Generation IV nuclear reactors

This EOE article is adapted from an information paper published by the World Nuclear Association (WNA). WNA information papers are frequently updated, so for greater detail or more up to date numbers, please see the latest version on WNA website (link at end of article).


 The Generation IV International Forum (GIF) was initiated in 2000 and formally chartered in mid 2001.  It is an international collective representing governments of 13 countries where nuclear energy is significant now and also seen as vital for the future.  Most are committed to joint development of the next generation of nuclear technology.  Led by the USA, Argentina, Brazil, Canada, France, Japan, South Korea, South Africa, Switzerland, and the UK are charter members of the GIF, along with the EU (Euratom).  Russia and China were admitted in 2006.  Argentina and Brazil have signed the GIF Charter but not the Framework Agreement (FA) which formally commits them to participate in the development of one or more Generation IV systems selected by GIF for further R&D.  The UK withdrew from the FA; accordingly, within the GIF, these three are designated as “inactive Members.”  Russia is working on the necessary approvals for its accession to the FA.

After some two years' deliberation, late in 2002, the Generation IV International Forum (GIF), then representing ten countries, announced the selection of six nuclear reactor technologies which they believe represent the future shape of nuclear energy. These were selected on the basis of being clean, safe, and cost-effective means of meeting increased energy demands on a sustainable basis, while being resistant to diversion of materials for weapons proliferation and secure from terrorist attacks. They will be the subject of further development internationally.

In addition to selecting these six concepts for deployment between 2010 and 2030, the GIF recognized a number of International Near-Term Deployment advanced nuclear power reactors available before 2015.

Most of the six systems employ a closed nuclear fuel cycle to maximize the resource base and minimize the amount of high-level wastes needed to be sent to a repository. Three of the six are fast reactors, one can be built as a fast reactor, one is described as epithermal, and only two operate with slow neutrons like today's plants.

Only one is cooled by light water, two are helium-cooled and the others have a lead-bismuth, sodium or fluoride salt coolant. The latter three operate at low pressure, with significant safety advantages. The last has the uranium fuel dissolved in the circulating coolant. Temperatures range from 510°C to 1000°C, compared with less than 330°C for today's light water reactors—this means that four of the reactors can be used for thermochemical hydrogen production.

The sizes range from 150 to 1500 MWe (or equivalent thermal), with the lead-cooled reactor optionally available as a 50-150 MWe "battery" with long core life (15-20 years without refuelling) as replaceable cassette or entire reactor module. This is designed for distributed generation or desalination.

At least four of the systems have significant operating experience already in most respects of their design, which may mean that they can be in commercial operation well before 2030.

However, it is significant that to address non-proliferation concerns, the fast neutron reactors are not conventional fast breeders, ie they do not have a blanket assembly where plutonium-239 is produced.  Instead, plutonium production takes place in the core, where burn-up is high and the proportion of plutonium isotopes other than Pu-239 remains high.  In addition, new reprocessing technologies will enable the fuel to be recycled without separating the plutonium.

In February 2005, five of the participants in GIF signed an agreement to take forward research and development on the six technologies. The USA, Canada, France, Japan and UK agreed to undertake joint research and exchange technical information.

While Russia was not initially part of GIF, one of GIF's design corresponds with the BREST reactor being developed there, and Russia is now the main operator of the sodium-cooled fast reactor for electricity—another of the technologies put forward by the GIF.

India is also not involved with GIF but is developing its own advanced technology to utilize thorium as a nuclear fuel. A three-stage program has the first stage well-established, with Pressurized Heavy Water Reactors (PHWRs, elsewhere known as CANDUs) fuelled by natural uranium to generate plutonium. Then, Fast Breeder Reactors (FBRs) use this plutonium-based fuel to breed uranium-233 (233U) from thorium, and finally, advanced nuclear power reactors will use the 233U. The spent fuel will be reprocessed to recover fissile materials for recycling. The two major options for the third stage, while continuing with the PHWR and FBR programs, are an Advanced Heavy Water Reactor and subcritical Accelerator-Driven Systems.

