Fast neutron reactors (FBR)
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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).
Contents
Introduction
In fast neutron reactors, most fission reactions are generated by neutrons with energy levels of the same order of magnitude as when they were produced by fission. These reactors exploit the improved efficiency, in terms of fission, of neutrons that maintain the speed acquired from previous fissions. Sometimes loosely referred to as "fast" reactors, they accept a wider variety of fuel isotopes than pressurized water reactors at the cost of a higher neutron flux in the reactor, and a high concentration of fissile isotopes in the fuel. The BN-350 fast nuclear reactor in Western Kazakhstan. (Source: The Institute of Physics and Power Engineering, Obninsk, Kalug) Approximately 20 fast neutron reactors have been in operation worldwide, some since the 1950s, with some supplying electricity commercially. Through these operations, over 300 reactor-years of operating experience have been accumulated. Fast neutron reactors more deliberately use the uranium (Fast neutron reactors (FBR)) isotope 238U as well as the fissile 235U isotope used in most reactors. If they are designed to produce more plutonium (Pu) than they consume, they are called fast breeder reactors (FBR). Conversely, if they are net consumers of plutonium they are sometimes called "burners".
Several countries have research and development programs underway with aims to improve fast neutron reactors, and the International Atomic Energy Agency (IAEA) INPRO program, involving 22 countries, focuses on fast neutron reactors in connection with a closed fuel cycle. For instance, one scenario for France is for half of the present nuclear capacity to be replaced by fast neutron reactors by 2050 (the first half being replaced by 3rd-generation EPR units).
The FBR was originally conceived to extend the world's uranium (U) resources, which they accomplished by a factor of approximatly 60. When uranium resources were perceived to be scarce, several countries embarked upon extensive FBR development programs. However, significant technical and materials problems were encountered, and geological exploration showed by the 1970s that scarcity should not be a concern for some time. Due to both factors, by the 1980s it was clear that FBRs would not be commercially competitive with existing light water reactors.
While there has been progress on the technical front, the economics of FBRs still depends on the value of the plutonium fuel which is bred, relative to the cost of fresh uranium. There is also international concern over the disposal of ex-military plutonium, and there are proposals to use fast reactors for this purpose. In both respects, the technology is important to long-term considerations of world energy sustainability.
The fast reactor has no moderator and uses plutonium as its basic fuel since it fissions sufficiently with fast neutrons to keep going. At the same time, the number of neutrons produced per fission is 25% more than from uranium, and this means that there are enough neutrons (after losses) not only to maintain the chain reaction but also to convert 238U in a "fertile blanket" around the core into fissile plutonium. In other words, the fast reactor "burns" and can "breed" plutonium.
Fast neutron reactors have a high power density and are normally cooled by liquid metal such as sodium, lead, or lead-bismuth, with high conductivity and boiling point and no moderating effect. They operate at around 500-550ºC, at or near atmospheric pressure. Fast reactors typically use boron carbide control rods.
In some respects, a liquid metal coolant is more benign overall than very high pressure water, which requires robust engineering on account of the pressure. However, the design needs to ensure that there is no chemical interaction (e.g., sodium-water), and if lead-cooled, the materials used need to be capable of managing the corrosivity of molten lead. In an effort to eliminate these complications, there are future plans for design of gas-cooled fast reactors. Also, fast reactors have a strong negative temperature coefficient (the reaction slows as the temperature rises unduly), an inherent safety feature, and the basis of automatic load following in many new designs.
Experiments on a 19-year-old UK breeder reactor, before it was decommissioned in 1977, and on EBR-II at the National Reactor Testing Station (now the Idaho National Laboratory) near Arco, Idaho, U.S. in 1986, showed that the metal fuel with a liquid sodium cooling system made them less sensitive to coolant failures than the more conventional very high pressure water and steam systems in light water reactors, which, with loss of coolant flow simply shut themselves down. More recent operating experience with large French and UK prototypes has confirmed this.
There is renewed interest in fast reactors due to their ability to fission actinides, including those which may be recovered from ordinary reactor used fuel. The fast neutron environment minimizes neutron capture reactions and maximizes fission in actinides. This results in less long-lived nuclides in high-level wastes (the fission products being preferable due to shorter lives).
