Neutron moderator

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https://en.wikipedia.org/wiki/Neutron_moderator 

 

Science with neutrons

In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, ideally without capturing any, leaving them as thermal neutrons with only minimal (thermal) kinetic energy. These thermal neutrons are immensely more susceptible than fast neutrons to propagate a nuclear chain reaction of uranium-235 or other fissile isotope by colliding with their atomic nucleus.

Water (sometimes called "light water" in this context) is the most commonly used moderator (roughly 75% of the world's reactors). Solid graphite (20% of reactors) and heavy water (5% of reactors) are the main alternatives. Beryllium has also been used in some experimental types, and hydrocarbons have been suggested as another possibility.

Moderation

Neutrons are normally bound into an atomic nucleus, and do not exist free for long in nature. The unbound neutron has a half-life of 10 minutes and 11 seconds. The release of neutrons from the nucleus requires exceeding the binding energy of the neutron, which is typically 7-9 MeV for most isotopes. Neutron sources generate free neutrons by a variety of nuclear reactions, including nuclear fission and nuclear fusion. Whatever the source of neutrons, they are released with energies of several MeV.

According to the equipartition theorem, the average kinetic energy, ${\displaystyle \overline{E}}$, can be related to temperature, ${\displaystyle T}$, via:

${\displaystyle \overline{E}=\frac{1}{2}{m}_n⟨v^2⟩=\frac{3}{2}{k}_{B}T}$,

where ${\displaystyle m_n}$ is the neutron mass, ${\displaystyle ⟨v^2⟩}$ is the average squared neutron speed, and ${\displaystyle k_B}$ is the Boltzmann constant. The characteristic neutron temperature of several-MeV neutrons is several tens of billions kelvin.

Moderation is the process of the reduction of the initial high speed (high kinetic energy) of the free neutron. Since energy is conserved, this reduction of the neutron speed takes place by transfer of energy to a material called a moderator.

The probability of scattering of a neutron from a nucleus is given by the scattering cross section. The first couple of collisions with the moderator may be of sufficiently high energy to excite the nucleus of the moderator. Such a collision is inelastic, since some of the kinetic energy is transformed to potential energy by exciting some of the internal degrees of freedom of the nucleus to form an excited state. As the energy of the neutron is lowered, the collisions become predominantly elastic, i.e., the total kinetic energy and momentum of the system (that of the neutron and the nucleus) is conserved.

Given the mathematics of elastic collisions, as neutrons are very light compared to most nuclei, the most efficient way of removing kinetic energy from the neutron is by choosing a moderating nucleus that has near identical mass.

Elastic collision of equal masses

A collision of a neutron, which has mass of 1, with a ¹H nucleus (a proton) could result in the neutron losing virtually all of its energy in a single head-on collision. More generally, it is necessary to take into account both glancing and head-on collisions. The mean logarithmic reduction of neutron energy per collision, ${\displaystyle \xi }$, depends only on the atomic mass, ${\displaystyle A}$, of the nucleus and is given by:

${\displaystyle \xi =\ln \frac{E_0}{E}=1+\frac{(A-1{)}^2}{2A}\ln \left(\frac{A-1}{A+1}\right)}$.

This can be reasonably approximated to the very simple form ${\displaystyle \xi \simeq \frac{2}{A+2/3}}$. From this one can deduce ${\displaystyle n}$, the expected number of collisions of the neutron with nuclei of a given type that is required to reduce the kinetic energy of a neutron from ${\displaystyle E_0}$ to ${\displaystyle E_1}$

${\displaystyle n=\frac{1}{\xi }(\ln E_0-\ln E_1)}$.

In a system at thermal equilibrium, neutrons (red) are elastically scattered by a hypothetical moderator of free hydrogen nuclei (blue), undergoing thermally activated motion. Kinetic energy is transferred between particles. As the neutrons have essentially the same mass as protons and there is no absorption, the velocity distributions of both particles types would be well-described by a single Maxwell–Boltzmann distribution.

