Fast-neutron reactor

From Wikipedia, the free encyclopedia
Jump to navigation Jump to search
Shevchenko BN350 nuclear fast reactor and desalination plant situated on the shore of the Caspian Sea. The plant generated 135 MWe and provided steam for an associated desalination plant. View of the interior of the reactor hall.

A fast-neutron reactor (FNR) or fast-spectrum reactor or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons (carrying energies above 1 MeV or greater, on average), as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.

Use of a moderator in nuclear reactors[edit]

Natural uranium consists mostly of two isotopes: 238
, 235
. 238
accounts for roughly 99.3% of natural uranium and undergoes fission only by fast neutrons.[1] About 0.7% of natural uranium is 235
, which will fission by neutrons. When either of these isotopes undergoes fission, it releases neutrons with a high energy ("Fast"). The flux of higher-energy fission neutrons (> 2 MeV) is too low to create sufficient fission in 238
, and the flux of lower-energy fission neutrons (< 2 MeV) is too low to do so easily in 235

The common solution to this problem is to slow the neutrons down using a neutron moderator, which interacts with the neutrons to slow them ("thermal" neutrons). The most common moderator is ordinary water, which acts by elastic scattering until the neutrons reach thermal equilibrium with the water, at which point the neutrons become highly reactive with the 235
. Other moderators include graphite. Moderation can be likened to a billiard ball (the neutron) striking another, slightly heavier billiard ball (water); the kinetic energy of the neutron is transferred in part to the heavier water. Many such collisions will slow the neutron down until it has about the same speeds as water molecules, the vibrations of which are observed on the macroscopic scale as temperature, hence the name "thermal". Thus, the slower neutrons have a much bigger chance (about a thousand times) of causing a fission in 235
than the faster neutrons. These thermal neutrons are also about a thousand times more likely to be absorbed by another heavy element, such as 238

Although 238
does not undergo fission by the neutrons released in fission, thermal neutrons (i.e. neutrons that have been slowed down by a moderator) can be captured by the nucleus to transmute the uranium into 239
. 239
has a neutron cross section similar to that of 235
. Unfortunately, only about 40% of the 239
created this way will undergo fission from the thermal neutrons. When 239
absorbs another neutron without undergoing fission, 240
is created, which virtually never fissions with the slower neutrons, when it in turn absorbs one, but just absorbs the neutron to become a heavier isotope. These effects combined have the result of creating, in a (water) moderated reactor, the presence of the transuranic elements. Such isotopes are themselves often not stable, and undergo Beta decay to create ever heavier elements, such as Americium and Curium. Thus, plutonium isotopes in many instances do not fission (and so do not create new neutrons), but instead just absorb the neutrons. Therefore, after a certain amount of time (around 12-18 months of stable operation in water moderated reactors), the nuclear reactor can no longer sustain the fission process, and the reactor has to be refueled.

In the spent fuel, therefore, several plutonium isotopes are present, along with the heavier, transuranic elements. Nuclear reprocessing, a complex series of chemical extraction processes, can then be used to extract the unchanged uranium, the fission products, the plutonium, and the heavier elements.[3]

Water has disadvantages as a moderator. It can absorb a neutron and remove it from the reaction. It does this just enough that the concentration of 235
in natural uranium is too low to sustain the chain reaction; the neutrons lost through absorption in the water and 238
, along with those lost to the environment, results in too few left in the fuel. The most common solution to this problem is to slightly concentrate the amount of 235
in the fuel to produce enriched uranium, with the leftover 238
known as depleted uranium. Other designs use different moderators, like heavy water, that are much less likely to absorb neutrons, allowing them to run on unenriched fuel. In either case, the reactor's neutron economy is based on thermal neutrons. Another drawback of using water is that it has a relatively low boiling point. The vast majority of electricity production uses steam turbines. These become more efficient as the pressure (and thus the temperature) of the steam is as high as possible. A water cooled and moderated nuclear reactor therefore needs to operate at high pressures (around 70 bars) to enable the efficient production of electricity. Thus, such reactors are constructed using very heavy steel vessel, that are for example 22 cm (10 inch) thick. This adds complexity to reactor design and requires safety measures. The vast majority of nuclear reactors in the world is water cooled and moderated with water.