Europe, Russia, Kazakhstan
Europe
France has operated its fast reactor Phenix since 1973, apart from a few years for refurbishing. Closure of the 1250 MWe commercial prototype Superphenix fast breeder reactor (FBR) in 1998 on political grounds after little operation over 13 years caused developmental setbacks. Research work on the 1450 MWe European FBR has almost ceased.
In the UK, the Dounreay Fast Reactor started operating in 1959 using sodium-potassium coolant. This was followed by the much larger Prototype Fast Reactor which operated for 20 years until withdrawal of government funding.
Russia
The Russian BN-600 fast breeder reactor - Beloyarsk unit 3 - has been supplying electricity to the grid since 1980 and is said to have the best operating and production record of all Russia's nuclear power units. It uses uranium oxide fuel, some enriched to over 20%, and the sodium coolant delivers 550°C at little more than atmospheric pressure. Russia plans to reconfigure the BN-600 to burn the plutonium from its military stockpiles and extend its life beyond the 30 year design span.
Construction has started on Beloyarsk-4, the first BN-800, a new larger (880 MWe) FBR with improved features including fuel Flexibility—U + Pu nitride, MOX, or metal—and a breeding ratio of up to 1.3. Its safety features and economy are considerably enhanced—operating cost is expected to be only 15% more than VVER (Russia's pressurized water reactor). It is capable of burning 2 tonnes of plutonium per year from dismantled weapons, and will test the recycling of minor actinides in the fuel.
Planning is underway for the development of more BN-800 units. However, industry spokesmen have warned the government that Russia's world leadership in FBR development is being threatened due to lack of funding for completion of BN-800.
Russia has experimented with several lead-cooled reactor designs, and has used lead-bismuth cooling for 40 years in reactors for its Alfa class submarines. Pb-208 (54% of naturally-occurring lead) is transparent to neutrons. A significant new Russian design is the BREST fast neutron reactor, of 300 MWe or more with lead as the primary coolant, at 540° C, and supercritical steam generators. It is inherently safe and uses a U+Pu nitride fuel. No weapons-grade plutonium can be produced (since there is no uranium blanket), and spent fuel can be recycled indefinitely at on-site facilities. A pilot unit is being built at Beloyarsk and 1200 MWe units are planned.
A smaller and newer Russian design is the Lead-Bismuth Fast Reactor (SVBR) of 75-100 MWe. This is an integral design, with the steam generators sitting in the same Pb-Bi pool at 400-480ºC as the reactor core, which could use a wide variety of fuels. The unit would be factory-made and shipped as a 4.5m diameter, 7.5m high module, then installed in a tank of water which gives passive heat removal and shielding. A power station with 16 such modules is expected to supply electricity at lower cost than any other new Russian technology, while achieving inherent safety and high proliferation resistance. Russia built 7 Alfa-class submarines, each powered by a compact 155 MWt Pb-Bi-cooled reactor, acquiring 70 reactor-years of operational experience.
Kazakhstan
Kazakhstan's BN-350 prototype FBR, preceeded by Russia's BOR-60 demonstration model, generated power for 27 years, until 1999. Approximately half of its 1000 MW (thermal) output was used for water desalination. It used uranium enriched to 17-26% and had a design life of 20 years; after 1993 it operated on the basis of annual licence renewal.
Japan, India, China
Japan
A significant portion of Japanese energy policy has been devoted to development of FBRs in an effort to dramatically improve uranium utilization. From 1961 to 1994 there was a strong commitment to FBRs, but in 1994 the FBR commercial timeline was pushed out to 2030, and in 2005 commercial FBRs were envisaged by 2050.
In 1999 Japan Nuclear Cycle Development Institute (JNC) initiated a program to review promising concepts, define a development plan by 2005, and establish a system of FBR technology by 2015. The parameters are: passive safety, economic competitiveness with light water reactors, efficient utilization of resources (burning transuranics and depleted uranium), reduced wastes, proliferation resistance, and versatility (include hydrogen production). Utilities are also involved in the development process.
Phase 2 of Japan's study focused on four basic reactor designs: sodium-cooled with MOX and metal fuels, helium-cooled with nitride and MOX fuels, lead-bismuth eutectic-cooled with nitride and metal fuels, and supercritical water-cooled with MOX fuel. All involve closed fuel cycle, and three reprocessing routes were considered: advanced aqueous, oxide electrowinning and metal pyroprocessing (electrorefining). This work is linked with the Generation IV initiative, in which Japan is playing a leading role with sodium-cooled FBRs.
