Breeder reactor

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Assembly of the core of Experimental Breeder Reactor I in Idaho, 1951

A breeder reactor is a nuclear reactor capable of generating more fissile material than it consumes.[1] These devices are able to achieve this feat because their neutron economy is high enough to breed more fissile fuel than they use from fertile material like uranium-238 or thorium-232. Breeders were at first considered attractive because of their superior fuel economy compared to light water reactors. Interest in breeders declined after the 1960s as more uranium reserves were found,[2] and new methods of uranium enrichment reduced fuel costs.

Fuel efficiency and Types of Nuclear Waste[edit]

Proportions of the isotopes, uranium-238 (blue) and uranium-235 (red) found naturally, versus grades that are enriched. light water reactors largely run on fuel enriched to (3-4%), and don't extract much energy from U-238, while by contrast uranium breeder reactors mostly use U-238/natural uranium as their fuel, reducing the need and expense of enrichment technology and further strengthening the potential of deeming nuclear power as renewable energy.

Breeder reactors could, in principle, extract almost all of the energy contained in uranium or thorium, decreasing fuel requirements by a factor of 100 compared to traditional once-through light water reactors. Conventional Light Water Reactors extract less than 1% of the energy in the uranium mined from the earth.[3] The high fuel efficiency of breeder reactors could greatly dampen concerns about fuel supply or energy used in mining. Adherents claim that with seawater uranium extraction, there would be enough fuel for breeder reactors to satisfy our energy needs for as long as the current relationship between the sun and Earth persists, about 5 billion years at the current energy consumption rate (thus making nuclear energy as sustainable in fuel availability terms as solar or wind renewable energy).[4][5]

Nuclear waste became a greater concern by the 1990s. In broad terms, spent nuclear fuel has two main components. The first consists of fission products, the left over fragments of fuel atoms after they have been split to release energy. Fission products come in dozens of elements and hundreds of isotopes, all of them lighter than uranium. The second main component of spent fuel is transuranics, which are made from uranium or heavier atoms in the fuel when they absorb neutrons but do not undergo fission. The transuranics are all heavier than uranium. All transuranic isotopes fall within the actinide series on the periodic table, and so they are frequently referred to as the actinides.

The physical behavior of the fission products is markedly different from that of the transuranics. In particular, fission products do not themselves undergo fission, and therefore cannot be used for nuclear weapons. Furthermore, only seven long-lived fission product isotopes have half-lives longer than a hundred years, which makes their geological storage or disposal less problematic than for transuranic materials.[6]

With increased concerns about nuclear waste, breeding fuel cycles became interesting again because they can reduce actinide wastes, particularly plutonium and minor actinides.[7] Breeder reactors are designed to fission the actinide wastes as fuel, and thus convert them to more fission products.

After "spent nuclear fuel" is removed from a light water reactor, it undergoes a complex decay profile as each nuclide decays at a different rate. Due to a physical oddity referenced below, there is a large gap in the decay half-lives of fission products compared to transuranic isotopes. If the transuranics are left in the spent fuel, after 1,000 to 100,000 years, the slow decay of these transuranics would generate most of the radioactivity in that spent fuel. Thus, removing the transuranics from the waste eliminates much of the long-term radioactivity of spent nuclear fuel.[8]

Today's commercial Light Water Reactors do breed some new fissile material, mostly in the form of plutonium. Because commercial reactors were never designed as breeders, they do not convert enough uranium-238 into plutonium to replace the uranium-235 consumed. Nonetheless, at least one-third of the power produced by commercial nuclear reactors comes from fission of plutonium generated within the fuel.[9] Even with this level of plutonium consumption, light water reactors consume only part of the plutonium and minor actinides they produce, and nonfissile isotopes of plutonium build up, along with significant quantities of other minor actinides.[10] Even with reprocessing, reactor-grade plutonium is normally recycled only once in LWRs as mixed oxide fuel, with limited reductions in long-term waste radioactivity.[citation needed]

Conversion Ratio, Breakeven, Breeding Ratio, Doubling Time, and Burnup[edit]

One measure of a reactor's performance is the "conversion ratio" (the average number of fissile atoms created per fission event). All proposed nuclear reactors except specially designed and operated actinide burners[11] experience some degree of conversion. As long as there is any amount of a fertile material within the neutron flux of the reactor, some new fissile material is always created.

The ratio of new fissile material in spent fuel to fissile material consumed from the fresh fuel is known as the "conversion ratio" or "breeding ratio" of a reactor.