Fast fission, breeders[edit]

Although 235
and 239
are less sensitive to higher-energy neutrons, they still remain somewhat reactive well into the MeV range. If enough 235
or 239
is present, eventually a threshold will be reached where there are enough fissile atoms in the fuel to maintain a chain reaction even with fast neutrons. Crucially, when a reactor runs on fast neutrons, the plutonium isotopes are far more likely to fission upon absorbing a neutron. Thus, fast neutrons have a smaller chance of being captured by the uranium and plutonium, but when they are captured, have a much bigger chance of causing a fission.

By removing the moderator, the size of the reactor can be greatly reduced, and to some extent the complexity. As 239
and particularly 240
are far more likely to fission when they capture a neutron, it is possible to fuel such reactors with a mixture of plutonium and natural uranium. During the operation, the natural uranium (mostly 238
will be turned into plutonium, or fission in the first place, and thus the reactor never runs out of neutrons, as new fuel is created during the operation, a process called breeding. Fast reactors can be used for breeding. By surrounding the reactor core with a blanket of 238
which captures the neutrons, the extra neutrons breed more 239

The blanket material can then be processed to extract the 239
, which is then mixed with uranium to produce MOX fuel that can be fed into both conventional slow-neutron reactors, as well as fast reactors. A single fast reactor can thereby feed several slow ones, greatly increasing the amount of energy extracted from the natural uranium: from less than 1% in a normal once-through cycle, to as much as 60% in the best existing fast reactor cycles. Given the current inventory of spent nuclear fuel (which contains the plutonium), it is possible to treat this waste material and reuse the fuel in fast reactors, effectively destroying the plutonium.

As the perception of the reserves of uranium ore in the 1960s was rather low, and the rate that nuclear power was expected to take over baseload generation, through the 1960s and 1970s fast breeder reactors were considered to be the solution to the world's energy needs. Using twice-through processing, a fast breeder increases the energy capacity of known ore deposits, meaning that existing ore sources would last hundreds of years. The disadvantage to this approach is that the breeder reactor has to be fed fuel that must be treated in a spent fuel treatment plant. It was widely expected that this would still be below the price of enriched uranium as demand increased and known resources dwindled.

Through the 1970s, experimental breeder designs were examined, especially in the US, France and the USSR. However, this coincided with a crash in uranium prices. The expected increased demand led mining companies to expand supply channels, which came online just as the rate of reactor construction stalled in the mid-1970s. The resulting oversupply caused fuel prices to decline from about US$40 per pound in 1980 to less than $20 by 1984. Breeders produced fuel that was much more expensive, on the order of $100 to $160, and the few units that reached commercial operation proved to be economically unfeasible.

US interest in breeder reactors were further muted by Jimmy Carter's April 1977 decision to defer construction of breeders in the US due to proliferation concerns, and the suboptimal operating record of France's Superphénix reactor. [4] The French reactors also met with serious opposition of environmentalist groups, who regarded these as very dangerous. [5]

Despite such setbacks, a large number of countries is still invested in the fast reactor technology. Around 25 reactors have been built since the 1970-ies, accumulating over 400 reactor years of experience. Russia currently operates two fast reactors [6]on commercial scale, which shows that the technology is viable, and international interest is strong. The GEN IV initiative, an international working group on new reactor designs has proposed six new reactor types, three of which operate with a fast spectrum. Argentina, Brazil, Canada, France, Japan, the Republic of Korea, the Republic of South Africa, the United Kingdom and the United States are part of this working group.[7]


Actinides[8] by decay chain Half-life
range (a)
Fission products of 235U by yield[9]
4n 4n+1 4n+2 4n+3
4.5–7% 0.04–1.25% <0.001%
228Ra 4–6 a 155Euþ
244Cmƒ 241Puƒ 250Cf 227Ac 10–29 a 90Sr 85Kr 113mCdþ
232Uƒ 238Puƒ 243Cmƒ 29–97 a 137Cs 151Smþ 121mSn
248Bk[10] 249Cfƒ 242mAmƒ 141–351 a

No fission products
have a half-life
in the range of
100 a–210 ka ...

