Fast-neutron reactor

Shevchenko BN350 nuclear fast reactor and desalination plant situated on the shore of the Caspian Sea. The plant generated 135 MW_e and provided steam for an associated desalination plant. View of the interior of the reactor hall.

A fast neutron reactor or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons. Such a reactor needs no neutron moderator, but must use fuel that is relatively rich in fissile material when compared to that required for a thermal reactor.

Introduction

Basic fission concepts

In order to sustain a fission chain reaction, the neutrons released in fission events have to react with other atoms in the fuel. The chance of this occurring depends on the energy of the neutron; most atoms will only undergo induced fission with high energy neutrons, although a smaller number prefer much lower energies.

Natural uranium consists mostly of three isotopes, U-238, U-235, and trace quantities of U-234, a decay product of U-238. U-238 accounts for roughly 99.3% of natural uranium and undergoes fission only by neutrons with energies of 5 MeV or greater, the so-called fast neutrons. About 0.7% of natural uranium is U-235, which undergoes fission by neutrons of any energy, but particularly by lower energy neutrons. When either of these isotopes undergoes fission they release neutrons around 1 to 2 MeV, too low to cause fission in U-238, and too high to do so easily in U-235.

The common solution to this problem is to slow the neutron from these fast speeds using a neutron moderator, any substance that reacts with the neutrons and makes them give up some of their energy. The most common moderator is normal water, which slows the neutrons through inelastic scattering until the neutrons reach thermal equilibrium with the water. The key to reactor design is to carefully lay out the fuel and water so the neutrons have time to slow enough to become highly reactive with the U-235, but not so far as to allow them easy pathways to escape the reactor core entirely.

Although U-238 will not undergo fission by the neutrons released in fission, thermal neutrons can be captured by the nucleus to transmute the atom into Pu-239. Pu-239 has a neutron cross section very similar to that of U-235, and most of the atoms created this way will undergo fission from the thermal neutrons. In most reactors this accounts for as much as ⅓ of the energy being generated. Not all of the Pu-239 is burned up during normal operation, and the leftover, along with leftover U-238, can be separated out to be used in new fuel during nuclear reprocessing.

Water is a common moderator for practical reasons, but has its disadvantages. From a nuclear standpoint, the primary problem is that water can absorb a neutron and remove it from the reaction. It does this just enough that the amount of U-235 in natural ore is too low to sustain the chain reaction; the neutrons lost though absorption in the water and U-238, along with those lost to the environment, results in too few left in the fuel. The most common solution to this problem is to slightly concentrate the amount of U-235 in the fuel to produce enriched uranium, with the leftover U-238 known as depleted uranium. Other designs use different moderators, like heavy water, that produce more neutrons than they absorb, allowing them to run on unenriched fuel. In either case, the reactor's neutron economy is based on thermal neutrons.

Fast fission, breeders

Although U-235 and Pu-239 are less sensitive to higher energy neutrons, they still remain somewhat reactive well into the MeV area. This means that if you enrich the fuel you will eventually reach a threshold where there are enough fissile atoms in the fuel that a chain reaction can be maintained even with fast neutrons.

The primary advantage is that by removing the moderator, the size of the reactor can be greatly reduced, and to some extent the complexity. This is commonly used for shipboard and submarine reactor systems, where size and weight are major concerns. The downside to the fast reaction is that fuel enrichment is an expensive process, so this is generally not suitable for electrical generation or other roles where cost is more important than size.

There is another advantage to the fast reaction that has led to considerable development for civilian use. Fast reactors lack a moderator, and thus lack one of the systems that remove neutrons from the system. Those running on Pu-239 further increase the number of neutrons, because its most common fission cycle gives off three neutrons rather than the mix of two and three neutrons released from U-235. By surrounding the reactor core with a moderator and then a blanket of U-238, those neutrons can be captured and used to breed more Pu-239. This is the same reaction that occurs internally in conventional designs, but in this case the blanket does not have to sustain a reaction and thus can be made of natural uranium or even depleted uranium.

Due to the surplus of neutrons from Pu-239 fission, the reactor will actually breed more Pu-239 than it consumes. The blanket material can then be processed to extract the Pu-239 to replace the losses in the reactor, and the surplus is then mixed with other fuel to produce MOX fuel that can be fed into conventional slow neutron reactors. A single fast reactor can thereby feed several slow ones, greatly increasing the amount of energy extracted from the natural uranium, from less than 1% in a normal once-through cycle, to as much as 60% in the best fast reactor cycles.