For example, commonly used Light Water Reactors have a conversion ratio of approximately 0.6. Pressurized Heavy Water Reactors (PHWR) running on natural uranium have a conversion ratio of 0.8.[12] In a breeder reactor, the conversion ratio is higher than 1. "Breakeven" occurs when the conversion ratio is exactly 1, and the reactor produces exactly as much fissile material as it uses.

"Doubling time" is the amount of time it would take for a breeder reactor to produce enough new fissile material to create a starting fuel load for another nuclear reactor. This was considered an important measure of breeder performance in early years, when uranium was thought to be scarce. However, since uranium is more abundant than thought, and given the amount of plutonium available in spent reactor fuel, doubling time has become a less important metric in modern breeder reactor design.[13][14]

"Burnup" is a measure of how much energy has been extracted from a given mass of heavy metal in fuel, often expressed (for power reactors) in terms of gigawatt-days per ton of heavy metal. Burnup is an important factor in determining the types and abundances of isotopes produced by a fission reactor. Breeder reactors, by design, have extremely high burnup compared to a conventional reactor. This burnup reflects the fact that breeder reactors produce much more of their waste in the form of fission products, while most or all of the actinides are meant to be fissioned and destroyed.[15]

Historically, breeder reactor development has focused on reactors with low breeding ratios, from 1.01 for the Shippingport Reactor[16][17] running on thorium fuel and cooled by conventional light water to over 1.2 for the Russian BN-350 liquid-metal-cooled reactor.[18] Theoretical models of breeders with liquid sodium coolant flowing through tubes inside fuel elements ("tube-in-shell" construction) suggest breeding ratios of at least 1.8 are possible.[19]

Types of breeder reactor[edit]

Many types of breeder reactor are possible:

A 'breeder' is simply a reactor designed for very high neutron economy with an associated conversion rate higher than 1.0. In principle, almost any reactor design could possibly be tweaked to become a breeder. An example of this process is the evolution of the Light Water Reactor, a very heavily moderated thermal design, into the Super Fast Reactor [20] concept, using light water in an extremely low-density supercritical form to increase the neutron economy high enough to allow breeding.

Aside from water cooled, there are many other types of breeder reactor currently envisioned as possible. These include molten-salt cooled and liquid metal cooled designs in many variations. Almost any of these basic design types may be fueled by uranium, plutonium, many minor actinides, or thorium, and they may be designed for many different goals, such as creating more fissile fuel, long-term steady-state operation, or active burning of nuclear wastes.

For convenience sake, it is perhaps simplest to divide the extant reactor designs into two broad categories based upon their neutron spectrum, which has the natural effect of dividing the reactor designs into those which are designed to utilize primarily uranium and transuranics, and those designed to use thorium and avoid transuranics.

  • Fast breeder reactor or FBR uses fast (unmoderated) neutrons to breed fissile plutonium and possibly higher transuranics from fertile uranium-238. The fast spectrum is flexible enough that it can also breed fissile uranium-233 from thorium, if desired.
  • Thermal breeder reactor use thermal spectrum (moderated) neutrons to breed fissile uranium-233 from thorium (thorium fuel cycle). Due to the behavior of the various nuclear fuels, a thermal breeder is thought commercially feasible only with thorium fuel, which avoids the buildup of the heavier transuranics.


Fission of the nuclear fuel in any reactor produces neutron-absorbing fission products. Because of this unavoidable physical process, it is necessary to reprocess the fertile material from a breeder reactor to remove those neutron poisons. This step is required if one is to fully utilize the ability to breed as much or more fuel than is consumed. All reprocessing can present a proliferation concern, since it extracts weapons usable material from spent fuel.[21] The most common reprocessing technique, PUREX, presents a particular concern, since it was expressly designed to separate pure plutonium. Early proposals for the breeder reactor fuel cycle posed an even greater proliferation concern because they would use PUREX to separate plutonium in a highly attractive isotopic form for use in nuclear weapons.[22][23]

Several countries are developing reprocessing methods that do not separate the plutonium from the other actinides. For instance, the non-water based pyrometallurgical electrowinning process, when used to reprocess fuel from the Integral Fast Reactor concept, leaves large amounts of radioactive actinides in the reactor fuel.[3] More conventional advanced reprocessing systems which are water based, like PUREX, include SANEX, UNEX, DIAMEX, COEX, and TRUEX, as well as proposals to combine PUREX with co-processes. All of these systems increase proliferation resistance compared to the classical PUREX system, although their adoption rate is low.[24][25]

