Introduction

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 are 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.

The GIF was initiated in 2000 and formally chartered in mid-2001. It is an international collective representing governments of countries where nuclear energy is now significantly developed and seen as vital for the future. They are committed to joint development of the next generation of nuclear technology, known as Generation IV reactors. Led by the USA, Argentina, Brazil, Canada, France, Japan, South Korea, South Africa, Switzerland, and the UK are members of the GIF, along with the EU. Russia and China were admitted in 2006.

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.

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.

GIF Reactor technologies

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, these will be high-temperature units—850°C—suitable for power generation, thermochemical hydrogen production or other process heat. For electricity, the gas will directly drive a gas turbine (Brayton cycle). Fuels would include depleted uranium and any other fissile or fertile materials. Spent 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.

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

Lead-cooled fast reactors: Liquid metal (lead (Pb) or lead-bismuth (Pb-Bi)) cooling is by natural convection. Fuel is depleted uranium metal or nitride, with full actinide recycling from regional or central reprocessing plants. A wide range of unit sizes is envisaged, from a factory-built "battery" with a 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, which would enable thermochemical hydrogen production.

This corresponds with Russia's BREST fast reactor technology which is lead-cooled and builds on 40 years experience of lead-bismuth cooling 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.

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

Molten salt reactors (MSR): 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. Fission products are removed continuously and the actinides are fully recycled, while plutonium and other actinides can be added along with uranium-238 (²³⁸U). 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.

During the 1960s, the USA developed the molten salt breeder 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.

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

Sodium-cooled fast reactors: This design builds on more than 300 reactor-years experience with fast neutron reactors over five decades and in eight countries. It utilizes depleted uranium in the fuel and has a coolant temperature of 550°C enabling electricity generation via a secondary sodium circuit, the primary one being at near atmospheric pressure. Two variants are proposed: a 150-500 MWe type with actinides incorporated into a metal fuel requiring pyrometallurgical processing on site; and a 500-1500 MWe type with conventional MOX fuel reprocessed in conventional facilities elsewhere.

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

Supercritical water-cooled reactors: These are very high-pressure water-cooled nuclear reactors that operates above the thermodynamic critical point of water 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. 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. Most research on the design has been in Japan.

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 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 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. Modules of 600 MW thermal are envisaged.

Further Reading

• DOE 19/9/02
• DOE EIA 2003 New Reactor Designs
• Generation IV International Forum (GIF) Homepage
• Generation IV Nuclear Energy Systems (U. S. Department of Energy, Idaho National Laboratory)
• What is Generation IV? (U. S. Department of Energy, Office of Nuclear Energy)