A fast-neutron reactor (FNR) or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons (carrying energies above 0.5 MeV or greater, on average), as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.
Natural uranium consists mostly of three isotopes: , , and trace quantities of (a decay product of ). accounts for roughly 99.3% of natural uranium and undergoes fission only by fast neutrons. About 0.7% of natural uranium is , which undergoes fission by neutrons of any energy, but particularly by lower-energy neutrons. When either of these isotopes undergoes fission, it releases neutrons with an energy distribution peaking around 1 to 2 MeV. The flux of higher-energy fission neutrons (> 2 MeV) is too low to create sufficient fission in , and the flux of lower-energy fission neutrons (< 2 MeV) is too low to do so easily in .
The common solution to this problem is to slow the neutrons using a neutron moderator, which interacts with the neutrons to slow them. The most common moderator is water, which acts by elastic 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 , but not so far as to allow them to escape the reactor core.
Although does not undergo fission by the neutrons released in fission, thermal neutrons can be captured by the nucleus to transmute the uranium into . has a neutron cross section similar to that of , and most of the atoms created this way will undergo fission from the thermal neutrons. In most reactors this accounts for as much as 1/3 of generated energy. Some remains, and the leftover, along with leftover , can be recycled during nuclear reprocessing.
Water has disadvantages as a moderator. It can absorb a neutron and remove it from the reaction. It does this just enough that the concentration of in natural uranium is too low to sustain the chain reaction; the neutrons lost through absorption in the water and , 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 in the fuel to produce enriched uranium, with the leftover known as depleted uranium. Other designs use different moderators, like heavy water, that are much less likely to absorb neutrons, allowing them to run on unenriched fuel. In either case, the reactor's neutron economy is based on thermal neutrons.
Although and are less sensitive to higher-energy neutrons, they still remain somewhat reactive well into the MeV range. If the fuel is enriched, eventually a threshold will be reached where there are enough fissile atoms in the fuel to maintain a chain reaction 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 was commonly used for many early 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.
Another advantage to the fast reaction 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 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 . By surrounding the reactor core with a moderator and then a layer (blanket) of , those neutrons can be captured and used to breed more . 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 depleted uranium.
Due to the surplus of neutrons from fission, the reactor produces more than it consumes. The blanket material can then be processed to extract the to replace losses in the reactor, and the surplus is then mixed with uranium 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 existing fast reactor cycles, or more than 99% in the Integral Fast Reactor.
Given the limited reserves of uranium ore known in the 1960s, and the rate that nuclear power was expected to take over baseload generation, through the 1960s and 1970s fast breeder reactors were considered to be the solution to the world's energy needs. Using twice-through processing, a fast breeder increases the energy capacity of known ore deposits by as much as 100 times, meaning that existing ore sources would last hundreds of years. The disadvantage to this approach is that the breeder reactor has to be fed expensive, highly-enriched fuel. It was widely expected that this would still be below the price of enriched uranium as demand increased and known resources dwindled.
Through the 1970s, experimental breeder designs were examined, especially in the US, France and the USSR. However, this coincided with a crash in uranium prices. The expected increased demand led mining companies to expand supply channels, which came online just as the rate of reactor construction stalled in the mid-1970s. The resulting oversupply caused fuel prices to decline from about US$40 per pound in 1980 to less than $20 by 1984. Breeders produced fuel that was much more expensive, on the order of $100 to $160, and the few units that reached commercial operation proved to be economically 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.
Actinides and fission products by half-life
|Actinides by decay chain||Half-life
|Fission products of 235U by yield|
No fission products
... nor beyond 15.7 Ma
Legend for superscript symbols
Fast-neutron reactors can reduce the total radiotoxicity of nuclear waste  using all or almost all of the waste as fuel. With fast neutrons, the ratio between splitting and the capture of neutrons by plutonium and the minor actinides is often larger than when the neutrons are slower, at thermal or near-thermal "epithermal" speeds. The transmuted even-numbered actinides (e.g. , ) split nearly as easily as odd-numbered actinides in fast reactors. After they split, the actinides become a pair of "fission products". These elements have less total radiotoxicity. Since disposal of the fission products is dominated by the most radiotoxic fission products, strontium-90, which has a half life of 28.8 years, and caesium-137, which has a half life of 30.1 years, the result is to reduce nuclear waste lifetimes from tens of millennia (from transuranic isotopes) to a few centuries. The processes are not perfect, but the remaining transuranics are reduced from a significant problem to a tiny percentage of the total waste, because most transuranics can be used as fuel.
Fast reactors technically solve the "fuel shortage" argument against uranium-fueled reactors without assuming undiscovered reserves, or extraction from dilute sources such as granite or seawater. They permit nuclear fuels to be bred from almost all the actinides, including known, abundant sources of depleted uranium and thorium, and light-water reactor wastes. On average, more neutrons per fission are produced by fast neutrons than from thermal neutrons. This results in a larger surplus of neutrons beyond those required to sustain the chain reaction. These neutrons can be used to produce extra fuel, or to transmute long half-life waste to less troublesome isotopes, as was done at the Phénix reactor in Marcoule, France, or some can be used for each purpose. Though conventional thermal reactors also produce excess neutrons, fast reactors can produce enough of them to breed more fuel than they consume. Such designs are known as fast breeder reactors.
The main disadvantage of fast-neutron reactors is that to date they have proven costly to build and operate, and none have been proven be cost-competitive with thermal-neutron reactors unless the price of uranium increased dramatically.
