PWR or BWR) flashes to steam upon overheating or depressurization. The use of gas also allows for higher operating temperatures than are possible with other coolants, increasing thermal efficiency, and allowing other non-mechanical applications of the energy, such as the production of hydrogen fuel.

Research history edit

Past pilot and demonstration projects have all used thermal designs with graphite moderators. As such, no true gas-cooled fast reactor design has ever been brought to criticality. The main challenges that have yet to be overcome are in-vessel structural materials, both in-core and out-of-core, that will have to withstand fast-neutron damage and high temperatures (up to 1,600 °C [2,910 °F]). Another problem is the low thermal inertia and poor heat removal capability at low helium pressures, although these issues are shared with thermal reactors which have been constructed. Peter Fortescue, whilst at General Atomic, was leader of the team responsible for the initial development of the High temperature gas-cooled reactor (HTGR), as well as the Gas-cooled Fast Reactor (GCFR) system.[1]

Gas-cooled projects (thermal spectrum) include decommissioned reactors such as the Dragon reactor, built and operated in the United Kingdom, the AVR and the THTR-300, built and operated in Germany, and Peach Bottom and Fort St. Vrain, built and operated in the United States. Ongoing demonstrations include the High-temperature engineering test reactor in Japan, which reached full power (30 MWth) using fuel compacts inserted in prismatic blocks in 1999, and the HTR-10 in China, which reached its full effect at 10 MWth in 2003 using pebble fuel. A 400 MWth pebble bed modular reactor demonstration plant was designed by PBMR Pty for deployment in South Africa but withdrawn in 2010, and a consortium of Russian institutes is designing a 600 MWth GT-MHR (prismatic block reactor) in cooperation with General Atomics. In 2010, General Atomics announced the Energy Multiplier Module reactor design, an advanced version of the GT-MHR.

A European gas cooled fast reactor (GFR) demonstrator, ALLEGRO, is currently being developed by Czech Republic, France, Hungary, Slovakia and Poland. The primary aim of ALLEGRO is to create a conceptual design of a helium-cooled fast reactor with passive decay heat removal during LOCA accidents based on nitrogen injections into the guard vessel containing the reactor pressure vessel and to design an air-tight guard vessel capable of withstanding the increased pressure (over 10 bar) and temperature during the LOCA accident.[2]

See also edit

References edit

  1. ^ Fouquet, Doug. "Peter Fortescue Dies at 102". Retrieved 20 November 2021 – via General Atomics.
  2. ^ Kvizda, Boris (2019). "ALLEGRO Gas-cooled Fast Reactor (GFR) demonstrator thermal hydraulic benchmark". Nuclear Engineering and Design. 345: 47–61. doi:10.1016/j.nucengdes.2019.02.006. S2CID 116688540. Retrieved June 11, 2022.
  • "Gas-Cooled Fast Reactor (GFR) Fact Sheet". Idaho National Laboratory.
  • Van Rooijen, W. F. G. (2009). "Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook". Science and Technology of Nuclear Installations. 2009: 1–11. doi:10.1155/2009/965757.