241Amƒ 251Cfƒ[11] 430–900 a
226Ra 247Bk 1.3–1.6 ka
240Pu 229Th 246Cmƒ 243Amƒ 4.7–7.4 ka
245Cmƒ 250Cm 8.3–8.5 ka
239Puƒ 24.1 ka
230Th 231Pa 32–76 ka
236Npƒ 233Uƒ 234U 150–250 ka 99Tc 126Sn
248Cm 242Pu 327–375 ka 79Se
1.53 Ma 93Zr
237Npƒ 2.1–6.5 Ma 135Cs 107Pd
236U 247Cmƒ 15–24 Ma 129I
244Pu 80 Ma

... nor beyond 15.7 Ma[12]

232Th 238U 235Uƒ№ 0.7–14.1 Ga

Legend for superscript symbols
₡  has thermal neutron capture cross section in the range of 8–50 barns
ƒ  fissile
metastable isomer
№  primarily a naturally occurring radioactive material (NORM)
þ  neutron poison (thermal neutron capture cross section greater than 3k barns)
†  range 4–97 a: Medium-lived fission product
‡  over 200 ka: Long-lived fission product

Fast-neutron reactors can reduce the total radiotoxicity of nuclear waste [13] using all or almost all of the waste as fuel. With fast neutrons, the ratio between splitting and the capture of neutrons by plutonium and the minor actinides is often larger than when the neutrons are slower, at thermal or near-thermal "epithermal" speeds. Simply put, fast neutrons have a smaller chance of being absorbed by plutonium or Uranium, but when they do, they almost always cause a fission. The transmuted even-numbered actinides (e.g. 240
, 242
) split nearly as easily as odd-numbered actinides in fast reactors. 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 products, strontium-90, which has a half life of 28.8 years, and caesium-137, which has a half life of 30.1 years,[13] 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 undiscovered reserves, or extraction from dilute sources such as granite or seawater. 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 by fast neutrons than from 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, as was done at the Phénix reactor in Marcoule, 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.[citation needed]

Another advantage is the inherent safety of the reactors. Since the most common fast reactor design is the sodium cooled "pool type" reactor, in which the reactor is submerged in a very large volume of molten sodium metal, there is a very large thermal inertia, which can absorb temperature fluctuations. Second, because the sodium has a boiling point of 883 oC, but the reactor operates typically around 530 - 550 oC, there is a large margin where sodium will stay liquid, and thermal increases can be easily absorbed. Third, because there is no water present at high temperatures, the reactor is essentially at atmospheric pressure. Most often, an inert gas blanket at a slight overpressure (e.g. 0.5 atmospheres) is present to ensure that any leak results in mass transport to the outside of the reactor. This means that there is no pressure vessel with associated problems (high pressure systems are complex). Fourth, because the entire vessel is at atmospheric pressure, and the sodium is very hot in operation, passive cooling (i.e. no pumping requirements) is possible. Accidents such as the Fukushima Daiichi nuclear accident [14] are impossible with such a design. Lastly, the higher temperature of the sodium, and therefore the higher temperature of the steam generated by this sodium, allows a considerable increase in the electric generating efficiency (around 40%). [15]


The main disadvantage of fast-neutron reactors is that to date they have proven costly to build and operate, and none have been proven cost-competitive with thermal-neutron reactors unless the price of uranium increased dramatically.[16]

Some other disadvantages are specific to some designs.

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 for long periods (notably the Phénix and EBR-II for 30 years, or the BN-600 still in operation since 1980 despite several minor leaks and fires. It is important to note that these fires had no radioactive releases, as the sodium fast reactors are always designed with a two loop system.

Another problem is related to neutron activation. Since liquid metals other than lithium and beryllium have low moderating ability, the primary interaction of neutrons with fast reactor coolant is the (n,gamma) reaction, which induces radioactivity in the coolant. Neutron irradiation activates a significant fraction of coolant in high-power fast reactors, up to around a terabecquerel of beta decays per kilogram of coolant in steady operation.[17] This is the reason that sodium-cooled reactors have a primary cooling loop embedded within a separate sodium pool. The sodium-24 that results from neutron capture undergoes beta decay to magnesium-24 with a half life of fifteen hours; the magnesium is removed in a cold trap.