Choice of moderator materials

Some nuclei have larger absorption cross sections than others, which removes free neutrons from the flux. Therefore, a further criterion for an efficient moderator is one for which this parameter is small. The moderating efficiency gives the ratio of the macroscopic cross sections of scattering, ${\displaystyle {\mathrm{\Sigma }}_s}$, weighted by ${\displaystyle \xi }$ divided by that of absorption, ${\displaystyle {\mathrm{\Sigma }}_a}$: i.e., ${\displaystyle \frac{\xi {\mathrm{\Sigma }}_s}{{\mathrm{\Sigma }}_a}}$. For a compound moderator composed of more than one element, such as light or heavy water, it is necessary to take into account the moderating and absorbing effect of both the hydrogen isotope and oxygen atom to calculate ${\displaystyle \xi }$. To bring a neutron from the fission energy of ${\displaystyle E_0}$ 2 MeV to an ${\displaystyle E}$ of 1 eV takes an expected ${\displaystyle n}$ of 16 and 29 collisions for H₂O and D₂O, respectively. Therefore, neutrons are more rapidly moderated by light water, as H has a far higher ${\displaystyle {\mathrm{\Sigma }}_s}$. However, it also has a far higher ${\displaystyle {\mathrm{\Sigma }}_a}$, so that the moderating efficiency is nearly 80 times higher for heavy water than for light water.

The ideal moderator is of low mass, high scattering cross section, and low absorption cross section.

	Hydrogen	Deuterium	Beryllium	Carbon	Oxygen	Uranium
Mass of kernels u	1	2	9	12	16	238
Energy decrement ${\displaystyle \xi }$	1	0.7261	0.2078	0.1589	0.1209	0.0084
Number of Collisions	18	25	86	114	150	2172

Distribution of neutron velocities once moderated

After sufficient impacts, the speed of the neutron will be comparable to the speed of the nuclei given by thermal motion; this neutron is then called a thermal neutron, and the process may also be termed thermalization. Once at equilibrium at a given temperature the distribution of speeds (energies) expected of rigid spheres scattering elastically is given by the Maxwell–Boltzmann distribution. This is only slightly modified in a real moderator due to the speed (energy) dependence of the absorption cross-section of most materials, so that low-speed neutrons are preferentially absorbed, so that the true neutron velocity distribution in the core would be slightly hotter than predicted.

Reactor moderators

In a thermal-neutron reactor, the nucleus of a heavy fuel element such as uranium absorbs a slow-moving free neutron, becomes unstable, and then splits ("fissions") into two smaller atoms ("fission products"). The fission process for ²³⁵U nuclei yields two fission products, two to three fast-moving free neutrons, plus an amount of energy primarily manifested in the kinetic energy of the recoiling fission products. The free neutrons are emitted with a kinetic energy of ~2 MeV each. Because more free neutrons are released from a uranium fission event than thermal neutrons are required to initiate the event, the reaction can become self-sustaining – a chain reaction – under controlled conditions, thus liberating a tremendous amount of energy (see article nuclear fission).

Fission cross section, measured in barns (a unit equal to 10⁻²⁸ m²), is a function of the energy (so-called excitation function) of the neutron colliding with a ²³⁵U nucleus. Fission probability decreases as neutron energy (and speed) increases. This explains why most reactors fueled with ²³⁵U need a moderator to sustain a chain reaction and why removing a moderator can shut down a reactor.

The probability of further fission events is determined by the fission cross section, which is dependent upon the speed (energy) of the incident neutrons. For thermal reactors, high-energy neutrons in the MeV-range are much less likely (though not unable) to cause further fission. The newly released fast neutrons, moving at roughly 10% of the speed of light, must be slowed down or "moderated", typically to speeds of a few kilometres per second, if they are to be likely to cause further fission in neighbouring ²³⁵U nuclei and hence continue the chain reaction. This speed happens to be equivalent to temperatures in the few hundred Celsius range.

In all moderated reactors, some neutrons of all energy levels will produce fission, including fast neutrons. Some reactors are more fully thermalised than others; for example, in a CANDU reactor nearly all fission reactions are produced by thermal neutrons, while in a pressurized water reactor (PWR) a considerable portion of the fissions are produced by higher-energy neutrons. In the proposed water-cooled supercritical water reactor (SCWR), the proportion of fast fissions may exceed 50%, making it technically a fast-neutron reactor.

A fast reactor uses no moderator, but relies on fission produced by unmoderated fast neutrons to sustain the chain reaction. In some fast reactor designs, up to 20% of fissions can come from direct fast neutron fission of uranium-238, an isotope which is not fissile at all with thermal neutrons.

