Жылдам нейтронды реакторлар: Артықшылықтары мен ерекшеліктері

Nuclear reactor where fast neutrons maintain a fission chain reaction

A fast neutron reactor (FNR) or fast spectrum reactor 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 1 MeV, on average), as opposed to slow thermal neutrons used in thermal neutron reactors. Such a fast 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. Around 20 land based fast reactors have been built, accumulating over 400 reactor years of operation globally. The largest of this was the Superphénix Sodium cooled fast reactor in France that was designed to deliver 1,242 MWe. Fast reactors have been intensely studied since the 1950s, as they provide certain advantages over the existing fleet of water cooled and water moderated reactors. These are:
More neutrons are produced when a fission occurs, resulting from the absorption of a fast neutron, than the comparable process with slow (thermal, or moderated) neutrons. Thus, criticality is easier to attain than with slower neutrons. All fast reactor designs built to this date use liquid metals as coolant, such as the sodium cooled fast reactor and the lead cooled fast reactor. As the boiling points of these metals are very high, the pressure in the reactor can be maintained at a low level, which improves safety considerably. As temperatures in the core can also be substantially higher than in a water cooled design, such reactors have a greater thermodynamic efficiency; a larger percentage of the heat generated is turned into usable electricity. Atoms heavier than uranium have a much greater chance of fission with a fast neutron, than with a thermal one. This means that the inventory of heavier atoms in the nuclear waste stream, for example curium, is greatly reduced, leading to a substantially lower waste management requirement. In the GEN IV initiative, about two thirds of the proposed reactors for the future use a fast spectrum for these reasons.

Fast reactors operate by the fission of uranium and other heavy atoms, similar to thermal reactors. However, there are crucial differences, arising from the fact that by far most commercial nuclear reactors use a moderator, and fast reactors do not.

The following disadvantages of the use of a moderator have instigated the research and development of fast reactors. Although cheap, readily available and easily purified, light water can absorb a neutron and remove it from the reaction. It does this 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 concentrate the amount of in the fuel to produce enriched uranium, with the leftover known as depleted uranium. Other thermal neutron designs use different moderators, like heavy water or graphite that are much less likely to absorb neutrons, allowing them to run on natural uranium fuel. See CANDU, X 10 Graphite Reactor. In either case, the reactor's neutron economy is based on thermal neutrons. A second drawback of using water for cooling is that it has a relatively low boiling point. The vast majority of electricity production uses steam turbines. These become more efficient as the pressure (and thus the temperature) of the steam is higher. A water cooled and moderated nuclear reactor therefore needs to operate at high pressures to enable the efficient production of electricity. Thus, such reactors are constructed using very heavy steel vessels, for example 30 cm (12 inch) thick. This high pressure operation adds complexity to reactor design and requires extensive physical safety measures. The vast majority of nuclear reactors in the world are water cooled and moderated with water. Examples include the PWR, the BWR and the CANDU reactors. In Russia and the UK, reactors are operational that use graphite as moderator, and respectively water in Russian and gas in British reactors as coolant. As the operational temperature and pressure of these reactors is dictated by engineering and safety constraints, both are limited. Thus, the temperatures and pressures that can be delivered to the steam turbine are also limited. Typical water temperatures of a modern pressurized water reactor are around , with pressures of around 85 bar (1233 psi). Compared to for example modern coal fired steam circuits, where main steam temperatures in excess of are obtained, this is low, leading to a relatively low thermal efficiency. In a modern PWR, around 30 33 % of the nuclear heat is converted into electricity. A third drawback is that when a (any) nuclear reactor is shut down after operation, the fuel in the reactor no longer undergoes fission processes. However, there is an inventory present of highly radioactive elements, some of which generate substantial amounts of heat. If the fuel elements were to be exposed (i. e. there is no water to cool the elements), this heat is no longer removed. The fuel will then start to heat up, and temperatures can then exceed the melting temperature of the zircaloy cladding. When this occurs the fuel elements melt, and a meltdown occurs, such as the multiple meltdowns that occurred in the Fukushima disaster. When the reactor is in shutdown mode, the temperature and pressure are slowly reduced to atmospheric, and thus water will boil at This relatively low temperature, combined with the thickness of the steel vessels used, could lead to problems in keeping the fuel cool, as was shown by the Fukushima accident. Lastly, the fission of uranium and plutonium in a thermal spectrum yields a smaller number of neutrons than in the fast spectrum, so in a fast reactor, more losses are acceptable. The proposed fast reactors solve all of these problems (next to the fundamental fission properties, where for example plutonium 239 is more likely to fission after absorbing a fast neutron, than a slow one.)