Some other disadvantages are specific to some designs.
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 for long periods (notably the Phénix and EBR-II for 30 years, or the BN-600 still in operation since 1980 despite several minor leaks and fires).
Another problem is related to neutron activation. Since liquid metals other than lithium and beryllium have low moderating ability, the primary interaction of neutrons with fast reactor coolant is the (n,gamma) reaction, which induces radioactivity in the coolant. Neutron irradiation activates a significant fraction of coolant in high-power fast reactors, up to around a terabecquerel of beta decays per kilogram of coolant in steady operation. This is the reason that sodium-cooled reactors have a primary cooling loop embedded within a separate sodium pool. The sodium-24 that results from neutron capture undergoes beta decay to magnesium-24 with a half life of fifteen hours; the magnesium is removed in a cold trap.
A defective fast reactor design could have positive void coefficient: boiling of the coolant in an accident would reduce coolant density and thus the absorption rate; no such designs are proposed for commercial service. This 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 both problems, since the elastic scattering and total cross sections are approximately equal, i.e. few (n,gamma) reactions are present in the coolant and the low density of helium at typical operating conditions means that neutrons have few interactions with coolant.
Due to the low cross sections of most materials at high neutron energies, critical mass in a fast reactor is much higher than in a thermal reactor. In practice, this means significantly higher enrichment: >20% enrichment in a fast reactor compared to <5% enrichment in typical thermal reactors.
Water, the most common coolant in thermal reactors, is generally not feasible 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 and moderated system using uranium enriched to ~90%.
All operating fast reactors are liquid metal cooled reactors. The early Clementine reactor used mercury coolant and plutonium metal fuel. 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 considered as a coolant. Molten lead and lead-bismuth eutectic alloys have been used in naval propulsion units, particularly the Soviet Alpha class of submarines, as well as some prototype reactors. Sodium-potassium alloy (NaK) is popular in test reactors due to its low melting point. All large-scale fast reactors have used molten sodium coolant.
Another proposed fast reactor is a molten salt reactor, in which the salt's moderating properties are insignificant. This is typically achieved by replacing the light metal fluorides (e.g. lithium fluoride - LiF, beryllium fluoride - BeF2) in the salt carrier with heavier metal chlorides (e.g., potassium chloride - KCI, rubidium chloride - RbCl, zirconium chloride - ZrCl4). Moltex Energy proposes to build a fast-neutron reactor called the Stable Salt Reactor. In this reactor design the nuclear fuel is dissolved in a molten salt. The salt is contained in stainless steel tubes similar to those used in solid fuel reactors. The reactor is cooled using the natural convection of another molten salt coolant. Moltex claims that their design is less expensive to build than a coal-fired power plant and can consume nuclear waste from conventional solid fuel reactors.
Gas-cooled fast reactors have been the subject of research commonly using helium, which has small absorption and scattering cross sections, thus preserving the fast neutron spectrum without significant neutron absorption in the coolant.
In practice, sustaining a fission chain reaction with fast neutrons means using relatively enriched uranium or plutonium. The reason for this is that fissile reactions are favored at thermal energies, since the ratio between the fission cross section and absorption cross section is ~100 in a thermal spectrum and 8 in a fast spectrum. Fission and absorption cross sections are low for both and at high (fast) energies, which means that fast neutrons are likelier to pass through fuel without interacting than thermal neutrons; thus, more fissile material is needed. Therefore a fast reactor cannot run on natural uranium fuel. However, it is possible to build a fast reactor that breeds fuel by producing more than it consumes. After the initial fuel charge such a reactor can be refueled by reprocessing. Fission products can be replaced by adding natural or even depleted uranium without further enrichment. This is the concept of the fast breeder reactor or FBR.
So far, most fast-neutron reactors have used either MOX (mixed oxide) or metal alloy fuel. Soviet fast-neutron reactors use (high enriched) uranium fuel. The Indian prototype reactor uses 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 been proposed with enrichments in the range of 20 percent for reasons including core lifetime: if a fast reactor were loaded with the minimal critical mass, then the reactor would become subcritical after the first fission. Rather, an excess of fuel is inserted with reactivity control mechanisms, such that the reactivity control is inserted fully at the beginning of life to bring the reactor from supercritical to critical; as the fuel is depleted, the reactivity control is withdrawn to support continuing fission. In a fast breeder reactor, the above applies, though the reactivity from fuel depletion is also compensated by breeding either or and from thorium-232 or , respectively.
They cannot, however, rely on changes to their moderators because there is no moderator. So Doppler broadening in the moderator, which affects thermal neutrons, does not work, nor does a negative void coefficient of the moderator. Both techniques are common in ordinary light-water reactors.
Doppler broadening from the molecular motion of the fuel, from its heat, can provide rapid negative feedback. The molecular movement of the fissionables themselves can tune the fuel's relative speed away from the optimal neutron speed. Thermal expansion of the fuel can provide negative feedback. Small reactors as in submarines may use Doppler broadening or thermal expansion of neutron reflectors.
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.
|Past||Clementine, EBR-I/II, SEFOR, FFTF||BN-350||Dounreay, Rapsodie, Superphénix, Phénix (stopped in 2010)|
|Cancelled||Clinch River, IFR||SNR-300|
|Under construction||MBIR||PFBR, CFR-600|
|Planned||Gen IV (Gas·sodium·lead·salt), TerraPower, Elysium MCSFR, DoE VTR||BN-1200||ASTRID, Moltex||4S, JSFR, KALIMER|