In the thorium cycle, thorium-232 breeds by converting first to protactinium-233, which then decays to uranium-233. If the protactinium remains in the reactor, small amounts of U-232 are also produced, which has the strong gamma emitter Tl-208 in its decay chain. Similar to uranium-fueled designs, the longer the fuel and fertile material remain in the reactor, the more of these undesirable elements build up. Inside the envisioned commercial thorium reactors high levels of U232 would be allowed to accumulate, leading to extremely high gamma radiation doses from any uranium derived from thorium. These gamma rays complicate the safe handling of a weapon and the design of its electronics; this explains why U-233 has never been pursued for weapons beyond proof-of-concept demonstrations.[26]

Waste reduction[edit]

Actinides and fission products by half-life
Actinides[27] by decay chain Half-life
range (a)
Fission products of 235U by yield[28]
4n 4n+1 4n+2 4n+3
4.5–7% 0.04–1.25% <0.001%
228Ra 4–6 155Euþ
244Cm 241Puƒ 250Cf 227Ac 10–29 90Sr 85Kr 113mCdþ
232Uƒ 238Pu 243Cmƒ 29–97 137Cs 151Smþ 121mSn
248Bk[29] 249Cfƒ 242mAmƒ 141–351

No fission products
have a half-life
in the range of
100–210k years…

241Am 251Cfƒ[30] 430–900
226Ra 247Bk 1.3k–1.6k
240Pu 229Th 246Cm 243Am 4.7k–7.4k
245Cmƒ 250Cm 8.3k–8.5k
239Puƒ 24.1k
230Th 231Pa 32k–76k
236Npƒ 233Uƒ 234U 150k–250k 99Tc 126Sn
248Cm 242Pu 327k–375k 79Se
1.53M 93Zr
237Np 2.1M–6.5M 135Cs 107Pd
236U 247Cmƒ 15M–24M 129I
244Pu 80M

...nor beyond 15.7M[31]

232Th 238U 235Uƒ№ 0.7G–14.1G

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

Nuclear waste became a greater concern by the 1990s. Breeding fuel cycles attracted renewed interest because of their potential to reduce actinide wastes, particularly plutonium and minor actinides.[7] After the spent nuclear fuel has been removed from a light water reactor for longer than 100,000 years, these transuranics would be the main source of radioactivity. Eliminating them would eliminate much of the long-term radioactivity from the spent fuel.[8]

In principle, breeder fuel cycles can recycle and consume all actinides,[4] leaving only fission products. As the graphic in this section indicates, fission products have a peculiar 'gap' in their aggregate half-lives, such that no fission products have a half-life longer than 91 years and shorter than two hundred thousand years. As a result of this physical oddity, after several hundred years in storage the waste's radioactivity would drop to the low level of the long-lived fission products. However, this benefit requires highly efficient separation of transuranics from spent fuel. If the fuel reprocessing methods used leave a large fraction of the transuranics in its final waste stream, this advantage would be reduced.[3]

Both types of breeding cycles can reduce actinide wastes:

  • The fast breeder reactor's fast neutrons can fission actinide nuclei with even numbers of both protons and neutrons. Such nucleii usually lack the low-speed "thermal neutron" resonances of fissile fuels used in LWRs.[32]
  • The thorium fuel cycle inherently produces lower levels of heavy actinides. The fertile material in the thorium fuel cycle has an atomic weight of 232, while the fertile material in the uranium fuel cycle has an atomic weight of 238. That mass difference means that thorium-232 requires six more neutron capture events per nucleus before the transuranic elements can be produced. In addition to this simple mass difference, the reactor gets two chances to fission the nuclei as the mass increases: First as the effective fuel nuclei U233, and as it absorbs two more neutrons, again as the fuel nuclei U235.[33][34]

A reactor whose main purpose is to destroy actinides, rather than increasing fissile fuel stocks, is sometimes known as a burner reactor. Both breeding and burning depend on good neutron economy, and many designs can do either. Breeding designs surround the core by a breeding blanket of fertile material. Waste burners surround the core with non-fertile wastes to be destroyed. Some designs add neutron reflectors or absorbers.[11]

Breeder reactor concepts[edit]

There exist several concepts for breeder reactors, however the two main distinct concepts are:

  • Reactors with a fast neutron spectrum are called fast breeder reactors (FBR) – these typically utilize uranium-238 as fuel.
  • Reactors with a thermal neutron spectrum are called thermal breeder reactors – these typically utilize thorium-232 as fuel.

