The BN-350 fast-neutron reactor at Aktau, Kazakhstan. It operated between 1973 and 1994.

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, on average), as opposed to slow thermal neutrons used in thermal-neutron reactors. Such a fast 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. Around 20 land based fast reactors have been built, accumulating over 400 reactor years of operation globally. The largest was the Superphénix sodium cooled fast reactor in France that was designed to deliver 1,242 MWe. Fast reactors have been studied since the 1950s, as they provide certain advantages over the existing fleet of water-cooled and water-moderated reactors. These are:

In the GEN IV initiative, about two thirds of the proposed reactors for the future use a fast spectrum for these reasons.

Fission processes

Fast reactors operate by the fission of uranium and other heavy atoms, similar to thermal reactors. However, there are crucial differences, arising from the fact that by far most commercial nuclear reactors use a moderator, and fast reactors do not.

Moderators in conventional nuclear reactors

Natural uranium consists mostly of two isotopes:

Of these two, 238
undergoes fission only by fast neutrons.[2] About 0.7% of natural uranium is 235
, which will fission by both fast and slow (thermal) neutrons. When the uranium undergoes fission, it releases neutrons with a high energy ("fast"). However, these fast neutrons have a much lower probability of causing another fission than neutrons which are slowed down after they have been generated by the fission process. Slower neutrons have a much higher chance (about 585 times greater) of causing a fission in 235
than the fast neutrons.

The common solution to this problem is to slow the neutrons down using a neutron moderator, which interacts with the neutrons to slow them. The most common moderator is ordinary water, which acts by elastic scattering until the neutrons reach thermal equilibrium with the water (hence the term "thermal neutron"), at which point the neutrons become highly reactive with the 235
. Other moderators include heavy water, beryllium and graphite. The elastic scattering of the neutrons can be likened to the collision of two ping pong balls; when a fast ping pong ball hits one that is stationary or moving slowly, they will both end up having about half of the original kinetic energy of the fast ball. This is in contrast to a fast ping pong ball hitting a bowling ball, where the ping pong ball keeps virtually all of its energy.

Such thermal neutrons are more likely to be absorbed by another heavy element, such as 238
, 232
or 235
. In this case, only the 235
has a high probability of fission.

Although 238
undergoes fission by the fast neutrons released in fission about 11% of the time this can not sustain the chain reaction alone. Neutrons produced by fission of 238
have lower energies than the original neutron, usually below 1 MeV, the fission threshold to cause subsequent fission of 238
, so fission of 238
does not sustain a nuclear chain reaction. When hit by thermal neutrons (i.e. neutrons that have been slowed down by a moderator) the neutron can be captured by the 238
nucleus to transmute the uranium into 239
which rapidly decays into 239
which in turn decays into 239
. 239
has a thermal neutron cross section larger than that of 235

About 73% of the 239
created this way will undergo fission from capturing a thermal neutron while the remaining 27% absorbs a thermal neutron without undergoing fission, 240
is created, which rarely fissions with thermal neutrons. When plutonium-240 in turn absorbs a thermal neutron to become a heavier isotope 241
which is also fissionable with thermal neutrons very close in probability to plutonium-239. In a fast spectrum reactor all three isotopes have a high probability of fission when absorbing a high energy neutron which limits their accumulation in the fuel.

These effects combined have the result of creating, in a moderated reactor, the presence of the transuranic elements. Such isotopes are themselves unstable, and undergo beta decay to create ever heavier elements, such as americium and curium. Thus, in moderated reactors, plutonium isotopes in many instances do not fission (and so do not release new fast neutrons), but instead just absorb the thermal neutrons. Most moderated reactors use natural uranium or low enriched fuel. As power production continues, around 12–18 months of stable operation in all moderated reactors, the reactor both consumes more fissionable material than it breeds and accumulates neutron absorbing fission products which make it difficult to sustain the fission process. When too much fuel has been consumed the reactor has to be refueled.

Drawbacks of light water as the moderator in conventional nuclear reactors

The following disadvantages of the use of a moderator have instigated the research and development of fast reactors.[3]

Although cheap, readily available and easily purified, light water can absorb a neutron and remove it from the reaction. It does this 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 concentrate the amount of 235
in the fuel to produce enriched uranium, with the leftover 238
known as depleted uranium.

Other thermal neutron designs use different moderators, like heavy water or graphite that are much less likely to absorb neutrons, allowing them to run on natural uranium fuel. See CANDU, X-10 Graphite Reactor. In either case, the reactor's neutron economy is based on thermal neutrons.