Moderators are also used in non-reactor neutron sources, such as plutonium-beryllium (using the ⁹
Be(α,n)¹²
C reaction) and spallation sources (using (p,xn) reactions with neutron rich heavy elements as targets).

Form and location

The form and location of the moderator can greatly influence the cost and safety of a reactor. Classically, moderators were precision-machined blocks of high purity graphite with embedded ducting to carry away heat. They were in the hottest part of the reactor, and therefore subject to corrosion and ablation. In some materials, including graphite, the impact of the neutrons with the moderator can cause the moderator to accumulate dangerous amounts of Wigner energy. This problem led to the infamous Windscale fire at the Windscale Piles, a nuclear reactor complex in the United Kingdom, in 1957. In a carbon dioxide cooled graphite moderated reactor where coolant and moderator are in contact with one another, the Boudouard reaction needs to be taken into account. This is also the case if fuel elements have an outer layer of carbon - as in some TRISO fuels - or if an inner carbon layer becomes exposed by failure of one or several outer layers.

The moderators of some pebble-bed reactors are not only simple, but also inexpensive: the nuclear fuel is embedded in spheres of reactor-grade pyrolytic carbon, roughly of the size of tennis balls. The spaces between the balls serve as ducting. The reactor is operated above the Wigner annealing temperature so that the graphite does not accumulate dangerous amounts of Wigner energy.

In CANDU and PWR reactors, the moderator is liquid water (heavy water for CANDU, light water for PWR). In the event of a loss-of-coolant accident in a PWR, the moderator is also lost and the reaction will stop. This negative void coefficient is an important safety feature of these reactors. In CANDU the moderator is located in a separate heavy-water circuit, surrounding the pressurized heavy-water coolant channels. The heavy water will slow down a significant portion of neutrons to the resonance integral of ²³⁸
U increasing the neutron capture in this isotope that makes up over 99% of the uranium in CANDU fuel thus decreasing the amount of neutrons available for fission. As a consequence, removing some of the heavy water will increase reactivity until so much is removed that too little moderation is provided to keep the reaction going. This design gives CANDU reactors a positive void coefficient, although the slower neutron kinetics of heavy-water moderated systems compensates for this, leading to comparable safety with PWRs.^[9] In the light water cooled graphite moderated RBMK, a reactor type originally envisioned to allow both production of weapons grade plutonium and large amounts of usable heat while using natural uranium and foregoing the use of heavy water, the light water coolant acts primarily as a neutron absorber and thus its removal in a Loss of Coolant Accident or by conversion of water into steam will increase the amount of thermal neutrons available for fission. Following the Chernobyl nuclear accident the issue was remedied so that all still operating RBMK type reactors have a slightly negative void coefficient but they now require a higher degree of uranium enrichment in their fuel.

Moderator impurities

Good moderators are free of neutron-absorbing impurities such as boron. In commercial nuclear power plants the moderator typically contains dissolved boron. The boron concentration of the reactor coolant can be changed by the operators by adding boric acid or by diluting with water to manipulate reactor power. The Nazi Nuclear Program suffered a substantial setback when its inexpensive graphite moderators failed to function. At that time, most graphites were deposited onto boron electrodes, and the German commercial graphite contained too much boron. Since the war-time German program never discovered this problem, they were forced to use far more expensive heavy water moderators. This problem was discovered by famous physicist Leó Szilárd.

Non-graphite moderators

Some moderators are quite expensive, for example beryllium, and reactor-grade heavy water. Reactor-grade heavy water must be 99.75% pure to enable reactions with unenriched uranium. This is difficult to prepare because heavy water and regular water form the same chemical bonds in almost the same ways, at only slightly different speeds.

The much cheaper light water moderator (essentially very pure regular water) absorbs too many neutrons to be used with unenriched natural uranium, and therefore uranium enrichment or nuclear reprocessing becomes necessary to operate such reactors, increasing overall costs. Both enrichment and reprocessing are expensive and technologically challenging processes, and additionally both enrichment and several types of reprocessing can be used to create weapons-usable material, causing proliferation concerns. Reprocessing schemes that are more resistant to proliferation are currently under development.