Although and have a lower capture cross section with higher energy neutrons, they still remain reactive well into the MeV range. If the density of or is sufficient, a threshold will be reached where there are enough fissile atoms in the fuel to maintain a chain reaction with fast neutrons. In fact, in the fast spectrum, when captures a fast neutron it will also undergo fission around 11% of the time with the remainder of captures being "radiative" and entering the decay chain to plutonium 239. Crucially, when a reactor runs on fast neutrons, the isotope is likely to fission 74% of the time instead of the 62% of fissions when it captures a thermal neutron. In addition the probability of a upon absorbing a fast neutron fissioning is 70% while for a thermal neutron it is less than 20%. Fast neutrons have a smaller chance of being captured by the uranium and plutonium, but when they are captured, have a significantly higher probability of causing a fission. The inventory of spent fast reactor fuel therefore contains virtually no actinides except for uranium and plutonium, which can be effectively recycled. Even when the core is initially loaded with 20% mass reactor grade plutonium (containing on average 2% , 53% , 25% , 15% , 5% and traces of ), the fast spectrum neutrons are capable of causing each of these to fission at significant rates. By the end of a fuel cycle of some 24 months, these ratios will have shifted with an increase of to over 80% while all the other plutonium isotopes will have decreased in proportion. By removing the moderator, the size of the reactor core volume can be greatly reduced, and to some extent the complexity. As and particularly are far more likely to fission when they capture a fast neutron, it is possible to fuel such reactors with a mixture of plutonium and natural uranium, or with enriched material, containing around 20% Test runs at various facilities have also been done using and The natural uranium (mostly ) will be turned into , while in the case of , is the result. As new fuel is created during the operation, this process is called breeding. Less than 1% of the total Uranium mined is consumed in a thermal once through cycle, while up to 60% of the natural uranium is fissioned in the best existing fast reactor cycles. Given the current inventory of spent nuclear fuel (which contains reactor grade plutonium), it is possible to process this spent fuel material and reuse the actinide isotopes as fuel in a large number of fast reactors. This effectively consumes the , reactor grade plutonium, , and Enormous amounts of energy are still present in the spent reactor fuel inventories; if fast reactor types were to be employed to use this material, that energy can be extracted for useful purposes.

All nuclear reactors produce heat which must be removed from the reactor core. Water, the most common coolant in thermal reactors, is generally not feasible for a fast reactor, because it acts as an effective neutron moderator. All operating fast reactors are liquid metal cooled reactors, which use sodium, lead, or lead bismuth eutectic as coolants. The early Clementine reactor used mercury coolant and plutonium metal fuel. In addition to its toxicity to humans, mercury has a high capture cross section (thus, it readily absorbs the neutrons, which causes nuclear reactions) 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 useful as a coolant. Russia has developed reactors that use molten lead and lead bismuth eutectic alloys, which have been used on a larger scale in naval propulsion units, particularly the Soviet Alfa class submarine, as well as some prototype reactors. Sodium potassium alloy (NaK) is popular in test reactors due to its low melting point. Another proposed fast reactor is a molten salt reactor, in which the salt's moderating properties are insignificant. The particular salt formula used is crucial as some formulas are effective moderators while others are not. 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. Purified nitrogen 15 has also been proposed as a coolant gas because it is more common than Helium and also has a very low neutron absorption cross section. However, all large scale fast reactors have used molten metal coolant. Advantages of molten metals are low cost, the small activation potential and the large liquid ranges. The latter means that the material has a low melting point, and a high boiling point. Examples of these reactors include Sodium cooled fast reactor, which are still being pursued worldwide. Russia currently operates two such reactors on a commercial scale. Additionally, Russia has around eighty reactor years of experience with the Lead cooled fast reactor which is rapidly gaining interest.

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 used (highly enriched) uranium fuel initially, then in 2022 switched to using MOX. The Indian prototype reactor uses uranium carbide fuel. While criticality at fast energies may be achieved with uranium enriched to 5.5 (weight) percent , fast reactor designs have been proposed with enrichment 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 or , respectively. Some designs use Burnable Poisons also known as Burnable Absorbers which contain isotopes with high neutron capture cross sections. Concentrated or & in natural Gadolinium are typically used for this purpose. As these isotopes absorb excess neutrons they are transmuted into isotopes with low absorption cross sections so that over the life of the fuel cycle they are eliminated as more fission products with high capture cross section are generated. This makes it easier to maintain control of the reactivity rate in the core at start up with fresh fuel.

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 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.

As the perception of the reserves of uranium ore in the 1960s was rather low, 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, meaning that existing ore sources would last hundreds of years. The disadvantage to this approach is that the breeder reactor has to be fed fuel that must be treated in a spent fuel treatment plant. 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 unfeasible.

