A fast neutron reactor or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons. Such a reactor needs no neutron moderator, but must use fuel that is relatively rich in fissile material when compared to that required for a thermal reactor.
- 1 Introduction
- 2 Advantages
- 3 Disadvantages
- 4 Nuclear reactor design
- 5 History
- 6 List of fast reactors
- 7 See also
- 8 References
- 9 External links
Neutrons released in fission events typically have energies greater than 1 MeV. The fission cross-sections for fissile materials, such as U-235, are greatest at much lower energies around 1 eV. The majority of nuclear reactors in operation are known as thermal reactors, which slow the high-energy ('fast') neutrons down to low ('thermal') energies, which are in thermal equilibrium with the reactor materials. This is achieved through elastic scattering of neutrons in a moderator. In a fast reactor, this process is avoided by eliminating any moderator and the fuel consists of materials with relatively large high-energy fission cross sections.
Actinides and fission products by half-life
|Actinides by decay chain||Half-life
|Fission products of 235U by yield|
No fission products
...nor beyond 15.7M years
Legend for superscript symbols
Fast neutron reactors can reduce the total radiotoxicity of nuclear waste, and dramatically reduce the waste's lifetime. They can also use all or almost all of the fuel in the waste. Fast neutrons have an advantage in the transmutation of nuclear waste. With fast neutrons, the ratio between splitting and the capture of neutrons of plutonium or minor actinide is often larger than when the neutrons are slower, at thermal or near-thermal "epithermal" speeds. The transmuted odd-numbered actinides (e.g. from Pu-240 to Pu-241) split more easily. After they split, the actinides become a pair of "fission products." These elements have less total radiotoxicity. Since disposal of the fission products is dominated by the most radiotoxic fission product, Cesium 137, which has a half life of 30.1 years, the result is to reduce nuclear waste lifetimes from tens of millennia (from transuranic isotopes) to a few centuries. The processes are not perfect, but the remaining transuranics are reduced from a significant problem to a tiny percentage of the total waste, because most transuranics can be used as fuel.
- Fast reactors technically solve the "fuel shortage" argument against uranium-fueled reactors without assuming unexplored reserves, or extraction from dilute sources such as ordinary granite or the ocean. They permit nuclear fuels to be bred from almost all the actinides, including known, abundant sources of depleted uranium and thorium, and light water reactor wastes. On average, more neutrons per fission are produced from fissions caused by fast neutrons than from those caused by thermal neutrons. This results in a larger surplus of neutrons beyond those required to sustain the chain reaction. These neutrons can be used to produce extra fuel, or to transmute long half-life waste to less troublesome isotopes, such as was done at the Phénix reactor in Marcoule in France, or some can be used for each purpose. Though conventional thermal reactors also produce excess neutrons, fast reactors can produce enough of them to breed more fuel than they consume. Such designs are known as fast breeder reactors.
- The fast reactor doesn't just transmute the inconvenient even-numbered transuranic elements (notably Pu-240 and U-238). It transmutes them, and then fissions them for power, so that these former wastes would actually become valuable.
- Breeder reactors are costly to build and operate, and are not likely to be cost-competitive with thermal reactors unless the price of uranium increases dramatically.
- Due to the low cross sections of most materials at high neutron energies, critical mass in a fast reactor is much higher than a thermal reactor. In practice, this means significantly higher enrichment: >20% enrichment in a fast reactor compared to <5% enrichment in typical thermal reactors. This raises greater Nuclear proliferation and nuclear security issues.
- Sodium is often used as a coolant in fast reactors, because it does not moderate neutron speeds much and has a high heat capacity. However, it burns and foams in air. It has caused difficulties in reactors (e.g. USS Seawolf (SSN-575), Monju), although some sodium-cooled fast reactors have operated safely (notably the Superphénix and EBR-II for 30 years).