The CANDU reactor's moderator doubles as a safety feature. A large tank of low-temperature, low-pressure heavy water moderates the neutrons and also acts as a heat sink in extreme loss-of-coolant accident conditions. It is separated from the fuel rods that actually generate the heat. Heavy water is very effective at slowing down (moderating) neutrons, giving CANDU reactors their important and defining characteristic of high "neutron economy". Unlike a light water reactor where adding water to the core in an accident might provide enough moderation to make a subcritical assembly go critical again, heavy water reactors will decrease their reactivity if light water is added to the core, which provides another important safety feature in the case of certain accident scenarios. However, any heavy water that becomes mixed with the emergency coolant light water will become too diluted to be useful without isotope separation.

Nuclear weapon design

Main article: Uranium hydride bomb

Early speculation about nuclear weapons assumed that an "atom bomb" would be a large amount of fissile material, moderated by a neutron moderator, similar in structure to a nuclear reactor or "pile". Only the Manhattan project embraced the idea of a chain reaction of fast neutrons in pure metallic uranium or plutonium. Other moderated designs were also considered by the Americans; proposals included using uranium deuteride as the fissile material. In 1943 Robert Oppenheimer and Niels Bohr considered the possibility of using a "pile" as a weapon. The motivation was that with a graphite moderator it would be possible to achieve the chain reaction without the use of any isotope separation. However, plutonium can be produced ("bred") sufficiently isotopically pure as to be usable in a bomb and then has to be "only" separated chemically, a much easier processes than isotope separation, albeit still a challenging one. In August 1945, when information of the atomic bombing of Hiroshima was relayed to the scientists of the German nuclear program, interred at Farm Hall in England, chief scientist Werner Heisenberg hypothesized that the device must have been "something like a nuclear reactor, with the neutrons slowed by many collisions with a moderator". The German program, which had been much less advanced, had never even considered the plutonium-option and didn't discover a feasible method of large scale isotope separation in uranium.

After the success of the Manhattan project, all major nuclear weapons programs have relied on fast neutrons in their weapons designs. The notable exception is the Ruth and Ray test explosions of Operation Upshot–Knothole. The aim of the University of California Radiation Laboratory designs was the exploration of deuterated polyethylene charge containing uranium as a candidate thermonuclear fuel, hoping that deuterium would fuse (becoming an active medium) if compressed appropriately. If successful, the devices could also lead to a compact primary containing minimal amount of fissile material, and powerful enough to ignite RAMROD a thermonuclear weapon designed by UCRL at the time. For a "hydride" primary, the degree of compression would not make deuterium to fuse, but the design could be subjected to boosting, raising the yield considerably. The cores consisted of a mix of uranium deuteride (UD₃), and deuterated polyethylene. The core tested in Ray used uranium low enriched in U²³⁵, and in both shots deuterium acted as the neutron moderator. The predicted yield was 1.5 to 3 kt for Ruth (with a maximum potential yield of 20 kt) and 0.5-1 kt for Ray. The tests produced yields of 200 tons of TNT each; both tests were considered to be fizzles.

The main benefit of using a moderator in a nuclear explosive is that the amount of fissile material needed to reach criticality may be greatly reduced. Slowing of fast neutrons will increase the cross section for neutron absorption, reducing the critical mass. A side effect is however that as the chain reaction progresses, the moderator will be heated, thus losing its ability to cool the neutrons.

Another effect of moderation is that the time between subsequent neutron generations is increased, slowing down the reaction. This makes the containment of the explosion a problem; the inertia that is used to confine implosion type bombs will not be able to confine the reaction. The end result may be a fizzle instead of a bang.

The explosive power of a fully moderated explosion is thus limited, at worst it may be equal to a chemical explosive of similar mass. Again quoting Heisenberg: "One can never make an explosive with slow neutrons, not even with the heavy water machine, as then the neutrons only go with thermal speed, with the result that the reaction is so slow that the thing explodes sooner, before the reaction is complete."

While a nuclear bomb working on thermal neutrons may be impractical, modern weapons designs may still benefit from some level of moderation. A beryllium tamper used as a neutron reflector will also act as a moderator.