As most fast reactors to date have been either sodium, lead or lead bismuth cooled, the disadvantages of such systems are described here. As a result of running the reactors on fast neutrons, the reactivity of the core is determined by these neutrons, as opposed to moderated reactors. In the moderated reactors, a significant amount of control of the reactivity is obtained from delayed neutrons, which allow time for operators or computers to adjust reactivity. As delayed neutrons play virtually no role in fast reactors, other mechanisms are required for the very short term reactivity control (e. g. within one second) in fast reactors, which are thermal expansion and Doppler broadening. Longer term reactivity is obtained from control rods, which are filled with a neutron absorption material. As the entire reactor is filled with large volumes of molten metal, refuelling is not trivial, as optical tools (cameras, etc.) are of no use. Costly, carefully calibrated and positioned robotic tools are needed for the operation of refueling. Also, completely removing fuel elements from the reactor is not easy. The fact that the entire reactor is filled with a metal that has a melting point much higher than room temperature, all the tubing, heat exchangers, and the entire reactor volume must be heated electrically, before any nuclear operation can take place. However, once the reactor produces heat, this is no longer of any concern. To date most fast reactor types have proven costly to build and operate, and are not very competitive with thermal neutron reactors unless the price of uranium increased dramatically, or building costs decreased. It is thought that given the perception of problematic nuclear waste disposal, such reactors will be necessary. As moderated reactor construction costs are rising (among other) due to ever more stringent safety mechanisms, this could mean a better economic viability of fast reactors. 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, although the combustion reaction of sodium in air should not be confused with the extremely violent reaction of sodium and water. Sodium leaks can ignite with air, causing difficulties in reactors such as (e. g. USS Seawolf (SSN 575) and Monju). Some sodium cooled fast reactors have operated safely for long periods (notably the Phénix and EBR II for 30 years, or the BN 600 and BN 800 in operation since resp. 1980 and 2016, despite several minor leaks and fires. It is important to note that sodium leaks (and possibly fires) do not release radioactive elements, as the sodium fast reactors are always designed with a two loop system. 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. Sodium 24 is created in the reactor loop of the sodium cooled fast reactor, from natural sodium 23 by neutron bombardment. With a 15 hour half life, decays to link=Isotopes of magnesium24Mg by emission of an electron and two gamma rays. As the half life of this isotope is very short, after e. g. two weeks, almost no is left. Fast spectrum reactors that use sodium must remove this magnesium from the sodium, which is achieved with a 'cold' trap. From the liquid lead or Lead bismuth eutectic designs, only the liquid eutectic lead bismuth will have activation. As pure lead will have virtually no activation, a pure lead reactor design could operate in a single loop, saving significant costs on heat exchanger and separate systems. 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, as they are potentially 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, reactivity control in a gas cooled fast reactor is difficult. 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. Alternatively, a mixture of plutonium from nuclear waste, combined with natural or depleted uranium could be used.

Clementine was the first fast reactor, built in 1946 at Los Alamos National Laboratory. It used plutonium metal fuel, mercury coolant, achieved 25 kW thermal and used for research, especially as a fast neutron source. Experimental Breeder Reactor I (EBR I) at Argonne West, now Idaho National Laboratory, near Arco, Idaho, in 1951 became the first reactor to generate significant amounts of power. Decommissioned in 1964. Fermi 1 near Detroit was a prototype fast breeder reactor that powered up in 1957 and shut down in 1972. Experimental Breeder Reactor II (EBR II) at Idaho National Laboratory, near Arco, Idaho, was a prototype for the Integral Fast Reactor, 1965–1994. SEFOR in Arkansas, was a 20 MWt research reactor that operated from 1969 to 1972. Fast Flux Test Facility (FFTF), 400 MWt, operated flawlessly from 1982 to 1992, at Hanford Washington. It used liquid sodium drained with argon backfill under care and maintenance. SRE in California, was a 20 MWt, 6.5 MWe commercial reactor operated from 1957 to 1964. LAMPRE 1 was a molten plutonium fueled 1 MWth reactor. It operated as a research reactor from 1961 to 1963 at Los Alamos national Lab.

Dounreay Loop type Fast Reactor (DFR), 1959–1977, was a 14 MWe and Prototype Fast Reactor (PFR), 1974–1994, 250 MWe, in Caithness, in the Highland area of Scotland. Dounreay Pool type Fast Reactor (PFR), 1975–1994, was a 600 MWt, 234 MWe which used mixed oxide (MOX) fuel. Rapsodie in Cadarache, France, (20 then 40 MW) operated between 1967 and 1982. Superphénix, in France, 1200 MWe, closed in 1997 due to a political decision and high costs. 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, in Germany a 21 MWe experimental compact sodium cooled fast reactor operated from Oct 1977 Aug 1991. The objective of the experiment was to eliminate nuclear waste while producing energy. There were minor sodium problems combined with public protests which resulted in the closure of the facility.