- Since liquid metals have low moderating power and ratio and no other moderator is present, the primary interaction of neutrons with liquid metal coolants is the (n,gamma) reaction, which induces radioactivity in the coolant. Boiling in the coolant, e.g. in an accident, would reduce coolant density and thus the absorption rate, such that the reactor has a positive void coefficient, which is dangerous and undesirable from a safety and accident standpoint. This can be avoided with a gas cooled reactor, since voids do not form in such a reactor during an accident; however, activation in the coolant remains a problem. A helium-cooled reactor would avoid this, since the elastic scattering and total cross sections are approximately equal, i.e. there are very few (n,gamma) reactions in the coolant and the low density of helium at typical operating conditions means that the amount neutrons have few interactions with coolant.
Nuclear reactor design
Water, the most common coolant in thermal reactors, is generally not a feasible coolant for a fast reactor, because it acts as a neutron moderator. However the Generation IV reactor known as the supercritical water reactor with decreased coolant density may reach a hard enough neutron spectrum to be considered a fast reactor. Breeding, which is the primary advantage of fast over thermal reactors, may be accomplished with a thermal, light-water cooled & moderated system using very high enriched (~90%) uranium.
All current fast reactors are liquid metal cooled reactors. The early Clementine reactor used mercury coolant and plutonium metal fuel. Sodium-potassium alloy (NaK) coolant is popular in test reactors due to its low melting point. In addition to its toxicity to humans, mercury has a high cross section for the (n,gamma) reaction, causing activation in the coolant and losing neutrons that could otherwise be absorbed in the fuel, which is why it is no longer used or considered as a coolant in reactors. Molten lead cooling has been used in naval propulsion units as well as some other prototype reactors. All large-scale fast reactors have used molten sodium coolant.
Another proposed fast reactor is a Molten Salt Reactor, one in which the molten salt's moderating properties are insignificant. This is typically achieved by replacing the light metal fluorides (e.g. LiF, BeF2) in the salt carrier with heavier metal chlorides (e.g., KCl, RbCl, ZrCl4).
Gas-cooled fast reactors have been the subject of research as well, as helium, the most commonly proposed coolant in such a reactor, has small absorption and scattering cross sections, thus preserving the fast neutron spectrum without significant neutron absorption in the coolant.
In practice, sustaining a fission chain reaction with fast neutrons means using relatively highly enriched uranium or plutonium. The reason for this is that fissile reactions are favored at thermal energies, since the ratio between the Pu239 fission cross section and U238 absorption cross section is ~100 in a thermal spectrum and 8 in a fast spectrum. Fission and absorption cross sections are low for both Pu239 and U238 at high (fast) energies, which means that fast neutrons are likelier to pass through fuel without interacting than thermal neutrons; thus, more fissile material is needed. Therefore it is impossible to build a fast reactor using only natural uranium fuel. However, it is possible to build a fast reactor that will breed fuel (from fertile material) by producing more fissile material than it consumes. After the initial fuel charge such a reactor can be refueled by reprocessing. Fission products can be replaced by adding natural or even depleted uranium with no further enrichment required. This is the concept of the fast breeder reactor or FBR.
So far, most fast neutron reactors have used either MOX (mixed oxide) or metal alloy fuel. Soviet fast neutron reactors have been using (high U-235 enriched) uranium fuel. The Indian prototype reactor has been using uranium-carbide fuel.
While criticality at fast energies may be achieved with uranium enriched to 5.5 weight percent Uranium-235, fast reactor designs have often been proposed with enrichments in the range of 20 percent for a variety of reasons, including core lifetime: If a fast reactor were loaded with the minimal critical mass, then the reactor would become subcritical after the first fission had occurred. Rather, an excess of fuel is inserted with reactivity control mechanisms, such that the reactivity control is inserted fully at the beginning of life to bring the reactor from supercritical to critical; as the fuel is depleted, the reactivity control is withdrawn to mitigate the negative reactivity feedback from fuel depletion and fission product poisons. In a fast breeder reactor, the above applies, though the reactivity from fuel depletion is also compensated by the breeding of either Uranium-233 or Plutonium-239 and 241 from Thorium 232 or Uranium 238, respectively.
They cannot, however, rely on changes to their moderators because there is no moderator. So Doppler broadening in the moderator, which affects thermal neutrons, does not work, nor does a negative void coefficient of the moderator. Both techniques are very common in ordinary light water reactors.