Materials used

• Hydrogen, as in ordinary "light water". Because protium also has a significant cross section for neutron capture only limited moderation is possible without losing too many neutrons. The less-moderated neutrons are relatively more likely to be captured by uranium-238 and less likely to fission uranium-235, so light water reactors require enriched uranium to operate.
  • There are also proposals to use the compound formed by the chemical reaction of metallic uranium and hydrogen (uranium hydride—UH₃) as a combination fuel and moderator in a new type of reactor.
  • Hydrogen is also used in the form of cryogenic liquid methane and sometimes liquid hydrogen as a cold neutron source in some research reactors: yielding a Maxwell–Boltzmann distribution for the neutrons whose maximum is shifted to much lower energies.
  • Hydrogen combined with carbon as in paraffin wax was used in some early German experiments.
• Deuterium, in the form of heavy water, in heavy water reactors, e.g. CANDU. Reactors moderated with heavy water can use unenriched natural uranium.
• Carbon, in the form of reactor-grade graphite or pyrolytic carbon, used in e.g. RBMK and pebble-bed reactors, or in compounds, e.g. carbon dioxide. As carbon dioxide contains twice as many oxygen atoms as it does carbon atoms and both have moderating and neutron absorbing effects in a similar range (see above), a significant share of the moderation in a (yet to be built) carbon dioxide moderated reactor would actually come from the oxygen. Lower-temperature reactors are susceptible to buildup of Wigner energy in the material. Like deuterium-moderated reactors, some of these reactors can use unenriched natural uranium.
  • Graphite is also deliberately allowed to be heated to around 2000 K or higher in some research reactors to produce a hot neutron source: giving a Maxwell–Boltzmann distribution whose maximum is spread out to generate higher energy neutrons.
• Beryllium, in the form of metal. Beryllium is expensive and toxic, so its use is limited.
• Lithium-7, in the form of a lithium fluoride salt, typically in conjunction with beryllium fluoride salt (FLiBe). This is the most common type of moderator in a molten salt reactor.

Other light-nuclei materials are unsuitable for various reasons. Helium is a gas and it requires special design to achieve sufficient density; lithium-6 and boron-10 absorb neutrons.

Currently operating nuclear power reactors by moderator

Moderator	Reactors	Design	Country
none (fast)	2	BN-600, BN-800	Russia (2)
graphite	25	AGR, Magnox, RBMK	United Kingdom (14), Russia (9)
heavy water	29	CANDU, PHWR	Canada (17), South Korea (4), Romania (2), China (2), India (18), Argentina, Pakistan
light water	359	PWR, BWR	27 countries

Breeder reactor

From Wikipedia, the free encyclopedia

https://en.wikipedia.org/wiki/Breeder_reactor 

 

Assembly of the core of Experimental Breeder Reactor I in Idaho, United States, 1951

A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. These reactors can be fuelled with more commonly available isotopes of uranium and thorium, such as uranium-238 or thorium-232, as opposed to the rare uranium-235 which is used in conventional reactors. These materials are called fertile materials since they can be bred into fuel by these breeder reactors.

Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use. These extra neutrons are absorbed by the fertile material that is loaded into the reactor along with fissile fuel. This irradiated fertile material in turns transmutes into fissile material which can undergo fission reactions.

Breeders were at first found attractive because they made more complete use of uranium fuel than light-water reactors, but interest declined after the 1960s, as more uranium reserves were found, and new methods of uranium enrichment reduced fuel costs.

Fuel resources

Breeder reactors could, in principle, extract almost all of the energy contained in uranium or thorium, decreasing fuel requirements by a factor of 100 compared to widely used once-through light water reactors, which extract less than 1% of the energy in the uranium mined from the earth. The high fuel-efficiency of breeder reactors could greatly reduce concerns about fuel supply, energy used in mining, and storage of radioactive waste. With seawater uranium extraction (currently too expensive to be economical), there is enough fuel for breeder reactors to satisfy the world's energy needs for 5 billion years at 1983's total energy consumption rate, thus making nuclear energy effectively a renewable energy. In addition to seawater the average crustal granite rocks contain significant quantities of Uranium and Thorium that with breeder reactors can supply abundant energy for the remaining lifespan of the sun on the main sequence of stellar evolution.

Fuel efficiency and types of nuclear waste

Nuclear waste became a greater concern by the 1990s. In broad terms, spent nuclear fuel has three main components. The first consists of fission products, the leftover fragments of fuel atoms after they have been split to release energy. Fission products come in dozens of elements and hundreds of isotopes, all of them lighter than uranium. The second main component of spent fuel is transuranics (atoms heavier than uranium), which are generated from uranium or heavier atoms in the fuel when they absorb neutrons but do not undergo fission. All transuranic isotopes fall within the actinide series on the periodic table, and so they are frequently referred to as the actinides. The largest component is the remaining Uranium which is around 98.25 % Uranium-238, 1.1% Uranium-235 and 0.65% Uranium-236. The U-236 comes from the non-fission capture reaction where U-235 absorbs a neutron but releases only a high energy gamma ray instead of undergoing fission.