Small lead cooled fast reactors were used for naval propulsion, particularly by the Soviet Navy. BR 5 was a research focused fast neutron reactor at the Institute of Physics and Energy in Obninsk from 1959 to 2002. BN 350 was constructed by the Soviet Union in Shevchenko (today's Aqtau) on the Caspian Sea, It produced 130 MWe plus 80,000 tons of fresh water per day. IBR was a research focused pulsed fast neutron reactor at the Joint Institute of Nuclear Research in Dubna (near Moscow) put into operation in 1984 till 2006. RORSATs 33 space fast reactors were launched by the Soviet Union from 1989 to 1990 as part of a program known as the Radar Ocean Reconnaissance Satellite (RORSAT) in the US. Typically, the reactors produced approximately 3 kWe. BES 5 was a sodium cooled space reactor launched as part of the RORSAT program which produced 5 kWe. BR 5 was a 5 MWt sodium fast reactor operated by the USSR in 1961 primarily for materials testing. Russian Alpha 8 PbBi was a series of lead bismuth cooled fast reactors used aboard submarines. The submarines functioned as killer submarines, staying in harbor then attacking due to the high speeds achievable by the sub.

Monju reactor, 300 MWe, in Japan, was closed in 1995 following a serious sodium leak and fire. It was restarted on May 6, 2010, but in August 2010 another accident, involving dropped machinery, shut down the reactor again. As of June 2011, the reactor had generated electricity for only one hour since its first test two decades prior. Aktau Reactor, 150 MWe, in Kazakhstan, was used for plutonium production, desalination, and electricity. It closed 4 years after the plant's operating license expired.

IBR 2 a pulsed fast neutron reactor at Joint Institute for Nuclear Research in Dubna. Located at the Laboratory of Neutron Physics (FLNP) named after. I. M. Frank as part of JINR. From 2006 to June 2011 it underwent modernization. The only nuclear reactor in the world with a movable reflector. Commissioned on February 10, 1984. BN 600 a pool type sodium cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. It provides 560 MWe to the Middle Urals power grid. In operation since 1980. BN 800 a sodium cooled fast breeder reactor at the Beloyarsk Nuclear Power Station. It generates 880 MW of electrical power and started producing electricity in October, 2014. It reached full power in August, 2016. BOR 60 a sodium cooled reactor at the Research Institute of Atomic Reactors in Dimitrovgrad, Russia. In operation since 1968. It produces 60MW for experimental purposes. FBTR a 40MWt,13.2MWe experimental reactor in India which focused on reaching significant burnup levels. China Experimental Fast Reactor, a 60 MWth, 20 MWe, experimental reactor which went critical in 2011 and is currently operational. It is used for materials and component research for future Chinese fast reactors. KiloPower/KRUSTY is a 1 10 kWe research sodium fast reactor built at Los Alamos National Laboratory. It first reach criticality in 2015 and demonstrates an application of a Stirling power cycle.

Jōyō (nuclear reactor) (Jōyō), 1977–1997 and 2004–2007, Japan, 140 MWt is an experimental reactor, operated as an irradiation test facility. After an incident in 2007, the reactor was suspended for repairing, recoworks were planned to be completed in 2014.

PFBR, Kalpakkam, India, 500 MWe reactor with criticality planned for 2021. It is a sodium fast breeder reactor. CFR 600, China, 600 MWe. MBIR Multipurpose fast neutron research reactor. The Research Institute of Atomic Reactors (NIIAR) site at Dimitrovgrad in the Ulyanovsk region of western Russia, 150 MWt. Construction started in 2016 with completion scheduled for 2024. BREST 300, Seversk, Russia. Construction started at 8 June 2021

+ Fast reactors U. S. Russia Europe AsiaPast Clementine, EBR I/II, SEFOR, FFTF BN 350 Dounreay, Rapsodie, Superphénix, Phénix (stopped in 2010)Cancelled Clinch River, IFR SNR 300, ASTRIDUnder decommissioningMonjuOperating IBR 2, BOR 60, BN 600, BN 800 FBTR, CEFRUnder repair Jōyō Under construction MBIR, BREST 300 PFBR, CFR 600Planned Gen IV (Gas·sodium·lead·salt), TerraPower, Elysium MCSFR, DoE VTR BN 1200 Moltex, ALFRED 4S, JSFR, KALIMER