Doppler broadening from the molecular motion of the fuel, from its heat, can provide rapid negative feedback. The molecular movement of the fissionables themselves can tune the fuel's relative speed away from the optimal neutron speed. Thermal expansion of the fuel itself can also provide quick negative feedback. Small reactors such as those used in submarines may use doppler broadening or thermal expansion of neutron reflectors.
during the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power.
List of fast reactors
Fast reactors of the past
- CLEMENTINE, the first fast reactor, built in 1946 at Los Alamos National Laboratory. Plutonium metal fuel, mercury coolant, power 25 kW thermal, used for research, especially as a fast neutron source.
- EBR-I at Idaho Falls, which in 1951 became the first reactor to generate significant amounts of electrical power. Decommissioned 1964.
- Fermi 1 near Detroit was a prototype fast breeder reactor that began operating in 1957 and shut down in 1972.
- EBR-II Prototype for the Integral Fast Reactor, 1965–1995?.
- SEFOR in Arkansas, a 20 MWt research reactor which operated from 1969 to 1972.
- Fast Flux Test Facility, 400MWt, Operated flawlessly from 1982 to 1992, at Hanford Washington, now deactivated, liquid sodium is drained with argon backfill under care and maintenance.
- DFR (Dounreay Fast Reactor, 1959–1977, 14MWe) and PFR (Prototype Fast Reactor, 1974–1994, 250MWe), in Caithness, in the Highland area of Scotland.
- Rhapsodie in Cadarache, France, (20 then 40 MW) between 1967 and 1982.
- Superphénix, in France, 1200MWe, closed in 1997 due to a political decision and very high costs of operation.
- Phénix, 1973, France, 233 MWe, restarted 2003 at 140 MWe for experiments on transmutation of nuclear waste for six years, ceased power generation in March 2009, though it will continue in test operation and to continue research programs by CEA until the end of 2009. Stopped in 2010.
- KNK-II, Germany
- Small lead-cooled fast reactors used for naval propulsion, particularly by the Soviet Navy.
- BR-5 - research fast neutron reactor at the Institute of Physics and Energy in Obninsk. Years of operation 1959-2002.
- BN-350, constructed by the Soviet Union in Shevchenko (today's Aqtau) on the Caspian Sea, 130MWe plus 80,000 tons of fresh water per day.
- IBR-2 - research fast neutron reactor at the Joint Institute of Nuclear Research in Dubna (near Moscow).
- BN-600 - sodium-cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. Provides 560 MW to the Middle Urals power grid. In operation since 1980.
- BN-800 - sodium-cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. Designed to generate 880MW of electrical power. Started producing electricity in October, 2014. BN-800 reactor design is to be sold by Russia to the China.
- Clinch River Breeder Reactor, USA
- Integral Fast Reactor, USA. Design emphasized fuel cycle based on on-site electrolytic reprocessing. Cancelled 1994 without construction.
- SNR-300, Germany
- Monju reactor, 300 MWe, in Japan. was closed in 1995 following a serious sodium leak and fire. It was restarted May 6, 2010 and in August 2010 another accident, involving dropped machinery, shut down the reactor again. As of June 2011, the reactor has only generated electricity for one hour since its first testing two decades prior.
- BN-600, 1981, Russia, 600 MWe, scheduled end of life 2010 but still in operation.
- BN-800, Russia, testing began June 27, 2014, estimated total power 880MW
- BOR-60 - sodium-cooled reactor at the Research Institute of Atomic Reactors in Dmitrovgrad. In operation since 1980.(experimental purposes)
- FBTR, 1985, India, 10.5 MWt (experimental purposes)
- China Experimental Fast Reactor, 65 MWt (experimental purposes), planned 2009, critical 2010
- Jōyō (常陽?), 1977–1997 and 2004–2007, Japan, 140 MWt. Experimental reactor, operated as an irradiation test facility. After an incident in 2007, the reactor is suspended for repairing, recovery works were planned to be completed in 2014.
- PFBR, Kalpakkam, India, 500 MWe.