Fast breeder reactor[edit]

Schematic diagram showing the difference between the Loop and Pool types of LMFBR.

As of 2006, all large-scale fast breeder reactor (FBR) power stations have been liquid metal fast breeder reactors (LMFBR) cooled by liquid sodium. These have been of one of two designs:[1]

  • Loop type, in which the primary coolant is circulated through primary heat exchangers outside the reactor tank (but inside the biological shield due to radioactive sodium-24 in the primary coolant)
  • Pool type, in which the primary heat exchangers and pumps are immersed in the reactor tank

All current fast neutron reactor designs use liquid metal as the primary coolant, to transfer heat from the core to steam used to power the electricity generating turbines. FBRs have been built cooled by liquid metals other than sodium—some early FBRs used mercury, other experimental reactors have used a sodium-potassium alloy called NaK. Both have the advantage that they are liquids at room temperature, which is convenient for experimental rigs but less important for pilot or full scale power stations. Lead and lead-bismuth alloy have also been used. The relative merits of lead vs sodium are discussed here. Looking further ahead, three of the proposed generation IV reactor types are FBRs:[35]

FBRs usually use a mixed oxide fuel core of up to 20% plutonium dioxide (PuO2) and at least 80% uranium dioxide (UO2). Another fuel option is metal alloys, typically a blend of uranium, plutonium, and zirconium (used because it is "transparent" to neutrons). Enriched uranium can also be used on its own.

In many designs, the core is surrounded in a blanket of tubes containing non-fissile uranium-238 which, by capturing fast neutrons from the reaction in the core, is converted to fissile plutonium-239 (as is some of the uranium in the core), which is then reprocessed and used as nuclear fuel. Other FBR designs rely on the geometry of the fuel itself (which also contains uranium-238), arranged to attain sufficient fast neutron capture. The plutonium-239 (or the fissile uranium-235) fission cross-section is much smaller in a fast spectrum than in a thermal spectrum, as is the ratio between the 239Pu/235U fission cross-section and the 238U absorption cross-section. This increases the concentration of 239Pu/235U needed to sustain a chain reaction, as well as the ratio of breeding to fission.[11]

On the other hand, a fast reactor needs no moderator to slow down the neutrons at all, taking advantage of the fast neutrons producing a greater number of neutrons per fission than slow neutrons. For this reason ordinary liquid water, being a moderator as well as a neutron absorber, is an undesirable primary coolant for fast reactors. Because large amounts of water in the core are required to cool the reactor, the yield of neutrons and therefore breeding of 239Pu are strongly affected. Theoretical work has been done on reduced moderation water reactors, which may have a sufficiently fast spectrum to provide a breeding ratio slightly over 1. This would likely result in an unacceptable power derating and high costs in an liquid-water-cooled reactor, but the supercritical water coolant of the SCWR has sufficient heat capacity to allow adequate cooling with less water, making a fast-spectrum water-cooled reactor a practical possibility.[20]

Integral Fast Reactor[edit]

One design of fast neutron reactor, specifically designed to address the waste disposal and plutonium issues, was the Integral Fast Reactor (also known as an Integral Fast Breeder Reactor, although the original reactor was designed to not breed a net surplus of fissile material).[36][37]

To solve the waste disposal problem, the IFR had an on-site electrowinning fuel reprocessing unit that recycled the uranium and all the transuranics (not just plutonium) via electroplating, leaving just short half-life fission products in the waste. Some of these fission products could later be separated for industrial or medical uses and the rest sent to a waste repository (where they would not have to be stored for anywhere near as long as wastes containing long half-life transuranics). The IFR pyroprocessing system uses molten cadmium cathodes and electrorefiners to reprocess metallic fuel directly on-site at the reactor.[38] Such systems not only commingle all the minor actinides with both uranium and plutonium, they are compact and self-contained, so that no plutonium-containing material ever needs to be transported away from the site of the breeder reactor. Breeder reactors incorporating such technology would most likely be designed with breeding ratios very close to 1.00, so that after an initial loading of enriched uranium and/or plutonium fuel, the reactor would then be refueled only with small deliveries of natural uranium metal. A quantity of natural uranium metal equivalent to a block about the size of a milk crate delivered once per month would be all the fuel such a 1 gigawatt reactor would need.[39] Such self-contained breeders are currently envisioned as the final self-contained and self-supporting ultimate goal of nuclear reactor designers.[3][11] The project was canceled in 1994, at the behest of then-United States Secretary of Energy Hazel O'Leary.[40][41]

Other fast reactors[edit]

Another proposed fast reactor is a fast 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).