Closely related to GIF is the The Multinational Design Evaluation Program (MDEP) set up in 2005, led by the OECD Nuclear Energy Agency and involving the IAEA.  It aims to develop multinational regulatory standards for design of Gen IV reactors.  The US Nuclear Regulatory Commission (NRC) has proposed a three-stage process culminating in international design certification for these.  Ten countries are involved so far: Canada, China, Finland, France, Japan, Korea, Russia, South Africa, UK, USA, but others which have or are likely to have firm commitments to building new nuclear plants may be admitted.  In September 2007 the NRC called for countries involved in development of Gen IV reactors to move to stage 3 of design evaluation, which means developing common design requirements so that regulatory standards can be harmonised.  NRC has published its draft design requirements.

GIF Reactor technologies

caption A Generation IV gas-cooled fast reactor concept. Photo: U.S. DOE

Gas-cooled fast reactors: Like other helium-cooled reactors that have operated or are under development, GFRs will be high-temperature units—850°C.  They employ similar reactor technology to the VHTR, for power generation, thermochemical hydrogen production or other process heat.  The reference GFR unit is 1200 MWe, with thick steel reactor pressure vessel and three 800 MWt loops.  For electricity, the gas will directly drive a gas turbine (Brayton cycle). Fuels would include depleted uranium and any other fissile or fertile materials as ceramic pins or plates.  As with the SFR, used fuel would be reprocessed on site and all actinides would be recycled to minimize production of long-lived radioactive wastes.

While General Atomics (USA) worked on the design in the 1970s (but not as a fast reactor), none has so far been built. An 80 MWt experimental demonstration GFR, ALLEGRO, is planned by Euratom by 2011.  It will incorporate all the architecture and the main materials and components foreseen for the GFR without the power conversion system.  Euratom, France, Japan and Switzerland have signed on to System Arrangements (SA) for the GFR under the Framework Agreement.

caption A Generation IV lead-cooled fast reactor concept. Photo: U.S. DOE

Lead-cooled fast reactors: The LFR is a flexible fast neutron reactor which can use depleted uranium or thorium fuel matrices, and burn actinides from LWR fuel.  Liquid metal (Pb or Pb-Bi eutectic) cooling is at low pressure by natural convection (at least for decay heat removal).  Fuel is metal or nitride, with full actinide recycle from regional or central reprocessing plants.  A wide range of unit sizes is envisaged, from factory-built "battery" with 15-20 year life for small grids or developing countries, to modular 300-400 MWe units and large single plants of 1400 MWe.  Operating temperature of 550°C is readily achievable but 800°C is envisaged with advanced materials to provide lead corrosion resistance at high temperatures and this would enable thermochemical  hydrogen production.  A two-stage development program leading to industrial deployment is envisaged: by 2025 for reactors operating with relatively low temperature and power density, and by 2035 for more advanced higher-temperature designs.

This corresponds with Russia's BREST fast reactor technology which is lead-cooled and builds on 80 reactor-years experience of lead-bismuth cooling, mostly in submarine reactors. Its fuel is uranium+plutonium (U+Pu) nitride. More immediately, the GIF proposal appears to arise from two experimental designs: the US STAR and Japan's LSPR, being lead- and lead-bismuth-cooled respectively.

Initial development work on the LFR is focused on two pool-type reactors: SSTAR - Small Secure Transportable Autonomous Reactor of 20 MWe in USA and the European Lead-cooled SYstem (ELSY) of 600 MWe in Europe.  

SSTAR is being developed by Toshiba and others in Japan.  It runs at 566°C and has integral steam generator inside the sealed unit, which would be installed below ground level.  It is expected to have 44% thermal efficiency.  After a 20-year life without refuelling, the whole reactor unit is then returned for recycling the fuel.  The core is one metre high and 1.2 m diameter (20 MWe version).  SSTAR will eventually be coupled to a Brayton cycle turbine using supercritical carbon dioxide with natural circulation to four heat exchangers.  

The ELSY project is led by Ansaldo Nucleare from Italy and is being financed by Euratom.  The 600 MWe design was nearly complete in 2008 and a small-scale demonstration facility is planned.  It runs on MOX fuel at 480°C and the molten lead is pumped to eight steam generators. 