In design phase
- BN-1200, Russia, build starting after 2014, operation in 2018–2020
- Toshiba 4S being developed in Japan and was planned to be shipped to Galena, Alaska (USA) but progress is stalled (see Galena Nuclear Power Plant)
- KALIMER, 600 MWe, South Korea, projected 2030. KALIMER is a continuation of the sodium cooled, metallic fueled, fast neutron reactor in a pool represented by the Advanced Burner Reactor (2006), S-PRISM (1998-present), Integral Fast Reactor (1984-1994), and EBR-II (1965-1995).
- Generation IV reactor (Helium·Sodium·Lead cooled) US-proposed international effort, after 2030
- JSFR, Japan, project for a 1500 MWe reactor began in 1998, but without success.
- ASTRID, France, project for a 600 MWe sodium-cooled reactor. Planned experimental operation in 2020.
|Past||Clementine, EBR-I/II, SEFOR, FFTF||BN-350||Dounreay, Rhapsodie, Superphénix, Phénix (stopped in 2010)|
|Cancelled||Clinch River, IFR||SNR-300|
|Under construction||Monju, PFBR,|
|Planned||Gen IV (Gas·Sodium·Lead)||BN-1200||ASTRID||4S, JSFR, KALIMER|
- Nuclear fuel cycle
- Fast breeder reactor
- Sodium-cooled fast reactor
- Lead-cooled fast reactor
- Gas-cooled fast reactor
- Generation IV reactor
- Energy amplifier
- Thermal-neutron reactor
- Plus radium (element 88). While actually a sub-actinide, it immediately precedes actinium (89) and follows a three element gap of instability after polonium (84) where no isotopes have half-lives of at least four years (the longest-lived isotope in the gap is radon-222 with a half life of less than four days). Radium's longest lived isotope, at 1600 years, thus merits the element's inclusion here.
- Specifically from thermal neutron fission of U-235, e.g. in a typical nuclear reactor.
- Milsted, J.; Friedman, A. M.; Stevens, C. M. (1965). "The alpha half-life of berkelium-247; a new long-lived isomer of berkelium-248". Nuclear Physics 71 (2): 299. doi:10.1016/0029-5582(65)90719-4.
"The isotopic analyses disclosed a species of mass 248 in constant abundance in three samples analysed over a period of about 10 months. This was ascribed to an isomer of Bk248 with a half-life greater than 9 y. No growth of Cf248 was detected, and a lower limit for the β− half-life can be set at about 104 y. No alpha activity attributable to the new isomer has been detected; the alpha half-life is probably greater than 300 y."
- This is the heaviest isotope with a half-life of at least four years before the "Sea of Instability".
- Excluding those "classically stable" isotopes with half-lives significantly in excess of 232Th, e.g. while 113mCd has a half-life of only fourteen years, that of 113Cd is nearly eight quadrillion.
- Smarter use of Nuclear Waste, by William H. Hannum, Gerald E. Marsh and George S. Stanford, Copyright Scientific American, 2005. Retrieved 2010-9-2.
- "Fast Breeder Reactor Programs: History and Status". International Panel on Fissile Materials. February 2010.
- "Fast Reactor Knowledge Preservation System: Taxonomy and Basic Requirements".
- "Beloyarsk Nuclear Power Plant".
-  Beloyarsk NPP website
- Fast reactor starts clean nuclear energy era in Russia
- China 's first Experimental Fast Reactor (CEFR) Put into Operation in 2009 – Zoom China Energy Intelligence-New site
- T. SOGA, W. ITAGAKI, Y. KIHARA, Y. MAEDA. Endeavor to improve in-pile testing techniques in the experimental fast reactor Joyo. 2013.
- ANL report on EARLY SOVIET FAST REACTORS
- Article on recent work on fast neutron reactors in Scientific American, December, 2005
- IAEA Fast Reactor Database
- Fast Reactor Data Retrieval and Knowledge Preservation seeks to establish a comprehensive, international inventory of fast reactor data and knowledge, which would be sufficient to form the basis for fast reactor development in 30 to 40 years from now.
- World Nuclear Association: Fast Neutron Reactors
- International Thorium Energy Organisation - www.IThEO.org