Several prototype FBRs have been built, ranging in electrical output from a few light bulbs' equivalent (EBR-I, 1951) to over 1,000 MWe. As of 2006, the technology is not economically competitive to thermal reactor technology—but India, Japan, China, South Korea and Russia are all committing substantial research funds to further development of Fast Breeder reactors, anticipating that rising uranium prices will change this in the long term. Germany, in contrast, abandoned the technology due to safety concerns. The SNR-300 fast breeder reactor was finished after 19 years despite cost overruns summing up to a total of 3.6 billion Euros, only to then be abandoned.[42]

As well as their thermal breeder program, India is also developing FBR technology, using both uranium and thorium feedstocks.

Thermal breeder reactor[edit]

The advanced heavy water reactor (AHWR) is one of the few proposed large-scale uses of thorium.[43] India is developing this technology, their interest motivated by substantial thorium reserves; almost a third of the world's thorium reserves are in India, which also lacks significant uranium reserves.

The third and final core of the Shippingport Atomic Power Station 60 MWe reactor was a light water thorium breeder, which began operating in August 1977 and after testing was brought to full power by the end of that year.[44] It used pellets made of thorium dioxide and uranium-233 oxide; initially, the U-233 content of the pellets was 5-6% in the seed region, 1.5-3% in the blanket region and none in the reflector region. It operated at 236 MWt, generating 60 MWe and ultimately produced over 2.1 billion kilowatt hours of electricity. After five years, the core was removed and found to contain nearly 1.4% more fissile material than when it was installed, demonstrating that breeding from thorium had occurred.[45][46]

The liquid fluoride thorium reactor (LFTR) is also planned as a thorium thermal breeder. Liquid-fluoride reactors may have attractive features, such as inherent safety, no need to manufacture fuel rods and possibly simpler reprocessing of the liquid fuel. This concept was first investigated at the Oak Ridge National Laboratory Molten-Salt Reactor Experiment in the 1960s. It has recently been the subject of a renewed interest worldwide.[47] Japan, China, the UK, as well as private US, Czech and Australian companies have expressed intent to develop and commercialize the technology.

Breeder reactor controversy[edit]

Like many aspects of nuclear power, fast breeder reactors have been subject to much controversy over the years. In 2010, The International Panel on Fissile Materials said "After six decades and the expenditure of the equivalent of tens of billions of dollars, the promise of breeder reactors remains largely unfulfilled and efforts to commercialize them have been steadily cut back in most countries". In Germany, the United Kingdom, and the United States, breeder reactor development programs have been abandoned.[48][49] The rationale for pursuing breeder reactors — sometimes explicit and sometimes implicit — was based on the following key assumptions:[49][50]

  • It was expected that uranium would be scarce and high-grade deposits would quickly become depleted if fission power were deployed on a large scale; the reality, however, is that since the end of the cold war, uranium has been much cheaper and more abundant than early designers expected.[51]
  • It was expected that breeder reactors would quickly become economically competitive with the light-water reactors that dominate nuclear power today, but the reality is that capital costs are at least 25% more than water cooled reactors.
  • It was thought that Breeder reactors could be as safe and reliable as light-water reactors, but safety issues are cited as a concern with fast reactors that use a sodium coolant, where a leak could lead to a sodium fire.
  • It was expected that the proliferation risks posed by breeders and their “closed” fuel cycle, in which plutonium would be recycled, could be managed. But since plutonium breeding reactors produce plutonium from U238, and thorium reactors produce fissile U233 from thorium, all breeding cycles could theoretically pose proliferation risks.[52]

These problems have stymied their deployment and lent credence to calls for their abandonment.

There are some past anti-nuclear advocates that have become pro-nuclear power as a clean source of electricity since breeder reactors effectively recycle most of their waste. This solves one of the most important negative issues of nuclear power. In the documentary "Pandora's Promise", a case is made for breeder reactors because they provide a real, high kW alternative to fossil fuel energy. According to the movie, one pound of uranium provides as much power as 5000 barrels of oil.[53]

FBRs have been built and operated in the United States, the United Kingdom, France, the former USSR, India and Japan.[1] An experimental FBR in Germany was built but never operated. As of now, one such reactor is currently being used for power generation (and the second one to come in the beginning of 2015). There are a few planned reactors, and quite a few are being used for research related to the Generation IV reactor initiative.[citation needed]

Breeder reactor development and notable breeder reactors[edit]


Superphénix, the world's largest fast breeder reactor

France's first fast reactor, Rapsodie first achieved criticality in 1967. Built at Cadarache near Aix-en-Provence, Rapsodie was a loop-type reactor with a thermal output of 40 MW and no electrical generation facilities, and closed in 1983. The plant was also a focus point of anti-nuclear political activity by the Green party and other groups. Right wing groups claim the plant was shut down for political reasons and not lack of power generation.