A second drawback of using water for cooling 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 higher. A water cooled and moderated nuclear reactor therefore needs to operate at high pressures to enable the efficient production of electricity. Thus, such reactors are constructed using very heavy steel vessels, for example 30 cm (12 inch) thick. This high pressure operation adds complexity to reactor design and requires extensive physical safety measures. The vast majority of nuclear reactors in the world are water cooled and moderated with water. Examples include the PWR, the BWR and the CANDU reactors. In Russia and the UK, reactors are operational that use graphite as moderator, and respectively water in Russian and gas in British reactors as coolant.

As the operational temperature and pressure of these reactors is dictated by engineering and safety constraints, both are limited. Thus, the temperatures and pressures that can be delivered to the steam turbine are also limited. Typical water temperatures of a modern pressurized water reactor are around 350 °C (660 °F), with pressures of around 85 bar (1233 psi). Compared to for example modern coal fired steam circuits, where main steam temperatures in excess of 500 °C (930 °F) are obtained, this is low, leading to a relatively low thermal efficiency. In a modern PWR, around 30-33 % of the nuclear heat is converted into electricity.

A third drawback is that when a (any) nuclear reactor is shut down after operation, the fuel in the reactor no longer undergoes fission processes. However, there is an inventory present of highly radioactive elements, some of which generate substantial amounts of heat. If the fuel elements were to be exposed (i.e. there is no water to cool the elements), this heat is no longer removed. The fuel will then start to heat up, and temperatures can then exceed the melting temperature of the zircaloy cladding. When this occurs the fuel elements melt, and a meltdown occurs, such as the multiple meltdowns that occurred in the Fukushima disaster. When the reactor is in shutdown mode, the temperature and pressure are slowly reduced to atmospheric, and thus water will boil at 100 °C (210 °F). This relatively low temperature, combined with the thickness of the steel vessels used, could lead to problems in keeping the fuel cool, as was shown by the Fukushima accident.

Lastly, the fission of uranium and plutonium in a thermal spectrum yields a smaller number of neutrons than in the fast spectrum, so in a fast reactor, more losses are acceptable.

The proposed fast reactors solve all of these problems (next to the fundamental fission properties, where for example plutonium-239 is more likely to fission after absorbing a fast neutron, than a slow one.)

Fast fission and breeding

Although 235
and 239
have a lower capture cross section with higher-energy neutrons, they still remain reactive well into the MeV range. If the density of 235
or 239
is sufficient, a threshold will be reached where there are enough fissile atoms in the fuel to maintain a chain reaction with fast neutrons. In fact, in the fast spectrum, when 238
captures a fast neutron it will also undergo fission around 11% of the time with the remainder of captures being "radiative" and entering the decay chain to plutonium-239.

Crucially, when a reactor runs on fast neutrons, the 239
isotope is likely to fission 74% of the time instead of the 62% of fissions when it captures a thermal neutron. In addition the probability of a 240
upon absorbing a fast neutron fissioning is 70% while for a thermal neutron it is less than 20%. Fast neutrons have a smaller chance of being captured by the uranium and plutonium, but when they are captured, have a significantly higher probability of causing a fission. The inventory of spent fast reactor fuel therefore contains virtually no actinides except for uranium and plutonium, which can be effectively recycled. Even when the core is initially loaded with 20% mass reactor-grade plutonium (containing on average 2% 238
, 53% 239
, 25% 240
, 15% 241
, 5% 242
and traces of 244
), the fast spectrum neutrons are capable of causing each of these to fission at significant rates. By the end of a fuel cycle of some 24 months, these ratios will have shifted with an increase of 239
to over 80% while all the other plutonium isotopes will have decreased in proportion.

By removing the moderator, the size of the reactor core volume 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 fast neutron, it is possible to fuel such reactors with a mixture of plutonium and natural uranium, or with enriched material, containing around 20% 235
. Test runs at various facilities have also been done using 233
and 232
. The natural uranium (mostly 238
) will be turned into 239
, while in the case of 232
, 233
is the result. As new fuel is created during the operation, this process is called breeding.[4] All fast reactors can be used for breeding, or by carefully selecting the materials in the core and eliminating the blanket they can be operated to maintain the same level of fissionable material without creating any excess material. This is a process called Conversion because it transmutes fertile materials into fissile fuels on a 1:1 basis. By surrounding the reactor core with a blanket of 238
or 232
which captures excess neutrons, the extra neutrons breed more 239
or 233

The blanket material can then be processed to extract the new fissile material, which can then be mixed with depleted uranium to produce MOX fuel, mixed with lightly enriched Uranium fuel to form REMIX fuel both for conventional slow-neutron reactors. Alternatively it can be mixed as in greater percentage of 17%-19.75% fissile fuel for fast reactor cores. A single fast reactor can thereby supply its own fuel indefinitely as well as feed several thermal ones, greatly increasing the amount of energy extracted from the natural uranium. The most effective breeder configuration theoretically is able to produce 14 239
nuclei for every 10 (14:10) actinide nuclei consumed, however real world fast reactors have so far achieved a ratio of 12:10 ending the fuel cycle with 20% more fissile material than they held at the start of the cycle.[5] Less than 1% of the total Uranium mined is consumed in a thermal once-through cycle, while up to 60% of the natural uranium is fissioned in the best existing fast reactor cycles.

