An International forum of GIF countries are working jointly on developing Fourth Generation Nuclear Reactor Technologies:
Generation IV Nuclear Reactors: WNA
Generation IV Nuclear Reactors
(updated December 2010)
•An international task force is developing six nuclear reactor technologies for deployment between 2020 and 2030. Four are fast neutron reactors.
•All of these operate at higher temperatures than today's reactors. In particular, four are designated for hydrogen production.
•All six systems represent advances in sustainability, economics, safety, reliability and proliferation-resistance.
•Europe is pushing ahead with three of the fast reactor designs.
GIF
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, China, France, Japan, Russia, South Korea, South Africa, Switzerland, and the UK are charter members of the GIF, along with the EU (Euratom). Most of these are party to 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. Argentina and Brazil did not sign the FA, and the UK withdrew from it; accordingly, within the GIF, these three are designated as “inactive Members.” Russia formalized its accession to the FA in August 2009 as its tenth member, with Rosatom as implementing agent. In 2011 the 13 members decided to modify and extend the GIF charter indefinitely.
GIF focus
After some two years' deliberation and review of about one hundred concepts, GIF (then representing ten countries) late in 2002 announced the selection of six 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 are the subject of further development internationally, and expenditure so far is in line with the initial estimate of $6 billion over 15 years. About 80% of the cost is being met by the USA, Japan and France.
In addition to selecting these six concepts for deployment between 2010 and 2030, the GIF recognised a number of International Near-Term Deployment advanced reactors available before 2015. (see Advanced Reactors paper )
Most of the six systems employ a closed fuel cycle to maximise the resource base and minimise high-level wastes to be sent to a repository. Three of the six are fast neutron reactors (FNR) and 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 lead-bismuth, sodium or fluoride salt coolant. The latter three operate at low pressure, with significant safety advantage. 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, and this means that four of them can be used for thermochemical hydrogen production.
The sizes range from 150 to 1500 MWe (or equivalent thermal) , with the lead-cooled one 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 provides a good basis for further R&D and is likely to 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 signed an agreement to take forward the R&D on the six technologies. The USA, Canada, France, Japan and UK agreed to undertake joint research and exchange technical information.
Associated ongoing programs
While Russia was not initially part of GIF, one 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 the GIF but is developing its own advanced technology to utilise thorium as a nuclear fuel. A three-stage program has the first stage well-established, with Pressurised 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 U-233 from thorium, and finally advanced nuclear power systems will use the U-233. 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 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.
A major project 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) under the US Advanced Fuel Cycle Initiative (AFCI). The first stage will lead to demonstration fuel containing minor actinides being used in Japan's Monju reactor.
GIF Reactor technologies:
There were originally six technologies chosen, but development on one has gone in two directions, so seven are listed in the Table below.
Gas-cooled fast reactors. Like other helium-cooled reactors which have operated or are under development, GFRs will be high-temperature units - 850°C. They employ similar reactor technology to the VHTR, suitable 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 helium will directly drive a gas turbine (Brayton cycle). It would have a self-generating (breeding) core with fast neutron spectrum and no fertile blanket. Robust nitride or carbide fuels would include depleted uranium and any other fissile or fertile materials as ceramic pins or plates, with plutonium content of 15 to 20%. As with the SFR, used fuel would be reprocessed on site and all the actinides recycled repeatedly to minimise production of long-lived radioactive wastes.
While General Atomics worked on the design in the 1970s (but not as fast reactor), none has so far been built. It is the only Gen IV design with no operating antecedent, so a prototype is not expected before 2025. However, an 80 MWt experimental technology demonstration GFR, ETDR or ALLEGRO, is planned by Euratom to be built from 2014. 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. See also European program section below.
An alternative GFR design has lower temperature (600-650ºC) helium cooling in a primary circuit and supercritical CO2 at 550ºC and 20 MPa in a secondary system for power generation. This reduces the metallurgical and fuel challenges associated with very high temperatures.
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 or lead-bismuth cooling, mostly in submarine reactors. Its fuel is U+Pu nitride. More immediately the GIF proposal appears to arise from two experimental designs: the US STAR and Japan's LSPR, these 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. In 2011 Russia joined them. A technology pilot plant is envisaged in operation by 2020, followed by a prototype of a large unit and deployment of small transportable units. See also European program section below.
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 U-238, 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 for about four years. Recent work has focused on lithium and beryllium fluoride coolant with dissolved thorium and U-233 fuel. The attractive features of the MSR fuel cycle include: the high-level waste comprising fission products only, hence shorter-lived radioactivity; small inventory of weapons-fissile material (Pu-242 being the dominant Pu isotope); low fuel use (the French self-breeding variant claims 50kg of thorium and 50kg U-238 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.
Sodium-cooled fast reactors . The SFR uses liquid sodium as the reactor coolant, allowing high power density with low coolant volume. It builds on some 390 reactor-years experienced with sodium-cooled fast neutron reactors over five decades and in eight countries, and is the main technology of interest in GIF. Most plants so far have had a core plus blanket configuration, but new designs are likely to have all the neutron action in the core. Other R&D is focused on safety in loss of coolant scenarios, and improved fuel handling.
The SFR utilises depleted uranium as the fuel matrix 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, China, France, Japan, Korea and the USA have signed on to System Arrangements (SA) for the SFR under the Framework Agreement, and in 2011 Russia joined them. 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. See also European program section below.
Supercritical water-cooled reactors . This is a very high-pressure water-cooled reactor which operates above the thermodynamic critical point of water (374ºC, 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. Two design options are considered: pressure vessel and pressure tube. 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 recycle 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).
* Today's supercritical coal-fired plants use supercritical water around 25 MPa which have "steam" temperatures of 500 to 600ºC and can give 45% thermal efficiency. At ultra supercritical levels (30+ MPa), 50% thermal efficiency may be attained. Over 400 such plants are operating world-wide.
Supercritical fluids are those above the thermodynamic critical point, defined as the highest temperature and pressure at which gas and liquid phases can co-exist in equilibrium. They have properties between those of gas and liquid. For water the critical point is at 374°C and 22 MPa, giving it a "steam" density one third that of the liquid so that it can drive a turbine in a similar way to normal steam.
Very high-temperature gas reactors . These are graphite-moderated, helium-cooled reactors, based on substantial experience.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).
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 pebble bed such as the Chinese HTR-10 or HTR-PM and the PBMR under development in South Africa, with international partners. Outlet temperature of over 900°C and aiming for 1000ºC 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 recycle 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.