The physical behavior of the fission products is markedly different from that of the Actinides. In particular, fission products do not themselves undergo fission, and therefore cannot be used as nuclear fuel, either for nuclear weapons or nuclear reactors. Indeed, because fission products are often neutron poisons (absorbing neutrons that could be used to sustain a chain reaction), fission products are viewed as nuclear 'ashes' left over from consuming fissile materials. Furthermore, only seven long-lived fission product isotopes have half-lives longer than a hundred years, which makes their geological storage or disposal less problematic than for transuranic materials.

With increased concerns about nuclear waste, breeding fuel cycles came under renewed interest as they can reduce actinide wastes, particularly plutonium and minor actinides. Breeder reactors are designed to fission the actinide wastes as fuel, and thus convert them to more fission products.

After spent nuclear fuel is removed from a light water reactor, it undergoes a complex decay profile as each nuclide decays at a different rate. Due to a physical oddity referenced below, there is a large gap in the decay half-lives of fission products compared to transuranic isotopes. If the transuranics are left in the spent fuel, after 1,000 to 100,000 years, the slow decay of these transuranics would generate most of the radioactivity in that spent fuel. Thus, removing the transuranics from the waste eliminates much of the long-term radioactivity of spent nuclear fuel.

Fission probabilities of selected actinides, thermal vs. fast neutrons

Isotope	Thermal fission cross section	Thermal fission %	Fast fission cross section	Fast fission %
Th-232	nil	1 n	0.350 barn	3 n
U-232	76.66 barn	59	2.370 barn	95
U-233	531.2 barn	89	2.450 barn	93
U-235	584.4 barn	81	2.056 barn	80
U-238	11.77 microbarn	1 n	1.136 barn	11
Np-237	0.02249 barn	3 n	2.247 barn	27
Pu-238	17.89 barn	7	2.721 barn	70
Pu-239	747.4 barn	63	2.338 barn	85
Pu-240	58.77 barn	1 n	2.253 barn	55
Pu-241	1012 barn	75	2.298 barn	87
Pu-242	0.002557 barn	1 n	2.027 barn	53
Am-241	600.4 barn	1 n	0.2299 microbarn	21
Am-242m	6409 barn	75	2.550 barn	94
Am-243	0.1161 barn	1 n	2.140 barn	23
Cm-242	5.064 barn	1 n	2.907 barn	10
Cm-243	617.4 barn	78	2.500 barn	94
Cm-244	1.037 barn	4 n	0.08255 microbarn	33
n=non-fissile

Today's commercial light water reactors do breed some new fissile material, mostly in the form of plutonium. Because commercial reactors were never designed as breeders, they do not convert enough uranium-238 into plutonium to replace the uranium-235 consumed. Nonetheless, at least one-third of the power produced by commercial nuclear reactors comes from fission of plutonium generated within the fuel. Even with this level of plutonium consumption, light water reactors consume only part of the plutonium and minor actinides they produce, and nonfissile isotopes of plutonium build up, along with significant quantities of other minor actinides.

Conversion ratio, break-even, breeding ratio, doubling time, and burnup

One measure of a reactor's performance is the "conversion ratio", defined as the ratio of new fissile atoms produced to fissile atoms consumed. All proposed nuclear reactors except specially designed and operated actinide burners experience some degree of conversion. As long as there is any amount of a fertile material within the neutron flux of the reactor, some new fissile material is always created. When the conversion ratio is greater than 1, it is often called the "breeding ratio".

For example, commonly used light water reactors have a conversion ratio of approximately 0.6. Pressurized heavy-water reactors (PHWR) running on natural uranium have a conversion ratio of 0.8. In a breeder reactor, the conversion ratio is higher than 1. "Break-even" is achieved when the conversion ratio reaches 1.0 and the reactor produces as much fissile material as it uses.