Given the limited stores of natural uranium ore, and the rate that nuclear power was expected to take over baseload generation, through the 1960s and 70s fast breeder reactors were seen as the solution to the world's energy needs. Using twice-through processing, a fast breeder economy increases the fuel capacity of known ore deposits by as much as 100 times, meaning that even existing ore sources would last hundreds of years. The disadvantage to this approach is that the breeder reactor has to be fed highly enriched fuel, which is very expensive to produce. Even though it breeds more fuel than it consumes, the resulting MOX is still expensive. 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, breeder designs were being widely experimented on, especially in the USA, France and the USSR. However, this coincided with a crash in uranium prices. The expected increased demand led mining companies to build up new 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 $40 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 had reached commercial operation proved to be economically disastrous. Interest in breeder reactors were further muted by Jimmy Carter's April 1977 decision to defer construction of breeders in the US due to proliferation concerns, and the terrible operating record of France's Superphénix reactor.

Advantages

Actinides and fission products by half-life
Actinides by decay chain				Half-life range (y)	Fission products of 235U by yield
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	249Cfƒ	242mAmƒ		141–351	No fission products have a half-life in the range of 100–210 k years ...
	241Amƒ		251Cfƒ	430–900
		226Ra№	247Bk	1.3 k – 1.6 k
240Pu	229Th№	246Cm	243Amƒ	4.7 k – 7.4 k
	245Cmƒ	250Cm		8.3 k – 8.5 k
			239Puƒ№	24.1 k
		230Th№	231Pa№	32 k – 76 k
236Npƒ	233Uƒ№	234U№		150 k – 250 k	‡	99Tc₡	126Sn
248Cm		242Pu		327 k – 375 k			79Se₡
				1.53 M		93Zr
	237Np№			2.1 M – 6.5 M		135Cs₡	107Pd
236U№			247Cmƒ	15 M – 24 M			129I₡
244Pu№				80 M	... nor beyond 15.7 M years
232Th№		238U№	235Uƒ№	0.7 G – 14.1 G
Legend for superscript symbols ₡  has thermal neutron capture cross section in the range of 8–50 barns ƒ  fissile m  metastable isomer №  naturally occurring radioactive material (NORM) þ  neutron poison (thermal neutron capture cross section greater than 3k barns) †  range 4–97 y: Medium-lived fission product ‡  over 200,000 y: Long-lived fission product

Fast neutron reactors can reduce the total radiotoxicity of nuclear waste, and dramatically reduce the waste's lifetime. They can also use all or almost all of the fuel in the waste. Fast neutrons have an advantage in the transmutation of nuclear waste. With fast neutrons, the ratio between splitting and the capture of neutrons of plutonium or minor actinide is often larger than when the neutrons are slower, at thermal or near-thermal "epithermal" speeds. The transmuted odd-numbered actinides (e.g. from Pu-240 to Pu-241) split more easily. 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 product, cesium-137, which has a half life of 30.1 years, 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 unexplored reserves, or extraction from dilute sources such as ordinary granite or the ocean. 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 from fissions caused by fast neutrons than from those caused by 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, such as was done at the Phénix reactor in Marcoule in 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.
• The fast reactor doesn't just transmute the inconvenient even-numbered transuranic elements (notably Pu-240 and U-238). It transmutes them, and then fissions them for power, so that these former wastes would actually become valuable.

Disadvantages

• Breeder reactors are costly to build and operate, and are not likely to be cost-competitive with thermal reactors unless the price of uranium increases dramatically.
• Due to the low cross sections of most materials at high neutron energies, critical mass in a fast reactor is much higher than a thermal reactor. In practice, this means significantly higher enrichment: >20% enrichment in a fast reactor compared to <5% enrichment in typical thermal reactors. This raises greater nuclear proliferation and nuclear security issues.
• Sodium is often used as a coolant in fast reactors, because it does not moderate neutron speeds much and has a high heat capacity. However, it burns and foams in air. It has caused difficulties in reactors (e.g. USS Seawolf (SSN-575), Monju), although some sodium-cooled fast reactors have operated safely (notably the Superphénix and EBR-II for 30 years).
• Since liquid metals have low moderating power and ratio and no other moderator is present, the primary interaction of neutrons with liquid metal coolants is the (n,gamma) reaction, which induces radioactivity in the coolant. Boiling in the coolant, e.g. in an accident, would reduce coolant density and thus the absorption rate, such that the reactor has a positive void coefficient, which is dangerous and undesirable from a safety and accident standpoint. This can be avoided with a gas cooled reactor, since voids do not form in such a reactor during an accident; however, activation in the coolant remains a problem. A helium-cooled reactor would avoid this, since the elastic scattering and total cross sections are approximately equal, i.e. there are very few (n,gamma) reactions in the coolant and the low density of helium at typical operating conditions means that the amount neutrons have few interactions with coolant.