Japan's Joyo experimental reactor which has been operating since 1977 is now being boosted to 140 MWt.
The 280 MWe Monju prototype FBR reactor began operation in April 1994, but a sodium leakage in its secondary heat transfer system during performance tests in 1995 necessitated shut down—the facility has not operated since. During its operation, it produced 246 MWe. Oversight of the facility was passed on to JNC, and the Minister for Science & Technology has stated its early restart as a key aim. A Supreme Court decision in May 2005 cleared the way for restarting it, probably in 2008.
Japan's LSPR is a lead-bismuth cooled reactor design of 150 MWt/53 MWe. Fuelled units would be supplied from a factory and operate for 30 years, then be returned.
A small-scale design developed by Toshiba Corporation in cooperation with Japan's Central Research Institute of Electric Power Industry (CRIEPI) and funded by the Japan Atomic Energy Research Institute (JAERI) is the 5 MWt, 200 kWe Rapid-L, using lithium-6 (a liquid neutron poison) as a control medium. It would have 2700 fuel pins of 40-50% enriched uranium nitride with a 2600ºC melting point integrated into a disposable cartridge. The reactivity control system is passive, using lithium expansion modules (LEM) which give burnup compensation, partial load operation, and negative reactivity feedback. As the reactor temperature rises, the lithium expands into the core, displacing an inert gas. Other kinds of lithium modules, also integrated into the fuel cartridge, shut down and start up the reactor. Cooling is by molten sodium, and with the LEM control system, reactor power is proportional to primary coolant flow rate. Refuelling would be every 10 years in an inert gas environment. Operation would require no skill, due to the inherent safety design features. The whole plant would be about 6.5 metres high and 2 metres diameter.
The Super-Safe, Small & Simple - the "4S" 'nuclear battery' system is being developed by Toshiba and CRIEPI in Japan in collaboration with STAR work in USA. It uses sodium as coolant (with electromagnetic pumps) and has passive safety features, notably negative temperature and void reactivity. The whole unit would be factory-built, transported to site, installed below ground level, and would drive a steam cycle. It is capable of three decades of continuous operation without refuelling. Metallic fuel (169 pins 10mm diameter) is uranium-zirconium or U-Pu-Zr alloy enriched to less than 20%. Steady power output over the core lifetime is achieved by progressively moving upwards an annular reflector around the slender core (0.68m diameter, 2m high). After 14 years, a neutron absorber at the center of the core is removed and the reflector repeats its slow movement up the core for 16 more years. In the event of power loss, the reflector falls to the bottom of the reactor vessel, slowing the reaction, and external air circulation gives decay heat removal.
Both 10 MWe and 50 MWe versions of 4S are designed to automatically maintain an outlet coolant temperature of 510°C - suitable for power generation with high temperature electrolytic hydrogen production. Plant cost is projected at US$ 2500/kW and power cost 5-7 cents/kWh for the small unit - very competitive with diesel in many locations. The design has gained considerable support in Alaska; toward the end of 2004 the town of Galena, Alaska granted initial approval for Toshiba to build a 4S reactor in their remote location. A pre-application review by the U.S. Nuclear Regulatory Commission (NRC) is being sought with hopes of starting operation of a demonstration unit by 2012. The Galena reactor's design is sufficiently similar to PRISM—GE's modular 150 MWe liquid metal-cooled inherently-safe reactor, which went part-way through U.S. NRC approval process—to give it favorable prospects of acquiring licensing.
India
In India, nuclear reactor research continues. At the Indira Gandhi Centre for Atomic Research, a 40 MWt fast breeder test reactor (FBTR) has been operating since 1985. In addition, the small-capacity (30 Kw) Kamini reactor (Kalpakkam Mini reactor) is employed to explore the use of thorium as a nuclear fuel by breeding fissile 233U.