For the LFR, no System Arrangements (SA) have been signed, and collaborative R&D is pursued by interested members under the auspices of a provisional steering committee.  A technology pilot plant is envisaged in operation by 2020, followed by a prototypes of a large unit and deployment of small transportable units.

caption A Generation IV molten salt reactor concept. Photo: U.S. DOE

Molten salt reactors (now two variants).   In an MSR, the Uranium fuel is dissolved in the sodium fluoride salt coolant, which circulates through graphite core channels to achieve some moderation and an epithermal neutron spectrum. The reference plant is up to 1000 MWe.  Fission products are removed continuously and the actinides are fully recycled, while plutonium and other actinides can be added along with uranium-238 (238U) without the need for fuel fabrication. Coolant temperature is 700°C at very low pressure, with 800°C envisaged. A secondary coolant system is used for electricity generation, and thermochemical hydrogen production is also feasible.

Compared with solid-fuelled reactors, MSR systems have lower fissile inventories, no radiation damage constraint on fuel burn-up, no spent nuclear fuel, no requirement to fabricate and handle solid fuel, and a homogeneous isotopic composition of fuel in the reactor.  These and other characteristics may enable MSRs to have unique capabilities and competitive economics for actinide burning and extending fuel resources.

During the 1960s, the USA developed the molten salt fast reactor as the primary back-up option for the conventional fast breeder reactor, and a small prototype was operated. Recent work has focused on lithium and beryllium fluoride coolant with dissolved thorium and 233U fuel. The attractive features of the MSR fuel cycle include: high-level waste comprising fission products only, hence shorter-lived radioactivity; small inventory of weapons-fissile material (242Pu being the dominant plutonium isotope); low fuel use (the French self-breeding variant claims 50kg of thorium and 50kg 238U per billion kWh); and safety due to passive cooling up to any size.

For the MSR, no System Arrangements (SA) have been signed, and collaborative R&D is pursued by interested members under the auspices of a provisional steering committee.  There will be a long lead time to prototypes, and the R&D orientation has changed since the project was set up, due to increased interest.  It now has two baseline concepts:
- the Molten Salt Fast Neutron Reactor (MSFR)
- the Advanced High-Temperature Reactor (AHTR) with the same graphite core structures as the VHTR and molten salt as coolant instead of helium, enabling power densities 4 to 6 times greater than HTRs and power levels up to 4000 MWt with passive safety systems.

caption A Generation IV sodium-cooled fast reactor concept. Photo: U.S. DOE

Sodium-cooled fast reactors: The SFR uses liquid sodium as the reactor coolant, allowing high power density with low coolant volume.  It builds on more than 300 reactor-years experience with fast neutron reactors over five decades and in eight countries, but moves from a core plus blanket configuration to having all the neutron action in the core. It utilizes depleted uranium in the fuel and has a coolant temperature of 500-550°C enabling electricity generation via a secondary sodium circuit, the primary one being at near atmospheric pressure. Three variants are proposed: a 50-150 MWe type with actinides incorporated into a U-Pu metal fuel requiring electrometallurgical processing (pyroprocessing) integrated on site; a 300-1500 MWe pool-type version of this, and a 600-1500 MWe type with conventional MOX fuel and advanced aqueous reprocessing in central facilities elsewhere.

Early in 2008, the USA, France and Japan signed an agreement to expand their cooperation on the development of sodium-cooled fast reactor technology.  The agreement relates to their collaboration in the Global Nuclear Energy Partnership, aimed at closing the nuclear fuel cycle through the use of advanced reprocessing and fast reactor technologies, and seeks to avoid duplication of effort. 

