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Fast breeder reactor

The fast breeder or fast breeder reactor (FBR) is a fast neutron reactor designed to breed fuel by producing more fissile material than it consumes. The FBR is one possible type of breeder reactor.


Reactor designs

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

  • Loop type, in which the primary coolant is circulated through primary heat exchangers external to the reactor tank (but within the biological shield owing to the presence of radioactive sodium-24 in the primary coolant).
  • Pool type, in which the primary heat exchangers and circulators are immersed in the reactor tank.

Prototype FBRs have also been built cooled by other liquid metals such as mercury, lead and NaK (an alloy of sodium (Na) and potassium (K)), and one generation IV reactor proposal is for a helium cooled FBR.

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. The plutonium used can be supplied by the reprocessing from reactor outputs or dismantled nuclear weapons.

In many FBR designs, the reactor 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 partially converted to fissile plutonium 239 (as is some of the uranium in the core), which can then be reprocessed for use as nuclear fuel. Other FBR designs rely on the geometry of the fuel itself (which also contains uranium-238) to attain sufficient fast neutron capture.

While fast neutrons are less likely to be absorbed by uranium-235 or plutonium-239 than thermal neutrons, the highly enriched fuel used in fast breeder reactors allows for a self-sustaining nuclear chain reaction. For this reason, no moderator is required to thermalize the fast neutrons.

All current fast reactor designs use liquid metal as the primary coolant, to transfer heat from the core to steam used to power the electricity generating turbines. Some early FBRs used mercury, and other experimental reactors have used NaK. Both of these choices 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.

Sodium is the normal coolant for large power stations, but lead has been used successfully for smaller generating rigs. Both coolant choices are being studied as possible Generation IV reactors, and each presents some advantages.[1] A gas-cooled option is also being studied, although no gas-cooled fast reactor has reached criticality.

Water cannot be used as the primary coolant since it acts as a moderator, slowing neutrons to thermal levels and preventing the breeding of uranium-238 into plutonium 239. However a heavy water moderated thermal breeder reactor, using thorium to produce uranium-233, is theoretically possible (see Advanced Heavy Water Reactor).


The breeding of plutonium fuel in FBRs, known as the plutonium economy, was for a time believed to be the future of nuclear power. It remains the strategic direction of the power program of Japan. However, cheap supplies of uranium and especially of enriched uranium have made current FBR technology uncompetitive with PWR and other thermal reactor designs. PWR designs remain the most common existing power reactor type and also represent most current proposals for new nuclear power stations.

Possible technology risks

Fission of the nuclear fuel in any reactor produces neutron-absorbing fission products, and because of this it is necessary to reprocess the fuel and breeder blanket from a breeder reactor if one is to fully utilise its ability to breed more fuel than it consumes. The most common reprocessing technique, PUREX, is generally considered a large proliferation concern because such reprocessing technologies can be used to extract weapons grade plutonium from a reactor operated on a short refuelling cycle. For this reason, the FBR closed fuel cycle is often seen as a greater proliferation concern than a once-through thermal fuel cycle.

However, to date all known weapons programs have used far more easily built thermal reactors to produce plutonium, and there are some designs such as the SSTAR which avoid proliferation risks by both producing low amounts of plutonium at any given time from the U-238, and by producing three different isotopes of plutonium (Pu-239, Pu-240, and Pu-242) making the plutonium used infeasible for atomic bomb use.

Furthermore, several countries are developing more proliferation resistant reprocessing methods that don't separate the plutonium from the other actinides. For instance, the pyrometallurgical process when used to reprocess fuel from the Integral Fast Reactor leaves large amounts of radioactive actinides in the reactor fuel. Removing these transuranics in a conventional reprocessing plant would be extremely difficult as many of the actinides emit strong neutron radiation, requiring all handling of the material to be done remotely, thus preventing the plutonium from being used for bombs while still being usable as reactor fuel.

Thorium fueled reactors may pose a slightly higher proliferation risk than uranium based reactors. The reason for this is that while Pu-239 will fairly often fail to undergo fission on neutron capture, producing Pu-240, the corresponding process in the Thorium cycle is relatively rare. Thorium-232 converts to U-233, which will almost always undergo fission successfully, meaning that there will be very little U-234 produced in the reactor's thorium/U-233 breeder blanket, and the resulting pure U-233 will be comparatively easy to extract and use for weapons. One proposed solution to this is to mix a small amount of natural or depleted uranium into the thorium breeder blanket. The irradiated material will then be useless for weapons purposes as then the U-233 would require isotopic separation from the U-238. A small amount of plutonium would be present but will also be low-grade.