This was followed by the 233 MWe Phénix, grid connected since 1973, both as a power reactor and as the center of work on reprocessing of nuclear waste by transmutation. It was shut down in 2009.[54] The life-time load factor was just below 40%, according to the IAEA data base PRIS.[55][56][57][58]

Superphénix, 1,200 MWe, entered service in 1984 and as of 2006 remains the largest FBR yet built. It was shut down in 1998,[59] having produced no electricity for most of the preceding ten years. The lifetime load factor was 7.79% according to IAEA.[55]


Germany has built two FBRs.

KNK-II as a Research reactor was converted from a thermal reactor, KNK-I, which had been used to study sodium cooling. KNK-II first achieved criticality as a fast reactor in 1977, and produced 20 MWe.[60] It was shut down in 1991 and is being dismantled[61]

Construction of the 300 MWe SNR-300 at Kalkar in North Rhine-Westphalia was completed in 1985 but never operated. The price had increased from 0.5 billion DM to 7.1 billion DM, the Three Mile Island accident had heightened public opposition to nuclear power, and the expected increase in electricity consumption had not occurred. The plant was maintained and staffed until a decision to close it was finally made in 1991, and has since been decommissioned. Today, it houses an amusement park (Wunderland Kalkar).[62]


India has an active development programme featuring both fast and thermal breeder reactors.[63]

India’s first 40 MWt Fast Breeder Test Reactor (FBTR) attained criticality on 18 October 1985. India has developed the technology to produce the plutonium rich U-Pu mixed carbide fuel. This can be used in the Fast Breeder Reactor.[64]

At present the scientists of the Indira Gandhi Centre for Atomic Research (IGCAR), one of the nuclear R & D institutions of India, are engaged in the construction (already in its final stages) of another FBR — the 500 MWe prototype fast breeder reactor - at Kalpakkam, near Chennai,[65] with plans to build more as part of its three stage nuclear power program.

India has the capability to use thorium cycle based processes to extract nuclear fuel. This is of special significance to the Indian nuclear power generation strategy as India has one of the world's largest reserves of thorium, which could provide power for more than 10,000 years,[66][67] and perhaps as long as 60,000 years.[68][69]


Jōyō is a test sodium-cooled fast reactor located in Ōarai, Ibaraki, operated by the Japan Atomic Energy Agency. The reactor was built in the 1970s for the purpose of experimental tests and the development of FBR technologies.

Japan has built one demonstration FBR, Monju, in Tsuruga, Fukui Prefecture, adding onto the research base developed by its older research FBR, the Joyo reactor. Monju is a sodium-cooled, MOX-fueled loop type reactor with 3 primary coolant loops, producing 714 MWt / 280 MWe.

Monju began construction in 1985 and was completed in 1991. It first achieved criticality on April 5, 1994. It was closed in December 1995 following a sodium leak and fire in a secondary cooling circuit, and was expected to restart in 2008. The reactor was restarted for tests in May 2010, for the goal to production usage in 2013.[70] However, on August 26, 2010, a 3.3-tonne "In‐Vessel Transfer Machine" fell into the reactor vessel when being removed after a scheduled fuel replacement operation.[71] The fallen device was not retrieved from the reactor vessel until 23 June 2011.[72]

In April 2007, the Japanese Government selected Mitsubishi Heavy Industries as the "core company in FBR development in Japan". Shortly thereafter, MHI started a new company, Mitsubishi FBR Systems (MFBR), with the explicit purpose of developing and eventually selling FBR technology.[73]


Main article: Dounreay

The UK fast reactor programme was conducted at Dounreay in Scotland, from 1957 until the programme was cancelled in 1994. Three reactors were constructed, two of them fast neutron power reactors, and the third, DMTR, being a heavy water moderated research reactor used to test materials for the program. Fabrication and reprocessing facilities for fuel for the two fast reactors and for the test rigs for DMTR were also constructed onsite. Dounreay Fast Reactor (DFR) achieved its first criticality in 1959. It used NaK coolant and produced 14MW of electricity. This was followed by the sodium-cooled 250 MWe Prototype Fast Reactor (PFR) in the 1970s. PFR was closed down in 1994 as the British government withdrew major financial support for nuclear energy development, DFR and DMTR both having previously been closed.[74][75] As of 2013, The UK has shown interest in the PRISM reactor and is working in concert with France to develop ASTRID.