Nuclear reactor design

Coolant

Water, the most common coolant in thermal reactors, is generally not a feasible coolant for a fast reactor, because it acts as a neutron moderator. However the Generation IV reactor known as the supercritical water reactor with decreased coolant density may reach a hard enough neutron spectrum to be considered a fast reactor. Breeding, which is the primary advantage of fast over thermal reactors, may be accomplished with a thermal, light-water cooled & moderated system using very high enriched (~90%) uranium.

All current fast reactors are liquid metal cooled reactors. The early Clementine reactor used mercury coolant and plutonium metal fuel. Sodium-potassium alloy (NaK) coolant is popular in test reactors due to its low melting point. In addition to its toxicity to humans, mercury has a high cross section 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 used or considered as a coolant in reactors. Molten lead cooling has been used in naval propulsion units as well as some other prototype reactors. All large-scale fast reactors have used molten sodium coolant.

Another proposed fast reactor is a 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).

Gas-cooled fast reactors have been the subject of research as well, as helium, the most commonly proposed coolant in such a reactor, has small absorption and scattering cross sections, thus preserving the fast neutron spectrum without significant neutron absorption in the coolant.

Nuclear fuel

In practice, sustaining a fission chain reaction with fast neutrons means using relatively highly enriched uranium or plutonium. The reason for this is that fissile reactions are favored at thermal energies, since the ratio between the Pu239 fission cross section and U238 absorption cross section is ~100 in a thermal spectrum and 8 in a fast spectrum. Fission and absorption cross sections are low for both Pu239 and U238 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 it is impossible to build a fast reactor using only natural uranium fuel. However, it is possible to build a fast reactor that will breed fuel (from fertile material) by producing more fissile material 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 with no further enrichment required. 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 have been using (high U-235 enriched) uranium fuel. The Indian prototype reactor has been using uranium-carbide fuel.

While criticality at fast energies may be achieved with uranium enriched to 5.5 weight percent Uranium-235, fast reactor designs have often been proposed with enrichments in the range of 20 percent for a variety of 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 had occurred. 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 mitigate the negative reactivity feedback from fuel depletion and fission product poisons. In a fast breeder reactor, the above applies, though the reactivity from fuel depletion is also compensated by the breeding of either Uranium-233 or Plutonium-239 and 241 from Thorium 232 or Uranium 238, respectively.

Control

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 very 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 itself can also provide quick negative feedback. Small reactors such as those used in submarines may use doppler broadening or thermal expansion of neutron reflectors.

Shevchenko BN350 desalination unit. View of the only nuclear-heated desalination unit in the world

History

A 2008 IAEA proposal for a Fast Reactor Knowledge Preservation System notes that:

during the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power.

List of fast reactors

USA

• CLEMENTINE, the first fast reactor, built in 1946 at Los Alamos National Laboratory. Plutonium metal fuel, mercury coolant, power 25 kW thermal, used for research, especially as a fast neutron source.
• EBR-I at Idaho Falls, which in 1951 became the first reactor to generate significant amounts of electrical power. Decommissioned 1964.
• Fermi 1 near Detroit was a prototype fast breeder reactor that began operating in 1957 and shut down in 1972.
• EBR-II Prototype for the Integral Fast Reactor, 1965–1995?.
• SEFOR in Arkansas, a 20 MWt research reactor which operated from 1969 to 1972.
• Fast Flux Test Facility, 400MWt, Operated flawlessly from 1982 to 1992, at Hanford Washington, now deactivated, liquid sodium is drained with argon backfill under care and maintenance.