A defective fast reactor design could have positive void coefficient: boiling of the coolant in an accident would reduce coolant density and thus the absorption rate; no such designs are proposed for commercial service. This 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 both problems, since the elastic scattering and total cross sections are approximately equal, i.e. few (n,gamma) reactions are present in the coolant and the low density of helium at typical operating conditions means that neutrons have few interactions with coolant.[citation needed]

Due to the low cross sections of most materials at high neutron energies, critical mass in a fast reactor is much higher than in a thermal reactor. In practice, this means significantly higher enrichment: >20% enrichment in a fast reactor compared to <5% enrichment in typical thermal reactors.

Reactor design[edit]


All nuclear reactors produce heat which must be removed from the reactor core. Water, the most common coolant in thermal reactors, is generally not feasible for a fast reactor, because it acts as a neutron moderator.

All operating fast reactors are liquid metal cooled reactors. The early Clementine reactor used mercury coolant and plutonium metal fuel. In addition to its toxicity to humans, mercury has a high cross section (thus, it readily absorbs the radiation, which causes nuclear reactions) 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 considered as a coolant.

Russia has developed reactors that use Molten lead and lead-bismuth eutectic alloys, which have been used on a larger scale in naval propulsion units, particularly the Soviet Alfa-class submarine, as well as some prototype reactors. Sodium-potassium alloy (NaK) is popular in test reactors due to its low melting point.

Another proposed fast reactor is a molten salt reactor, in which the salt's moderating properties are insignificant. [18]

Gas-cooled fast reactors have been the subject of research commonly using helium, which has small absorption and scattering cross sections, thus preserving the fast neutron spectrum without significant neutron absorption in the coolant.[citation needed]

However, all large-scale fast reactors have used molten sodium coolant. Advantages of molten sodium are its low cost, the small activation potential and its large liquid range. The latter means that the material has a low melting point, and a high boiling point. These reactors are called the Sodium cooled fast reactor and are still being pursued worldwide. Russia currently operates two such reactors on a commercial scale.


In practice, sustaining a fission chain reaction with fast neutrons means using relatively enriched uranium or plutonium. The reason for this is that fissile reactions are favored at thermal energies, since the ratio between the 239
fission cross section and 238
absorption cross section is ~100 in a thermal spectrum and 8 in a fast spectrum. Fission and absorption cross sections are low for both 239
and 238
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 a fast reactor cannot run on natural uranium fuel. However, it is possible to build a fast reactor that breeds fuel by producing more 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 without further enrichment. 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 use (high 235
enriched) uranium fuel. The Indian prototype reactor uses 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 been proposed with enrichments in the range of 20 percent for 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. 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 support continuing fission. In a fast breeder reactor, the above applies, though the reactivity from fuel depletion is also compensated by breeding either 233
or 239
and 241
from thorium-232 or 238
, respectively.


Like thermal reactors, fast-neutron reactors are controlled by keeping the criticality of the reactor reliant on delayed neutrons, with gross control from neutron-absorbing control rods or blades.

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 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 can provide negative feedback. Small reactors as in submarines may use Doppler broadening or thermal expansion of neutron reflectors.

Shevchenko BN350 desalination unit, the only nuclear-heated desalination unit in the world


A 2008 IAEA proposal for a Fast Reactor Knowledge Preservation System[19] noted that:

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[edit]

Decommissioned reactors[edit]

United States[edit]


  • Dounreay Loop type Fast Reactor (DFR), 1959–1977, was a 14 MWe and Prototype Fast Reactor (PFR), 1974–1994, 250 MWe, in Caithness, in the Highland area of Scotland.
  • Dounreay Pool type Fast Reactor (PFR), 1975–1994, was a 600 MWt, 234 MWe which used mixed oxide (MOX) fuel.
  • Rapsodie in Cadarache, France, (20 then 40 MW) operated between 1967 and 1982.
  • Superphénix, in France, 1200 MWe, closed in 1997 due to a political decision and high costs.
  • 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, in Germany a 21 MWe experimental compact sodium-cooled fast reactor operated from Oct 1977-Aug 1991. The objective of the experiment was to eliminate nuclear waste while producing energy. There were minor sodium problems combined with public protests which resulted in the closure of the facility.