In 2002 the regulatory authority issued approval to start construction of a 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam; the facility is now under construction by BHAVINI. It is expected to be operating in 2010, fuelled with uranium-plutonium oxide (the reactor-grade Pu coming from its existing PHWRs) and with a thorium blanket to breed fissile 233U. Completion of this facility will take India's ambitious thorium program to stage 2 and set the scene for an eventual full utilization of the country's abundant thorium resources for reactor fuelling. Four more fast reactors have been announced for construction by 2020. Initial Indian FBRs will run on mixed oxide fuel but will be followed by metallic-fuelled reactors that will enable shorter doubling time.
China
In China, a 65 MWt fast neutron reactor—the Chinese Experimental Fast Reactor (CEFR)—is under construction near Beijing and due to achieve criticality in 2008. There has been some Russian assistance in its development. Chinese research and development on fast neutron reactors started in 1964. A full-scale prototype fast reactor is envisaged by 2020, and the China National Nuclear Corporation (CNNC) expects the technology to become predominant by mid-century.
USA
In the USA, five fast neutron reactors have operated, and several more have been designed. In 1951, the experimental breeder reactor EBR-1, located at the National Reactor Testing Station (now the Idaho National Laboratory), near Arco, Idaho, produced enough power to run its own building—a milestone achievement.
The EBR-2 (62.5 MW thermal) was a demonstration reactor that typically operated at 19 MWe, providing heat and power to the Idaho facility. The idea was to demonstrate a complete sodium-cooled breeder reactor power plant with on-site reprocessing of metallic fuel, successfully done from 1964-69. The emphasis then shifted to testing materials and fuels (metal and ceramic oxides, carbides and nitrides of uranium and plutonium) for larger fast reactors. The EBR-2 would eventually become the prototype for the U.S. Integral Fast Reactor (IFR) program, using metallic alloy U-Pu-Zr fuels. All the time, it generated some 1 TWh of power as well.
The EBR-2 was integral to the IFR program, considered by the National Academy of Sciences to be the nation's highest research priority for development of future reactors. Research focused on developing a fully-integrated system with pyro-reprocessing, fuel fabrication and fast reactor in the same complex; the reactor could be operated as a breeder or not. Some US$46 million of IFR funding was provided by a Japanese utility consortium.
IFR program goals included demonstrating inherent safety apart from engineered controls, improved management of high-level nuclear wastes by recycling all actinides so that only fission products remain as high-level waste (HLW), and achieving the full energy potential of uranium rather than only about one percent. Although all of these goals were demonstrated, the program was aborted before recycling of neptunium and americium was properly evaluated. Fuel developed in the IFR program, first used in 1986, reached 19% burnup (compared with 3-4% for conventional reactors), and 22% was targeted.
A further political goal was demonstrating a proliferation-resistant closed fuel cycle, with plutonium being recycled with other actinides.
In 1994, Congress under the Clinton administration shut down EBR-2. The IFR program is now being reinvented as part of the Global Nuclear Energy Partnership (see below), while EBR-2 is being decommissioned. An EBR-3 of 200-300 MWe was proposed but never developed.
The first commercial fast breeder reactor (FBR) in the U.S. was Fermi-1 in Michigan, which operated for only three years before a coolant problem caused overheating and shut down with some damage to the fuel. After repair, it was restarted in 1970, but its licence was not renewed in 1972.
The 400 MWt Fast Flux Test Facility was in full operation 1982-92 at the Hanford Site, Washington as a major national research reactor. It was closed down at the end of 1993, and has been in decommission since 2001.
GE was involved in designing a modular 150 MWe liquid metal-cooled inherently-safe reactor—PRISM. GE and Argonne National Laboratory (ANL) have also been developing an advanced liquid-metal fast breeder reactor (ALMR) of over 1400 MWe, but both designs at an early stage were withdrawn from review of the Nuclear Regulatory Commission. Thus far, no US fast neutron reactor has exceeded capacity greater than 66 MWe, and none has supplied electricity commercially.
The Super-PRISM is a GE advanced reactor design for compact modular pool-type reactors with passive cooling and decay heat removal. Modules have a capacity of 1000 MWt and operate at higher temperature—510ºC—than the original PRISM. The pool-type modules contain the complete primary system with sodium coolant. The plutonium and depleted uranium fuel can be oxide or metal, but minor actinides are not removed in reprocessing so that even fresh fuel is intensely radioactive and hence resistant to misappropriation. The fission products are removed in reprocessing and resultant wastes are shorter-lived than usual. Fuel stays in the reactor six years, with one-third removed every two years. The commercial plant concept uses six reactor modules to provide 2280 MWe, and the design meets Generation IV criteria including generation cost of under 3 cents/kWh.
