Gas-cooled fast reactor

Gas-cooled fast reactor

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 using a direct Brayton 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.

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

Nuclear reactor design

Fast reactors were originally being developed for breeding fuel, because there was a concern that time that there was not enough uranium to fuel all of the nuclear reactors being built at the time, however; fuel supply was never an issue.

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, and of course there is no neutron moderator. Due to the higher fissile fuel content, the design has a higher power density than the HTGR.


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 normal 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.


The gas used can be many different types, including carbon dioxide or helium. The advantage of using gas as a coolant is gas does not absorb neutrons, making void coefficient not an issue. The use of gas also removes the vulnerability of a steam explosion, where the water in a water cooled reactor (PWR or BWR) flashes to steam, resulting in an explosion. Unlike water, gas stays as a gas and never changes, it does not turn to liquid, does not boil. The use of gas also allows for higher temperatures that are possible than other coolants, which increases the thermal efficiency, and such reactor could be used to produce hydrogen fuel. Most gas coolants do not absorb radiation, reducing the risk to workers around the primary coolant loop; and thus the primary loop is not radioactive.

Research History

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 1600°C). 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.

Gas cooled projects include decommissioned reactors such as the Dragon Project, 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 HTTR in Japan, which reached full power (30 MWth) using fuel compacts inserted in prismatic blocks in 1999, and the HTR-10 in China, which may reach 10 MWth in 2002 using pebble fuel. A 400 MWth pebble bed modular reactor demonstration plant is being designed by PBMR Pty for deployment in South Africa, and a consortium of Russian institutes is designing a 600 MWth GT-MHR (prismatic block reactor) in cooperation with General Atomics.

ee also

*Very high temperature reactor
*Fast breeder reactor
*Fast neutron reactor
*Generation IV reactor


# [ Flexibility of the Gas Cooled Fast Reactor to Meet the Requirements of the 21st Century]
# [ INL GFR summary]
# [ Generation IV International Forum GFR website]

External links

* [ IAEA Fast Reactors and Accelerator Driven Systems Knowledge Base]

* [ INL webpage]

* [ INL GFR workshop summary]

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