  • Small lead-cooled fast reactors were used for naval propulsion, particularly by the Soviet Navy.
  • BR-5 - was a research-focused fast-neutron reactor at the Institute of Physics and Energy in Obninsk from 1959-2002.
  • BN-350 was constructed by the Soviet Union in Shevchenko (today's Aqtau) on the Caspian Sea, It produced 130 MWe plus 80,000 tons of fresh water per day.
  • IBR-2 - was a research focused fast-neutron reactor at the Joint Institute of Nuclear Research in Dubna (near Moscow).
  • RORSATs - 33 space fast reactors were launched by the Soviet Union from 1989-1990 as part of a program known as the Radar Ocean Reconnaissance Satellite (RORSAT) in the US. Typically, the reactors produced approximately 3 kWe.
  • BES-5 - was a sodium cooled space reactor launched as part of the RORSAT program which produced 5 kWe.
  • BR-5 - was a 5 MWt sodium fast reactor operated by the USSR in 1961 primarily for materials testing.
  • Russian Alpha 8 PbBi - was a series of lead bismuth cooled fast reactors used aboard submarines. The submarines functioned as killer submarines, staying in harbor then attacking due to the high speeds achievable by the sub.


  • Monju reactor, 300 MWe, in Japan, was closed in 1995 following a serious sodium leak and fire. It was restarted on May 6, 2010 but in August 2010 another accident, involving dropped machinery, shut down the reactor again. As of June 2011, the reactor had generated electricity for only one hour since its first test two decades prior.[citation needed]
  • Aktau Reactor, 150 MWe, in Kazakhstan, was used for plutonium production, desalination, and electricity. It closed 4 years after the plant's operating license expired.[citation needed]

Never operated[edit]


  • BN-600 - a pool type sodium-cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. It provides 560 MWe to the Middle Urals power grid. In operation since 1980.
  • BN-800 - a sodium-cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. It generates 880 MW of electrical power and started producing electricity in October, 2014. It reached full power in August, 2016.
  • BOR-60 - a sodium-cooled reactor at the Research Institute of Atomic Reactors in Dimitrovgrad, Russia. In operation since 1968. It produces 60MW for experimental purposes.[citation needed]
  • FBTR - a 10.5 MW experimental reactor in India which focused on reaching significant burnup levels.
  • China Experimental Fast Reactor, a 60 MWth, 20 MWe, experimental reactor which went critical in 2011 and is currently operational.[20] It is used for materials and component research for future Chinese fast reactors.
  • KiloPower/KRUSTY is a 1-10 kWe research sodium fast reactor built at Los Alamos National Laboratory. It first reach criticality in 2015 and demonstrates an application of a Stirling power cycle.

Under repair[edit]

  • Jōyō (常陽), 1977–1997 and 2004–2007, Japan, 140 MWt is an experimental reactor, operated as an irradiation test facility. After an incident in 2007, the reactor was suspended for repairing, recoworks were planned to be completed in 2014.[21]

Under construction[edit]

  • PFBR, Kalpakkam, India, 500 MWe reactor with criticality planned for 2021. It is a sodium fast breeder reactor.
  • CFR-600, China, 600 MWe.
  • MBIR Multipurpose fast neutron research reactor. The Research Institute of Atomic Reactors (NIIAR) site at Dimitrovgrad in the Ulyanovsk region of western Russia, 150 MWt. Construction started in 2016 with completion scheduled for 2024.
  • BREST-300, Seversk, Russia. Construction started at 8 June 2021[22]

In design[edit]

  • BN-1200, Russia, built starting after 2014,[23] with operation planned for 2018–2020,[24] now delayed until at least 2035.[25]
  • Toshiba 4S was planned to be shipped to Galena, Alaska (USA) but progress stalled (see Galena Nuclear Power Plant)
  • KALIME is a 600 MWe project in South Korea, projected for 2030.[26] KALIMER is a continuation of the sodium-cooled, metal-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, a project for a 1500 MWe reactor began in 1998, but without success.
  • ASTRID, France, canceled project for a 600 MWe sodium-cooled reactor.
  • Mars Atmospherically Cooled Reactor (MACR) is a 1 MWe project, planned to complete in 2033. MACR is a gas-cooled (carbon dioxide coolant) fast-neutron reactor intended to provide power to proposed Mars colonies.
  • TerraPower is designing a molten salt reactor in partnership with Southern Company, Oak Ridge National Laboratory, Idaho National Laboratory, Vanderbilt University and the Electric Power Research Institute. They expect to begin testing a loop facility in 2019 and is scaling up their salt manufacturing process. Data will be used to assess thermal hydraulics and safety analysis codes.[27]
  • Elysium Industries is designing a fast spectrum molten salt reactor.[28]
  • ALFRED (Advanced Lead Fast Reactor European Demonstrator) is a lead cooled fast reactor demonstrator designed by Ansaldo Energia from Italy, it represents the last stage of the ELSY and LEADER projects.[29]


