Fast breeder reactor
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2008-07-14T18:48:46Z
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{{Refimprove|date=May 2008}}
{{mergeto|Breeder reactor|date=July 2008}}
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==
[[Image:LMFBR schematics2.svg|thumb|right|400px|Schematic diagram showing the difference between the Loop and Pool types of LMFBR.]]
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 (element)|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]] (PuO<sub>2</sub>) and at least 80% [[uranium dioxide]] (UO<sub>2</sub>). Another fuel option is [[Nuclear fuel#Metal fuel|metal alloys]], typically a blend of [[uranium]], [[plutonium]], and [[zirconium]]. The plutonium used can be supplied by the reprocessing from reactor outputs or 'off the shelf' dismantled [[nuclear weapon]]s.
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.
The ratio between the Pu239 (or U235) fission cross-section and the U238 absorption cross-section is much higher in a thermal spectrum than in a fast spectrum. Therefore a higher enrichment of the fuel is needed in a fast reactor in order to reach a self-sustaining [[nuclear chain reaction]].
Since a fast reactor uses a fast spectrum 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 reactor#Fast reactors|Generation IV reactors]], and each presents some advantages.<ref name="LeadvsSodium">[http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V4D-4K2T5D7-5&_user=10&_coverDate=08%2F31%2F2006&_rdoc=1&_fmt=&_orig=search&_sort=d&view=c&_acct=C000050221&_version=1&_urlVersion=0&_userid=10&md5=df85272db6aac0f870d0004ed28d513e Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues]</ref> A [[Gas-cooled fast reactor|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]]).
==Economics==
{{Unreferencedsection|date=August 2007}}
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 'off the shelf' uranium and especially of [[enriched uranium]] have made current FBR technology uncompetitive with [[Pressurized water reactor|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 ==
{{Unreferencedsection|date=May 2008}}
Fission of the nuclear fuel in any reactor produces neutron-absorbing fission products, and because of this it is necessary to [[nuclear reprocessing|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 [[nuclear proliferation|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 [[Nuclear fuel cycle|closed fuel cycle]] is often seen as a greater proliferation concern than a once-through [[Nuclear fuel cycle|thermal fuel cycle]].
However, to date all known weapons programs have used far more easily built [[thermal reactor]]s 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).<ref>http://www.nuc.berkeley.edu/designs/ifr/</ref><ref>[http://www.nationalcenter.org/NPA378.html National Policy Analysis #378: Integral Fast Reactors: Source of Safe, Abundant, Non-Polluting Power - December 2001<!-- Bot generated title -->]</ref>
To solve the waste disposal problem, the IFR had an on-site [[electrowinning]] fuel reprocessing unit that recycled the uranium and all the [[transuranic]]s (not just plutonium) via [[electroplating]], leaving just short [[half-life]] [[fission product]]s 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-[[United States Secretary of Energy|Secretary of Energy]] [[Hazel R. O'Leary|Hazel O'Leary]].
==FBR generating plants==
===History===
{{Unreferencedsection|date=May 2008}}
FBRs have been built and operated in the [[United States|USA]], the [[United Kingdom|UK]], [[France]], the former [[Soviet Union|USSR]], [[India]] and [[Japan]]. An experimental FBR in [[Germany]] was built but never operated. As of 2004, a prototype FBR was under construction in [[People's Republic of China|China]].
====USA====
{{Unreferencedsection|date=May 2008}}
On [[December 20]], [[1951]], the fast reactor [[Experimental Breeder Reactor I|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 [[Experimental Breeder Reactor II|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 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 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 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. 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 in cold standby.
====India====
{{Unreferencedsection|date=May 2008}}
[[India]] has an active development programme featuring both fast and [[thermal breeder reactor]]s.
India’s first 40 MWt Fast Breeder Test Reactor ([[FBTR]]) attained criticality on [[18 October]] [[1985]]. Thus, India became the sixth nation to have the technology to build and operate an 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. The higher construction expense of the Fast Breeder Reactor in comparison with the [[Pressurised Heavy Water Reactor]]s ([[PHWR]]) in use is one of the main reasons why India is looking at the cheaper option - uranium fuel.
====France====
{{Unreferencedsection|date=May 2008}}
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 [[Nuclear transmutation|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. 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.
====UK====
{{main|Dounreay}}
{{Unreferencedsection|date=May 2008}}
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====
{{Unreferencedsection|date=May 2008}}
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]]).
====USSR====
{{Unreferencedsection|date=May 2008}}
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 reactor|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 130MW of electricity plus 80,000 tonnes per day of [[de-salination|desalinated]] fresh water to the city of [[Aktau]]. Its total output was regarded as the equivalent of 350MWe, hence the designation.
'''[[BN-600 reactor|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====
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 [[5 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 FBR Systems]] (MFBR), with the explicit purpose of developing and eventually selling FBR technology.<ref>[http://www.neimagazine.com/story.asp?sectioncode=132&storyCode=2045341 Nuclear Engineering International<!-- Bot generated title -->]</ref>
===Future plants===
As of 2003 one indigenous FBR was planned for [[India]], and another for [[People's Republic of China|China]] using Soviet technology.
[[South Korea]] is developing a design for a standardised modular FBR for export, to complement the standardised [[Pressurized Water Reactor|PWR]] (Pressurized Water Reactor) and [[CANDU reactor|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|Zarechny]], [[Sverdlovsk Oblast]]) is still operational. A second reactor ([[BN-800]]) is scheduled to be constructed before 2015.<ref>[http://www.bellona.no/en/international/russia/npps/beloyarsk/35835.html Russian parliament and Kremlin party defends fast-breeder reactors - Bellona<!-- Bot generated title -->]</ref>
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]].<ref>[http://www.energy.gov/news/3218.htm Department of Energy - Generation IV International Forum Signs Agreement to Collaborate on Sodium Cooled Fast Reactors<!-- Bot generated title -->]</ref>
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.<ref>[http://www.india-defence.com/reports/2854 Home | India Defence<!-- Bot generated title -->]</ref>
{{Nuclear Technology}}
==See also==
*[[Breeder reactor]]
*[[Fast neutron reactor]]
*[[Sodium-cooled fast reactor]]
*[[Integral Fast Reactor]]
*[[Lead cooled fast reactor|Lead-cooled fast reactor]]
*[[Gas-cooled fast reactor]]
==References==
<references/>
==External links==
*[http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1350/index.html US Nuclear Program]
*[http://www-frdb.iaea.org/index.html IAEA Fast Reactors Database]
*[http://www.iaea.org/inis/aws/fnss/abstracts/index.html IAEA Technical Documents on Fast Reactors]
*[http://www.atomicheritage.org/ebr1.htm Atomic Heritage Foundation - EBR-I]
*[http://www.world-nuclear.org/sym/1999/pdfs/wilson.pdf The Changing Need for a Breeder Reactor] by Richard Wilson at The Uranium Institute 24th Annual Symposium, September 1999
*[http://www.ans.org/store/vi-300032 Experimental Breeder Reactor-II (EBR-II): An Integrated Experimental Fast Reactor Nuclear Power Station]
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