Europe

• DFR (Dounreay Fast Reactor, 1959–1977, 14MWe) and PFR (Prototype Fast Reactor, 1974–1994, 250MWe), in Caithness, in the Highland area of Scotland.
• Rhapsodie in Cadarache, France, (20 then 40 MW) between 1967 and 1982.
• Superphénix, in France, 1200MWe, closed in 1997 due to a political decision and very high costs of operation.
• Phénix, 1973, France, 233 MWe, restarted 2003 at 140 MWe for experiments on transmutation of nuclear waste for six years, ceased power generation in March 2009, though it will continue in test operation and to continue research programs by CEA until the end of 2009. Stopped in 2010.
• KNK-II, Germany

USSR/Russia

• Small lead-cooled fast reactors used for naval propulsion, particularly by the Soviet Navy.
• BR-5 - research fast neutron reactor at the Institute of Physics and Energy in Obninsk. Years of operation 1959-2002.
• BN-350, constructed by the Soviet Union in Shevchenko (today's Aqtau) on the Caspian Sea, 130MWe plus 80,000 tons of fresh water per day.
• IBR-2 - research fast neutron reactor at the Joint Institute of Nuclear Research in Dubna (near Moscow).
• BN-600 - sodium-cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. Provides 560 MW to the Middle Urals power grid. In operation since 1980.
• BN-800 - sodium-cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. Designed to generate 880MW of electrical power. Started producing electricity in October, 2014. BN-800 reactor design is to be sold by Russia to the China.

Never operated

• Clinch River Breeder Reactor, USA
• Integral Fast Reactor, USA. Design emphasized fuel cycle based on on-site electrolytic reprocessing. Cancelled 1994 without construction.
• SNR-300, Germany
• Monju reactor, 300 MWe, in Japan. was closed in 1995 following a serious sodium leak and fire. It was restarted May 6, 2010 and in August 2010 another accident, involving dropped machinery, shut down the reactor again. As of June 2011, the reactor has only generated electricity for one hour since its first testing two decades prior.

Currently operating

• BN-600, 1981, Russia, 600 MWe, scheduled end of life 2010 but still in operation.
• BN-800, Russia, testing began June 27, 2014, estimated total power 880MW
• BOR-60 - sodium-cooled reactor at the Research Institute of Atomic Reactors in Dmitrovgrad. In operation since 1980.(experimental purposes)
• FBTR, 1985, India, 10.5 MWt (experimental purposes)
• China Experimental Fast Reactor, 65 MWt (experimental purposes), planned 2009, critical 2010

Under repair

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

Under construction

• PFBR, Kalpakkam, India, 500 MWe.

In design phase

• BN-1200, Russia, build starting after 2014, operation in 2018–2020
• Toshiba 4S being developed in Japan and was planned to be shipped to Galena, Alaska (USA) but progress is stalled (see Galena Nuclear Power Plant)
• KALIMER, 600 MWe, South Korea, projected 2030. KALIMER is a continuation of the sodium cooled, metallic fueled, fast neutron reactor in a pool represented by the Advanced Burner Reactor (2006), S-PRISM (1998-present), Integral Fast Reactor (1984-1994), and EBR-II (1965-1995).
• Generation IV reactor (Helium·Sodium·Lead cooled) US-proposed international effort, after 2030
• JSFR, Japan, project for a 1500 MWe reactor began in 1998, but without success.
• ASTRID, France, project for a 600 MWe sodium-cooled reactor. Planned experimental operation in 2020.

Chart

Fast reactors

	U.S.	Russia	Europe	Asia
Past	Clementine, EBR-I/II, SEFOR, FFTF	BN-350	Dounreay, Rhapsodie, Superphénix, Phénix (stopped in 2010)
Cancelled	Clinch River, IFR		SNR-300
Operating		BOR-60, BN-600, BN-800		FBTR, CEFR
Under repairing				Jōyō
Under construction				Monju, PFBR,
Planned	Gen IV (Gas·Sodium·Lead)	BN-1200	ASTRID	4S, JSFR, KALIMER

See also

• Nuclear fuel cycle
• Fast breeder reactor
• Sodium-cooled fast reactor
• Lead-cooled fast reactor
• Gas-cooled fast reactor
• Generation IV reactor
• Energy amplifier
• Thermal-neutron reactor

External links

• http://www.amazon.com/Concepts-Behind-Breeder-Reactor-Design/dp/3659180009
• ANL report on EARLY SOVIET FAST REACTORS
• Article on recent work on fast neutron reactors in Scientific American, December, 2005
• IAEA Fast Reactor Database
• Fast Reactor Data Retrieval and Knowledge Preservation seeks to establish a comprehensive, international inventory of fast reactor data and knowledge, which would be sufficient to form the basis for fast reactor development in 30 to 40 years from now.
• World Nuclear Association: Fast Neutron Reactors
• International Thorium Energy Organisation - www.IThEO.org