Associated reactor types

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).[2][3]

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). It is thought that it would not be possible to divert fuel from this reactor to make bombs, as several of the transuranics spontaneously undergo fission so rapidly that any assembly would melt before it could be completed. The project was canceled in 1994, at the behest of then-Secretary of Energy Hazel O'Leary.

FBR generating plants


FBRs have been built and operated in the USA, the UK, France, the former USSR, India and Japan. An experimental FBR in Germany was built but never operated. As of 2004, a prototype FBR was under construction in China.


On December 20, 1951, the fast reactor EBR-I (Experimental Breeder Reactor-1) at the Idaho National Laboratory in Idaho Falls, Idaho 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.

The next generation experimental breeder was EBR-II (Experimental Breeder Reactor-2), which went into service at the INEEL in 1964 and operated until 1994. It was designed to be an "integral" nuclear plant, equipped to handle fuel recycling onsite. It typically operated at 20 megawatts out of its 62.5 megawatt maximum design power, and provided the bulk of heat and electricity to the surrounding facilities.

The world's first commercial LMFBR, and the only one yet built in the USA, was the 94MWe 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 Association 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 of 1968, and the plant was ready to resume operation by May of 1970, but a sodium coolant fire delayed its restart until July. It subsequently ran until August of 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. Funding for this project was halted by Congress on October 26, 1983.

The Fast Flux Test Facility, first critical in 1980, is not a breeder but is a sodium-cooled fast reactor. It is now (2005) in cold standby.


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

India’s first 40 MWt Fast Breeder Test Reactor (FBTR) attained criticality on 18th October 1985. Thus, India became the sixth nation to have the technology to build and operate a FBTR after US, UK, France, Japan and the former USSR. India has developed the technology to produce the plutonium rich U-Pu mixed carbide fuel. This can be used in the Fast Breeder Reactor.

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 of another FBR - the 500 MWe prototype fast breeder reactor - at Kalpakkam, near Chennai.

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 large reserves of thorium — about 360,000 tonnes — that can fuel nuclear projects for an estimated 2,500 years. But the hitch is with the expensive nature of the construction of Fast Breeder Reactor in comparison with the Pressurised Heavy Water Reactors (PHWR) in use. This is one of the main reasons why India is looking at the cheaper option - Uranium fuel.


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 40MW and no electrical generation facilities, and closed in 1983.

This was followed by the 233 MWe Phénix, grid connected since 1973 and still operating, both as a power reactor and more importantly as the center of work on reprocessing of nuclear waste by transmutation.

Superphénix, 1200 MWe, entered service in 1984 and as of 2006 remains the largest FBR yet built. It was shut down in 1997 due to political commitment of the left-wing government to competitive market forces. Ironically the power plant had not produced electricity for most of the preceding ten years prior to its closure.

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.


Main article: Dounreay

The UK fast reactor programme was conducted at Dounreay, 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.


Germany has built two FBRs, but both were closed in 1991 without the larger ever having achieved criticality.

KNK-II 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 20MWe.

Construction of the 300MWe SNR-300 at Kalkar in North Rhine-Westphalia was completed in 1985, but owing to political pressure it was never operated. The plant was maintained and staffed until a decision to close it was finally made in 1990, and has since been decommissioned. Today it houses an amusement park (Wunderland Kalkar).


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.

BN-350 (1973) was the first full-scale Soviet FBR. Constructed on the Mangyshlak Peninsula in Kazakstan and on the shore of the Caspian Sea, it supplied 130MW 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) is 1470MWth / 600MWe.

At the time of the break up of the Soviet Union, plans were well underway for the construction of two larger plants, BN-800 (800 MWe) at Beloyarsk and BN-1600 (1600 MWe).


Japan has built one demonstration FBR, Monju, in Tsuruga, Fukui Prefecture, adding on to 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 the 5th of April 1994. It was closed in December 1995 following a sodium leak and fire in a secondary cooling circuit, and is expected to restart in 2008.

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 Fast Breeder Reactor Systems (MFBR), with the explicit purpose of developing and eventually selling FBR technology.[4]

Future plants

As of 2003 one indigenous FBR was planned for India, and another for China using Soviet technology.

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

The FBR program of India includes the concept of using fertile thorium-232 to breed fissile uranium-233. India is also pursuing the thermal breeder reactor again using thorium. 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 their large reserves of thorium, but worldwide known reserves of thorium are also some three times those of uranium.

The BN-600 (Beloyarsk NNP in the town of Zarechny, Sverdlovsk Oblast) is still operational. A second reactor (BN-800) is scheduled to be constructed before 2015.[5]

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.[6]

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.[7]

See also


  1. ^ Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues
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This article is licensed under the GNU Free Documentation License. It uses material from the Wikipedia article "Fast_breeder_reactor". A list of authors is available in Wikipedia.
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