On December 20, 1951, the fast reactor EBR-I (Experimental Breeder Reactor-1) at the Argonne National Laboratory-West in Idaho[76] produced enough electricity to power four light bulbs, and the next day produced enough power to run the entire EBR-I building. This was a milestone in the development of nuclear power reactors.[77][78] The reactor was decommissioned in 1964.

The next generation experimental breeder was EBR-II (Experimental Breeder Reactor-2), which went into service at Argonne National Laboratory in 1964 and operated until 1994.[76] It was designed to be an "integral" nuclear plant (based on the Integral Fast Reactor design), equipped to handle fuel recycling onsite. It typically operated at 20 MW out of its 62.5 MW maximum design power, and provided the bulk of heat and electricity to the surrounding facilities.[79]

The world's first commercial LMFBR (Liquid Metal Fast Breeder Reactor), and the only one yet built in the USA, was the 94 MWe Unit 1 at Enrico Fermi Nuclear Generating Station. Designed in a joint effort between Dow Chemical and Detroit Edison as part of the Atomic Power Development Associates consortium, groundbreaking in Lagoona Beach, Michigan (near Monroe, Michigan) took place in 1956. The plant went into operation in 1963. It shut down on October 5, 1966 due to high temperatures caused by a loose piece of zirconium which was blocking the molten sodium coolant nozzles. Partial melting damage to six subassemblies within the core was eventually found. (This incident was the basis for a controversial book by investigative reporter John G. Fuller titled We Almost Lost Detroit.) The zirconium blockage was removed in April 1968, and the plant was ready to resume operation by May 1970, but a sodium coolant fire delayed its restart until July. It subsequently ran until August 1972, when its operating license renewal was denied.

The Clinch River Breeder Reactor Project was announced in January 1972. A government/business cooperative effort, construction proceeded fitfully and abandoned in 1982 because the US has since halted its spent-fuel reprocessing program.[80] Funding for this project was halted by Congress on October 26, 1983.[81][82]

The Fast Flux Test Facility, first critical in 1980, is not a breeder but is a sodium-cooled fast reactor. It is in cold standby.


The Soviet Union constructed a series of fast reactors, the first being mercury-cooled and fueled with plutonium metal, and the later plants sodium-cooled and fueled with plutonium oxide.

BR-1 (1955) was 100W (thermal) was followed by BR-2 at 100 kW and then the 5MW BR-5.

BOR-60 (first criticality 1969) was 60 MW, with construction started in 1965.[83]

BN-350 (1973) was the first full-scale Soviet FBR. Constructed on the Mangyshlak Peninsula in Kazakhstan and on the shore of the Caspian Sea, it supplied 130 MW of electricity plus 80,000 tonnes per day of desalinated fresh water to the city of Aktau. Its total output was regarded as the equivalent of 350MWe, hence the designation.

BN-600 (1986, end of life 2020) is 1,470 MWth / 600 MWe.[84][85]


Russia has a plan for increasing its fleet of fast breeder reactor significantly. A construction of new plant BN-800 (800 MWe) at Beloyarsk, was completed in 2012. It is a successor to a smaller BN-600. At 27.06.2014 the BN-800 was started in the minimum power mode.[86] It is expected to start to work in nominal power mode in 2015.

There are plans for the construction of an even larger plant BN-1200 (1,200 MWe). It is expected to be completed in 2018.