External links edit

  • INL GFR workshop summary

January 01, 1970
cooled, fast, reactor, this, article, multiple, issues, please, help, improve, discuss, these, issues, talk, page, learn, when, remove, these, template, messages, this, article, needs, additional, citations, verification, please, help, improve, this, article, . This article has multiple issues Please help improve it or discuss these issues on the talk page Learn how and when to remove these template messages This article needs additional citations for verification Please help improve this article by adding citations to reliable sources Unsourced material may be challenged and removed Find sources Gas cooled fast reactor news newspapers books scholar JSTOR November 2015 Learn how and when to remove this template message This article includes a list of general references but it lacks sufficient corresponding inline citations Please help to improve this article by introducing more precise citations November 2015 Learn how and when to remove this template message Learn how and when to remove this template message The gas cooled fast reactor GFR system is a nuclear reactor design which is currently in development Classed as a Generation IV reactor it features a fast neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides The reference reactor design is a helium cooled system operating with an outlet temperature of 850 C 1 560 F using a direct Brayton closed cycle gas turbine for high thermal efficiency Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products composite ceramic fuel advanced fuel particles or ceramic clad elements of actinide compounds Core configurations are being considered based on pin or plate based fuel assemblies or prismatic blocks which allows for better coolant circulation than traditional fuel assemblies Gas cooled fast reactor scheme The reactors are intended for use in nuclear power plants to produce electricity while at the same time producing breeding new nuclear fuel Contents 1 Reactor design 2 Fuel 3 Coolant 4 Research history 5 See also 6 References 7 External linksReactor design editFast reactors were originally designed to be primarily breeder reactors This was because of a view at the time of their conception that there was an imminent shortage of uranium fuel for existing reactors The projected increase in uranium price did not materialize but if uranium demand increases in the future then there may be renewed interest in fast reactors The GFR base design is a fast reactor but in other ways similar to a high temperature gas cooled reactor It differs from the HTGR design in that the core has a higher fissile fuel content as well as a non fissile fertile breeding component There is no neutron moderator as the chain reaction is sustained by fast neutrons Due to the higher fissile fuel content the design has a higher power density than the HTGR Fuel editIn a GFR reactor design the unit operates on fast neutrons no moderator is needed to slow neutrons down This means that apart from nuclear fuel such as uranium other fuels can be used The most common is thorium which absorbs a fast neutron and decays into Uranium 233 This means GFR designs have breeding properties they can use fuel that is unsuitable in light water reactor designs and breed fuel Because of these properties once the initial loading of fuel has been applied into the reactor the unit can go years without needing fuel If these reactors are used for breeding it is economical to remove the fuel and separate the generated fuel for future use Coolant editThe gas used can be many different types including carbon dioxide or helium It must be composed of elements with low neutron capture cross sections to prevent positive void coefficient and induced radioactivity The use of gas also removes the possibility of phase transition induced explosions such as when the water in a water cooled reactor PWR or BWR flashes to steam upon overheating or depressurization The use of gas also allows for higher operating temperatures than are possible with other coolants increasing thermal efficiency and allowing other non mechanical applications of the energy such as the production of hydrogen fuel Research history editPast pilot and demonstration projects have all used thermal designs with graphite moderators As such no true gas cooled fast reactor design has ever been brought to criticality The main challenges that have yet to be overcome are in vessel structural materials both in core and out of core that will have to withstand fast neutron damage and high temperatures up to 1 600 C 2 910 F Another problem is the low thermal inertia and poor heat removal capability at low helium pressures although these issues are shared with thermal reactors which have been constructed Peter Fortescue whilst at General Atomic was leader of the team responsible for the initial development of the High temperature gas cooled reactor HTGR as well as the Gas cooled Fast Reactor GCFR system 1 Gas cooled projects thermal spectrum include decommissioned reactors such as the Dragon reactor built and operated in the United Kingdom the AVR and the THTR 300 built and operated in Germany and Peach Bottom and Fort St Vrain built and operated in the United States Ongoing demonstrations include the High temperature engineering test reactor in Japan which reached full power 30 MWth using fuel compacts inserted in prismatic blocks in 1999 and the HTR 10 in China which reached its full effect at 10 MWth in 2003 using pebble fuel A 400 MWth pebble bed modular reactor demonstration plant was designed by PBMR Pty for deployment in South Africa but withdrawn in 2010 and a consortium of Russian institutes is designing a 600 MWth GT MHR prismatic block reactor in cooperation with General Atomics In 2010 General Atomics announced the Energy Multiplier Module reactor design an advanced version of the GT MHR A European gas cooled fast reactor GFR demonstrator ALLEGRO is currently being developed by Czech Republic France Hungary Slovakia and Poland The primary aim of ALLEGRO is to create a conceptual design of a helium cooled fast reactor with passive decay heat removal during LOCA accidents based on nitrogen injections into the guard vessel containing the reactor pressure vessel and to design an air tight guard vessel capable of withstanding the increased pressure over 10 bar and temperature during the LOCA accident 2 See also editEnergy Multiplier Module Fast breeder reactor Fast neutron reactor Generation IV reactor PBMR Very high temperature reactor HTR 10 HTR PMReferences edit Fouquet Doug Peter Fortescue Dies at 102 Retrieved 20 November 2021 via General Atomics Kvizda Boris 2019 ALLEGRO Gas cooled Fast Reactor GFR demonstrator thermal hydraulic benchmark Nuclear Engineering and Design 345 47 61 doi 10 1016 j nucengdes 2019 02 006 S2CID 116688540 Retrieved June 11 2022 Gas Cooled Fast Reactor GFR Fact Sheet Idaho National Laboratory Van Rooijen W F G 2009 Gas Cooled Fast Reactor A Historical Overview and Future Outlook Science and Technology of Nuclear Installations 2009 1 11 doi 10 1155 2009 965757 External links editIAEA Fast Reactors and Accelerator Driven Systems Knowledge Base INL GFR workshop summary Retrieved from https en wikipedia org w index php title Gas cooled fast reactor amp oldid 1185610231, wikipedia, wiki, book, books, library,

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