The Encapsulated Nuclear Heat Source (ENHS) concept is a liquid metal-cooled reactor of 50 MWe being developed by the University of California. The core is in a metal-filled module sitting in a large pool of secondary molten metal coolant that also accommodates the separate and unconnected steam generators. Fuel is a uranium-zirconium alloy with 13% U enrichment (or U-Pu-Zr with 11% Pu) with a 15-year life. After this period, the module is removed, stored on site until the primary lead (or Pb-Bi) coolant solidifies, and then shipped as a self-contained and shielded item. A new fuelled module would be supplied complete with primary coolant. The ENHS is designed for developing countries but is not yet close to commercialization.
A related project is the Secure Transportable Autonomous Reactor (STAR) being developed by Argonne under the leadership of Lawrence Livermore Laboratory of the U.S. Department of Energy (DOE). The design is a lead-cooled fast neutron modular reactor with passive safety features. Its 400 MWt capacity means that it can be shipped by rail and cooled by natural circulation. It uses U-transuranic nitride fuel in a cassette that is replaced every 15-20 years. The STAR-LM was conceived for power generation, running at 578ºC and producing 180 MWe.
STAR-H2 is an adaptation for hydrogen production, with reactor heat of up to 800ºC being conveyed by a helium circuit to drive a separate thermochemical hydrogen production plant, while lower grade heat is harnessed for desalination (multi-stage flash process). Any commercial electricity generation would then be powered by fuel cells run on the hydrogen. Development of the facility is further off.
A smaller STAR variant is the Small Sealed Transportable Autonomous Reactor (SSTAR) being developed in collaboration with Toshiba and others in Japan (see 4S below). It has lead or Pb-Bi cooling, runs at 566°C, and has an integral steam generator inside the sealed unit that would be installed below ground level. Conceived in sizes 10-100 MWe, development is now focused on a 45 MWt/20 MWe version as part of the U.S. Generation IV effort. After a 20-year life without refuelling, the whole reactor unit is then returned for recycling of the fuel. The core is one meter in diameter and 0.8m high. SSTAR will eventually be coupled with a Brayton cycle turbine using supercritical carbon dioxide. Prototype envisaged for 2015.
Generation IV fast reactors
In 2003 the Generation IV International Forum (GIF) representing ten countries announced the selection of six reactor technologies that they believe represent the future shape of nuclear energy. These technologies 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. 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; India has applied to join.
Most of the six systems employ a closed fuel cycle to maximize the resource base and minimize high-level wastes, which would need to be sent to a repository. Three of the six systems are fast reactors, one can be built as a fast reactor, and one is described as epithermal - these five are described below. Only two operate with slow neutrons like today's plants.
Of the five, only one is cooled by light water, one is 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 system has the uranium fuel dissolved in the circulating coolant. Temperatures range from 510ºC to 850ºC, compared with less than 330ºC in today's light water reactors, which means that three of the systems can be used for thermochemical hydrogen production.
The reactors range in size from 150 to 1500 MWe (or equivalent thermal), with the lead-cooled system optionally available as a 50-150 MWe "battery" with long core life (15-20 years without refuelling) as a replaceable cassette or entire reactor module. This system is designed for distributed generation or desalination. At least three of the five 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 Generation IV participants signed an agreement to move forward the R&D on the six technologies. The USA, Canada, France, Japan, and UK agreed to undertake joint research and exchange technical information.
While Russia is not a part of GIF, one design corresponds with Russia's under-development reactor, BREST, and Russia is now the main operator of the sodium-cooled fast reactor for electricity—another of the technologies put forward by the GIF.
Gas-cooled fast reactors
Like other helium-cooled reactors which have operated or are under development, gas-cooled fast reactors will be high-temperature units (850ºC), suitable for power generation, thermochemical hydrogen production, or other process heat. To produce 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 of the actinides would be recycled to minimize production of long-lived radioactive wastes. While General Atomics worked on the design in the 1970s (but not as a fast reactor), no such reactor has thus far been built.