Given the current inventory of spent nuclear fuel (which contains reactor grade plutonium), it is possible to process this spent fuel material and reuse the actinide isotopes as fuel in a large number of fast reactors. This effectively consumes the 237
, reactor-grade plutonium, 241
, and 244
. Enormous amounts of energy are still present in the spent reactor fuel inventories; if fast reactor types were to be employed to use this material, that energy can be extracted for useful purposes.

Waste recycling

Fast-neutron reactors can potentially reduce the radiotoxicity of nuclear waste. Each commercial scale reactor would have an annual waste output of a little more than a ton of fission products, plus trace amounts of transuranics if the most highly radioactive components could be recycled. The remaining waste should be stored for about 500 years.[6]

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 are, 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,[6] 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.[3]

In the spent fuel from water moderated reactors, several plutonium isotopes are present, along with the heavier, transuranic elements. Nuclear reprocessing, a complex series of chemical extraction processes, mostly based on the PUREX process, can be used to extract the unchanged uranium, the fission products, the plutonium, and the heavier elements.[7] Such waste streams can be divided in categories; 1) unchanged uranium-238, which is the vast bulk of the material and has a very low radioactivity, 2) a collection of fission products and 3) the transuranic elements.


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 an effective neutron moderator.[4]

All operating fast reactors are liquid metal cooled reactors, which use sodium, lead, or lead-bismuth eutectic as coolants.[8] The early Clementine reactor used mercury coolant and plutonium metal fuel. In addition to its toxicity to humans, mercury has a high capture cross section (thus, it readily absorbs the neutrons, 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 useful 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. The particular salt formula used is crucial as some formulas are effective moderators while others are not.[9]

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. Purified nitrogen-15 has also been proposed as a coolant gas because it is more common than Helium and also has a very low neutron absorption cross section.[10][11]

However, all large-scale fast reactors have used molten metal coolant. Advantages of molten metals are low cost, the small activation potential and the large liquid ranges. The latter means that the material has a low melting point, and a high boiling point. Examples of these reactors include Sodium cooled fast reactor, which are still being pursued worldwide. Russia currently operates two such reactors on a commercial scale. Additionally, Russia has around eighty reactor years of experience with the Lead-cooled fast reactor which is rapidly gaining interest.


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 used (highly 235
enriched) uranium fuel initially, then in 2022 switched to using MOX.[12] The Indian prototype reactor uses uranium-carbide fuel.

While criticality at fast energies may be achieved with uranium enriched to 5.5 (weight) percent 235
, fast reactor designs have been proposed with enrichment 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 232
or 238
, respectively. Some designs use Burnable Poisons also known as Burnable Absorbers which contain isotopes with high neutron capture cross sections. Concentrated 10
or 155
& 157
in natural Gadolinium are typically used for this purpose. As these isotopes absorb excess neutrons they are transmuted into isotopes with low absorption cross sections so that over the life of the fuel cycle they are eliminated as more fission products with high capture cross section are generated. This makes it easier to maintain control of the reactivity rate in the core at start up with fresh fuel.[13]


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.


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.


Fast reactors are widely seen as an essential development because of several advantages over moderated designs.[14] The most studied and built fast reactor type is the sodium-cooled fast reactor. Some of the advantages of this design are discussed below; other designs such as the lead-cooled fast reactor and FMSR, Fast Molten Salt Reactor[15] have similar advantages.

These two effects serve to reduce the reactivity because it allows more neutrons to escape the core, as was shown in a demonstration at EBR-II in 1986.[16] In this test, the additional heat was readily absorbed by the large volume of liquid sodium, and the reactor shut itself down, without operator intervention.


As most fast reactors to date have been either sodium, lead or lead-bismuth cooled, the disadvantages of such systems are described here.

Some sodium-cooled fast reactors have operated safely for long periods (notably the Phénix and EBR-II for 30 years, or the BN-600 and BN-800 in operation since resp. 1980 and 2016, despite several minor leaks and fires. It is important to note that sodium leaks (and possibly fires) do not release radioactive elements, as the sodium fast reactors are always designed with a two loop system.


The BN-350 reactor was also used for desalination of sea water.

US interest in breeder reactors were 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.[20] The French reactors also met with serious opposition of environmentalist groups, who regarded these as very dangerous.[21] Despite such setbacks, a number of countries still invest in the fast reactor technology. Around 25 reactors have been built since the 1970s, accumulating over 400 reactor years of experience.