Russia announced that it is going to build 2 additional BN-1200 (1,200 MWe) reactors by the end of 2030.[87]

An experimental lead-cooled fast reactor, BREST-300 will be built at the Siberian Chemical Combine (SCC) in Seversk. The BREST design is seen as a successor to the BN series and the 300 MWe unit at the SCC could be the forerunner to a 1,200 MWe version for wide deployment as a commercial power generation unit. The development program is as part of an Advanced Nuclear Technologies Federal Program 2010-2020 that seeks to exploit fast reactors as a way to be vastly more efficient in the use of uranium while 'burning' radioactive substances that otherwise would have to be disposed of as waste. BREST refers to bystry reaktor so svintsovym teplonositelem, Russian for 'fast reactor with lead coolant'. Its core would measure about 2.3 metres in diameter by 1.1 metres in height and contain 16 tonnes of fuel. The unit would be refuelled every year with each fuel element spending five years in total within the core. Lead coolant temperature would be around 540 °C, giving a high efficiency of 43% - primary heat production of 700 MWt yielding electrical power of 300 MWe. The operational lifespan of the unit could be 60 years. The design is expected to be completed by NIKIET in 2014 for construction between 2016 and 2020.[88]

Future plants[edit]

An FBR called the Prototype Fast Breeder Reactor is under construction in India, and is due to be completed in 2012. The commissioning date should be known by mid year.[89][90] The FBR program of India includes the concept of using fertile thorium-232 to breed fissile uranium-233. India is also pursuing the thorium thermal breeder reactor. A thermal breeder is not possible with purely uranium/plutonium based technology. Thorium fuel is the strategic direction of the power program of India, owing to the nation's large reserves of thorium, but worldwide known reserves of thorium are also some four times those of uranium. India's Department of Atomic Energy (DAE) says that it will simultaneously construct four more breeder reactors of 500 MWe each including two at Kalpakkam.[91]

The China Experimental Fast Reactor (CEFR) is a 25 MW(e) prototype for the planned China Prototype Fast Reactor (CFRP).[92] It started generating power on 21 July 2011.[93]

China has also initiated a research and development project in thorium molten-salt thermal breeder reactor technology (Liquid fluoride thorium reactor). It was formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011. Its ultimate target is to investigate and develop a thorium based molten salt nuclear system in about 20 years.[94][95]

Kirk Sorensen, former NASA scientist and Chief Nuclear Technologist at Teledyne Brown Engineering, has been a long time promoter of thorium fuel cycle and particularly liquid fluoride thorium reactors. In 2011, Sorensen founded Flibe Energy, a company aimed to develop 20-50 MW LFTR reactor designs to power military bases.[96][97][98][99]

South Korea is developing a design for a standardized modular FBR for export, to complement the standardized PWR (Pressurized Water Reactor) and CANDU designs they have already developed and built, but has not yet committed to building a prototype.

The BN-600 (Beloyarsk NNP in the town of Zarechny, Sverdlovsk Oblast) is still operational. A second reactor (the BN-800) is scheduled to finish construction before 2015. A third and possible fourth reactor is scheduled to begin construction in 2015. These include the BN-1200, and may possibly be expanded to include a second large design, the BN-1600[100]

On February 16, 2006, the U.S., France and Japan signed an "arrangement" to research and develop sodium-cooled fast reactors in support of the Global Nuclear Energy Partnership.[101]

In September 2010, French government allocated 651.6 millions euros to the Commissariat à l'énergie atomique to finalize the design of "Astrid" (Advanced Sodium Technological Reactor for Industrial Demonstration), a 600 MW reactor design of the 4th generation to be operational in 2020.[102][103]

In October 2010, GE Hitachi Nuclear Energy signed a memorandum of understanding with the operators of the Department of Energy's Savannah River site, which should allow the construction of a demonstration plant based on the company's S-PRISM fast breeder reactor prior to the design receiving full NRC licensing approval.[104] In October 2011, The Independent reported that the UK Nuclear Decommissioning Authority (NDA) and senior advisers within the Department for Energy and Climate Change (DECC) had asked for technical and financial details of the PRISM, partly as a means of reducing the country's plutonium stockpile.[105]

The traveling wave reactor proposed in a patent by Intellectual Ventures is a fast breeder reactor designed to not need fuel reprocessing during the decades-long lifetime of the reactor. The breed-burn wave in the TWR design does not move from one end of the reactor to the other but gradually from the inside out. Moreover, as the fuel's composition changes through nuclear transmutation, fuel rods are continually reshuffled within the core to optimize the neutron flux and fuel usage at any given point in time. Thus, instead of letting the wave propagate through the fuel, the fuel itself is moved through a largely stationary burn wave. This is contrary to many media reports, which have popularized the concept as a candle-like reactor with a burn region that moves down a stick of fuel. By replacing a static core configuration with an actively managed "standing wave" or "soliton" core, TerraPower's design avoids the problem of cooling a highly variable burn region. Under this scenario, the reconfiguration of fuel rods is accomplished remotely by robotic devices; the containment vessel remains closed during the procedure, and there is no associated downtime.[106]

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External links[edit]