Lead-cooled fast reactors
Liquid metal (Pb or Pb-Bi) cooling is by natural convection. Fuel used is of depleted uranium metal or nitride, with full actinide recycled 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 such as the SSTAR described above, to modular 300-400 MWe units and large single plants of 1400 MWe. Operating temperature of 550ºC is readily achievable, but 800ºC is envisaged with advanced materials, enabling 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 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.
Sodium-cooled fast reactors
This builds on more than 300 reactor-years experience with fast neutron reactors over five decades and in eight countries. It utilizes depleted uranium fuel and has a coolant temperature of 550ºC, enabling electricity generation via a secondary sodium circuit, the primary one being at near atmospheric pressure. Two variants were 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. In light of the 2006 announcement by the U.S. Department of Energy to launch the Global Nuclear Energy Partnership (GNEP), with aims to "expand safe, clean, reliable, affordable nuclear energy worldwide", the latter course seems less likely.
Supercritical water-cooled reactors
This is a very high-pressure water-cooled reactor which 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 510-550ºC) 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 of supercritical water-cooled reactors has been in Japan.
Molten salt reactors
While not strictly a fast neutron reactor, 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. Fission products are removed continuously and the actinides are fully recycled, while plutonium and other actinides can be added along with 238U. Coolant temperature is 700ºC at very low pressure, with 800ºC envisaged. A secondary coolant system is used for electricity generation, and thermochemical hydrogen production is also feasible.
During the 1960s, the U.S. 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. There are several attractive features of the MSR fuel cycle, including: 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 238U per billion kWh); and enhanced safety due to passive cooling in any size reactor.
INPRO
Along with GIF, another program with similar aims is being coordinated by the International Atomic Energy Agency (IAEA)—the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). INPRO was launched in 2001 and has 22 members including Russia. The program aims "to support the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century." Efforts to reach these aims involve examining issues related to the development and deployment of Innovative Nuclear Energy Systems (INS) for sustainable energy supplys.
One of the case studies in phase 1 of INPRO was undertaken by Russia on its BN-800 fast reactor, though the emphasis was on methodology rather than technology. Nevertheless, fast reactor systems will feature further INPRO developments.
Global Nuclear Energy Partnership (GNEP)
This concept, announced in 2006, builds on earlier US work with the Integral Fast Reactor (IFR) project and international work on fast reactors. GNEP's primary aim is to counter proliferation concerns, but will have the effect of much greater resource utilization.
GNEP envisages fabrication and leasing of fuel for conventional reactors, with the used fuel being returned to fuel supplier countries and pyro-processed to recover uranium and actinides, leaving only fission products as high-level waste. The actinide mix is then burned in on-site fast reactors.
Physics of fast neutron reactors
In an idealized Fast Neutron Reactor, the fuel in the core is Pu-239 and the abundant neutrons designed to leak from the core would breed more Pu-239 in the fertile blanket of 238U around the core. A minor fraction of 238U might be subject to fission, but most of the neutrons reaching the 238U blanket will have lost some of their original energy and are therefore subject only to capture and the eventual generation of Pu-239. Cooling of the fast reactor core requires a heat transfer medium which has minimal moderation of the neutrons, and hence liquid metals are used, typically sodium or a mixture of sodium and potassium.
Such reactors are more efficient at converting fertile material than ordinary thermal reactors because of the arrangement of fissile and fertile materials, and there is some advantage from the fact that Pu-239 yields more neutrons per fission than 238U. Although both yield more neutrons per fission when split by fast rather than slow neutrons, this is incidental since the fission cross-sections are much smaller at high neutron energies. Fast neutron reactors may be designed as breeders to yield more fissile material than they consume or to be plutonium burners to dispose of excess plutonium. A plutonium burner would be designed without a breeding blanket, and simply with a core optimized for plutonium fuel.
Further Reading
- WNA paper on Fast neutron reactors
- American Nuclear Society, 2005. Fast Reactor Technology: A Path to Long-Term Energy Sustainability (Position Paper), November 2005. (PDF)
- Global Nuclear Energy Partnership (GNEP) Website
- International Atomic Energy Agency. Fast Reactor Database.
- The Generation IV International Forum (GIF) Website
- U.S. Department of Energy. Generation IV Nuclear Energy Systems: What is Generation IV?.