Euratom, France, Japan, Korea and the USA have signed on to System Arrangements (SA) for the SFR under the Framework Agreement.  Three Project Arrangements have been signed within the SFR system: the Advanced Fuel PA; the Global Actinide Cycle International Demonstration (GACID) PA; and the Component Design and Balance-Of-Plant PA.

caption A Generation IV supercritical water-cooled reactor concept. Photo: U.S. DOE

Supercritical water-cooled reactors: This is a very high-pressure water-cooled nuclear reactor that operates above the thermodynamic critical point of water (374C, 22 MPa)  to give a thermal efficiency about one-third higher than today's light water reactors from which the design evolves. The supercritical water (25 MPa and 510-550°C) directly drives the turbine without any secondary steam system, simplifying the plant. Passive safety features are similar to those of simplified boiling water reactors. Fuel is uranium oxide, enriched in the case of the open fuel cycle option. However, it can be built as a fast reactor with full actinide recycling based on conventional reprocessing.

Euratom, Canada and Japan have signed on to System Arrangements (SA) for the SCWR under the Framework Agreement. Project Arrangements are pending for thermal-hydraulics and safety.  Pre-conceptual SCWR designs include Candu (Canada), LWR (Euratom) and Fast Neutron (Japan).

caption A Generation IV very high-temperature reactor concept. Photo: U.S. DOE

Very high-temperature gas reactors: These are graphite-moderated, helium-cooled reactors, based on substantial experience. The VHTR is the next step in the evolutionary development of high-temperature reactors and is primarily dedicated to the cogeneration of electricity and hydrogen.  Its high outlet temperature makes it attractive also for the chemical, oil and iron industries.

The core can be built of prismatic blocks such as the Japanese HTTR and the GTMHR under development by General Atomics and others in Russia, or it may be a pebble bed design, such as the Chinese HTR-10 and the PBMR under development in South Africa with international partners. Outlet temperature of over 900°C and aiming for 1000°C enables thermochemical hydrogen production via an intermediate heat exchanger, with electricity cogeneration, or direct high-efficiency driving of a gas turbine (Brayton cycle). There is some flexibility in fuels, but no recycling initially. Modules of 600 MW thermal are envisaged.  The VHTR has potential for high burn-up (150-200 GWd/t), completely passive safety, low operation and maintenance costs, and modular construction.

Euratom, Canada, France, Japan, China, Korea, Switzerland and the USA have signed on to the System Arrangement (SA) for the VHTR under the Framework Agreement.  South Africa is expected to do so in 2009.  Two Project Arrangements have been signed within the VHTR system: the Fuel and Fuel Cycle PA and the Hydrogen Production PA.  A Materials PA is pending and will involve PBMR Pty Ltd.

 neutron spectrum (fast/ thermal) coolant temperature(°C) pressure* fuel fuel cycle size(s)(MWe) uses
Gas-cooled fast reactors fast helium 850 high U-238 + closed, on site 1200electricity & hydrogen
Lead-cooled fast reactors fast lead or Pb-Bi 480-800 low U-238 + closed, regional 20-180**, 300-1200, 600-1000 electricity & hydrogen
Molten salt fast reactors fast fluoride salts 700-800 low UF in salt closed 1000 electricity & hydrogen
 Molten salt reactor - Advanced High Temperature Reactor thermal fluoride salts750-1000
 low UO2 particles in prism
 open 1000-1500 hydrogen
Sodium-cooled fast reactors fast sodium 550 low U-238 & MOX closed 30-150 300-1500, 1000-2000 electricity
Supercritical water-cooled reactors thermal or fast water 510-625very high UO2 open (thermal), closed (fast) 300-700 1000-1500 electricity
Very high temperature gas reactors thermal helium 900-1000 high UO2 prism or pebbles open 250-300 hydrogen & electricity
* high = 7-15 Mpa + = with some U-235 or Pu-239 ** 'battery' model with long cassette core life (15-20 yr) or replaceable reactor module.

 Research programs under GIF

A major project under the GIF is investigating the use of actinide-laden fuel assemblies in fast reactors as part of the sodium-cooled fast reactor program. The Global Actinide Cycle International Demonstration (GACID) is being undertaken by France's atomic energy commission (CEA), Japan's Atomic Energy Agency (JAEA) and the US Department of Energy (DOE).  The first stage will lead to demonstration fuel containing minor actinides being used in Japan's Monju reactor.

Further Reading



Hore-Lacy, I., & Association, W. (2010). Generation IV nuclear reactors. Retrieved from


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