A 2008 IAEA proposal for a Fast Reactor Knowledge Preservation System[22] 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.

As of 2021, Russia operates two fast reactors on commercial scale.[23] The GEN IV initiative, an international working group on new reactor designs has proposed six new reactor types, three of which would operate with a fast spectrum.[24]

List of fast reactors

Decommissioned reactors

United States




Never operated


Under repair

Under construction

In design



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 IBR-2, BOR-60, BN-600,
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, ALFRED 4S, JSFR, KALIMER

See also


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  2. ^ "What is Neutron - Neutron Definition". Retrieved 2017-09-19.
  3. ^ a b Pintu 14/10/2019 Nuclear Power Plant (2019-10-14). "Difference Between Thermal Reactor and Fast Reactor". Retrieved 2022-04-13.((cite web)): CS1 maint: numeric names: authors list (link)
  4. ^ a b "How do fast breeder reactors differ from regular nuclear power plants?". Scientific American. Retrieved 2023-12-05.
  5. ^ "Conversion Factor - Breeding Ratio". Nuclear Power.
  6. ^ a b Smarter use of Nuclear Waste, by William H. Hannum, Gerald E. Marsh and George S. Stanford, Copyright Scientific American, Dec 2005.
  7. ^ "Fast Neutron Reactors | FBR - World Nuclear Association".
  8. ^ "Fast Neutron Reactors | FBR - World Nuclear Association". Retrieved 2023-12-05.
  9. ^ "Moltex Energy | Safer Cheaper Cleaner Nuclear | Stable Salt Reactors | SSR". Retrieved 2016-10-20.
  10. ^ Van Rooijen, W. F. G. (2009). "Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook". Science and Technology of Nuclear Installations. 2009: 1–11. doi:10.1155/2009/965757.
  11. ^ "Will heavy nitrogen become a widely used fission reactor coolant? - Atomic Insights". 17 November 2020.
  12. ^ "Beloyarsk BN-800 fast reactor running on MOX : Uranium & Fuel - World Nuclear News".
  13. ^
  14. ^ "What is a fast reactor?".
  15. ^ Alsayyari, Fahad; Tiberga, Marco; Perkó, Zoltán; Kloosterman, Jan Leen; Lathouwers, Danny (2021). "Analysis of the Molten Salt Fast Reactor using reduced-order models". Progress in Nuclear Energy. 140: 103909. doi:10.1016/j.pnucene.2021.103909.
  16. ^ "The Integral Fast Reactor". YouTube.
  17. ^ Mausolff, Zander; DeHart, Mark; Goluoglu, Sedat (2021). "Design and assessment of a molten chloride fast reactor". Nuclear Engineering and Design. 379: 111181. doi:10.1016/j.nucengdes.2021.111181. S2CID 234814975.
  18. ^ "Fukushima Daiichi Accident - World Nuclear Association".
  19. ^ Archived 2021-11-25 at the Wayback Machine [bare URL PDF]
  20. ^ "Process Banned by President Carter Could Solve U.S. Nuclear Waste Problem".
  21. ^ SCHNEIDER, MYCLE (June 26, 2009). "Fast Breeder Reactors in France". Science & Global Security. 17 (1): 36–53. Bibcode:2009S&GS...17...36S. doi:10.1080/08929880902953013. S2CID 122789053 – via Taylor and Francis+NEJM.
  22. ^ "Fast Reactor Knowledge Preservation System: Taxonomy and Basic Requirements" (PDF).
  23. ^ a b PRIS data base (2021)
  24. ^ "Home - Generation IV Systems". GIF Portal.
  25. ^ "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.
  26. ^ T. SOGA, W. ITAGAKI, Y. KIHARA, Y. MAEDA. Endeavor to improve in-pile testing techniques in the experimental fast reactor Joyo. 2013.
  27. ^ "Russia starts building lead-cooled fast reactor : New Nuclear - World Nuclear News".
  28. ^ "Решение о строительстве БН-1200 будет принято в 2014 году".
  29. ^ "В 2012 году на Белоярской АЭС начнется строительство пятого энергоблока БН-1800. РИА Новый День]". November 1, 2007. Retrieved 11 May 2017.
  30. ^ "Russia defers BN-1200 until after 2035". 2 January 2020.
  31. ^ "***지속가능원자력시스템***".
  32. ^ Wang, Brian (August 24, 2018). "Southern Company partnering with Bill Gates backed Terrapower on molten chloride fast reactor". Retrieved 2018-08-25.
  33. ^ "Technology".
  34. ^ "Generation IV & SMR".
  35. ^ "Overview of Indian Fast Breeder Nuclear Reactor Programme - Nuclear Power - Nuclear Reactor". Scribd.