  • Future FBR, India, 600 MWe, after 2025[30]


Fast reactors
U.S. Russia Europe Asia
Past Clementine, EBR-I/II, SEFOR, FFTF BN-350 Dounreay, Rapsodie, Superphénix, Phénix (stopped in 2010)
Cancelled Clinch River, IFR SNR-300, ASTRID
Under decommissioning Monju
Operating BOR-60, BN-600,
BN-800[citation needed]
Under repair Jōyō
Under construction MBIR, BREST-300 PFBR, CFR-600
Planned Gen IV (Gas·sodium·lead·salt), TerraPower, Elysium MCSFR, DoE VTR BN-1200 Moltex 4S, JSFR, KALIMER

See also[edit]


  1. ^ "What is Neutron - Neutron Definition". Retrieved 2017-09-19.
  2. ^ "Neutron Flux Spectra - Nuclear Power". Retrieved 2017-08-29.
  3. ^ "Fast Neutron Reactors | FBR - World Nuclear Association".
  4. ^,to%20make%20new%20nuclear%20fuel.
  5. ^
  6. ^
  7. ^ "Home - Generation IV Systems". GIF Portal.
  8. ^ 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 nuclides have half-lives of at least four years (the longest-lived nuclide in the gap is radon-222 with a half life of less than four days). Radium's longest lived isotope, at 1,600 years, thus merits the element's inclusion here.
  9. ^ Specifically from thermal neutron fission of uranium-235, e.g. in a typical nuclear reactor.
  10. ^ 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. Bibcode:1965NucPh..71..299M. 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 [years]. No growth of Cf248 was detected, and a lower limit for the β half-life can be set at about 104 [years]. No alpha activity attributable to the new isomer has been detected; the alpha half-life is probably greater than 300 [years]."
  11. ^ This is the heaviest nuclide with a half-life of at least four years before the "Sea of Instability".
  12. ^ Excluding those "classically stable" nuclides 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 years.
  13. ^ a b Smarter use of Nuclear Waste, by William H. Hannum, Gerald E. Marsh and George S. Stanford, Copyright Scientific American, 2005. Retrieved 2010-9-2,
  14. ^
  15. ^
  16. ^ "Fast Breeder Reactor Programs: History and Status" (PDF). International Panel on Fissile Materials. February 2010.
  17. ^ "Is BN-800 the best nuclear reactor for now?". January 2017.
  18. ^ "Moltex Energy | Safer Cheaper Cleaner Nuclear | Stable Salt Reactors | SSR". Retrieved 2016-10-20.
  19. ^ "Fast Reactor Knowledge Preservation System: Taxonomy and Basic Requirements" (PDF).
  20. ^ "China 's first Experimental Fast Reactor (CEFR) Put into Operation in 2009 – Zoom China Energy Intelligence-New site". Archived from the original on 2011-07-07. Retrieved 2008-06-01.
  21. ^ T. SOGA, W. ITAGAKI, Y. KIHARA, Y. MAEDA. Endeavor to improve in-pile testing techniques in the experimental fast reactor Joyo. 2013.
  22. ^ "Russia starts building lead-cooled fast reactor : New Nuclear - World Nuclear News".
  23. ^ "Решение о строительстве БН-1200 будет принято в 2014 году".
  24. ^ "В 2012 году на Белоярской АЭС начнется строительство пятого энергоблока БН-1800. РИА Новый День]". November 1, 2007. Retrieved August 2018. Check date values in: |access-date= (help)
  25. ^ "Russia defers BN-1200 until after 2035". 2 January 2020.
  26. ^ "***지속가능원자력시스템***".
  27. ^ Wang, Brian (August 24, 2018). "Southern Company partnering with Bill Gates backed Terrapower on molten chloride fast reactor". Retrieved 2018-08-25.
  28. ^
  29. ^ "Generation IV & SMR".
  30. ^ "Overview of Indian Fast Breeder Nuclear Reactor Programme - Nuclear Power - Nuclear Reactor". Scribd.

External links[edit]