The doubling time is the amount of time it would take for a breeder reactor to produce enough new fissile material to replace the original fuel and additionally produce an equivalent amount of fuel for another nuclear reactor. This was considered an important measure of breeder performance in early years, when uranium was thought to be scarce. However, since uranium is more abundant than thought in the early days of nuclear reactor development, and given the amount of plutonium available in spent reactor fuel, doubling time has become a less important metric in modern breeder-reactor design.

"Burnup" is a measure of how much energy has been extracted from a given mass of heavy metal in fuel, often expressed (for power reactors) in terms of gigawatt-days per ton of heavy metal. Burnup is an important factor in determining the types and abundances of isotopes produced by a fission reactor. Breeder reactors, by design, have extremely high burnup compared to a conventional reactor, as breeder reactors produce much more of their waste in the form of fission products, while most or all of the actinides are meant to be fissioned and destroyed.

In the past, breeder-reactor development focused on reactors with low breeding ratios, from 1.01 for the Shippingport Reactor running on thorium fuel and cooled by conventional light water to over 1.2 for the Soviet BN-350 liquid-metal-cooled reactor. Theoretical models of breeders with liquid sodium coolant flowing through tubes inside fuel elements ("tube-in-shell" construction) suggest breeding ratios of at least 1.8 are possible on an industrial scale. The Soviet BR-1 test reactor achieved a breeding ratio of 2.5 under non-commercial conditions.

Types of breeder reactor

Production of heavy transuranic actinides in current thermal-neutron fission reactors through neutron capture and decays. Starting at uranium-238, isotopes of plutonium, americium, and curium are all produced. In a fast neutron-breeder reactor, all these isotopes may be burned as fuel.

Many types of breeder reactor are possible:

A "breeder" is simply a reactor designed for very high neutron economy with an associated conversion rate higher than 1.0. In principle, almost any reactor design could be tweaked to become a breeder. For example, the Light Water Reactor, a very heavily moderated thermal design, evolved into the Super Fast Reactor concept, using light water in an extremely low-density supercritical form to increase the neutron economy enough to allow breeding.

Aside from water-cooled, there are many other types of breeder reactor currently envisioned as possible. These include molten-salt cooled, gas cooled, and liquid-metal cooled designs in many variations. Almost any of these basic design types may be fueled by uranium, plutonium, many minor actinides, or thorium, and they may be designed for many different goals, such as creating more fissile fuel, long-term steady-state operation, or active burning of nuclear wastes.

Extant reactor designs are sometimes divided into two broad categories based upon their neutron spectrum, which generally separates those designed to use primarily uranium and transuranics from those designed to use thorium and avoid transuranics. These designs are:

• Fast breeder reactors (FBR's) which use 'fast' (i.e. unmoderated) neutrons to breed fissile plutonium (and possibly higher transuranics) from fertile uranium-238. The fast spectrum is flexible enough that it can also breed fissile uranium-233 from thorium, if desired.
• Thermal breeder reactors which use 'thermal-spectrum' or 'slow' (i.e. moderated) neutrons to breed fissile uranium-233 from thorium (thorium fuel cycle). Due to the behavior of the various nuclear fuels, a thermal breeder is thought commercially feasible only with thorium fuel, which avoids the buildup of the heavier transuranics.

Reprocessing

Fission of the nuclear fuel in any reactor unavoidably produces neutron-absorbing fission products. One must reprocess the fertile material from a breeder reactor to remove those neutron poisons. This step is required to fully utilize the ability to breed as much or more fuel than is consumed. All reprocessing can present a proliferation concern, since it can extract weapons-usable material from spent fuel. The most common reprocessing technique, PUREX, presents a particular concern, since it was expressly designed to separate pure plutonium. Early proposals for the breeder-reactor fuel cycle posed an even greater proliferation concern because they would use PUREX to separate plutonium in a highly attractive isotopic form for use in nuclear weapons.

Several countries are developing reprocessing methods that do not separate the plutonium from the other actinides. For instance, the non-water-based pyrometallurgical electrowinning process, when used to reprocess fuel from an integral fast reactor, leaves large amounts of radioactive actinides in the reactor fuel. More conventional water-based reprocessing systems include SANEX, UNEX, DIAMEX, COEX, and TRUEX, and proposals to combine PUREX with those and other co-processes.

All these systems have moderately better proliferation resistance than PUREX, though their adoption rate is low.

In the thorium cycle, thorium-232 breeds by converting first to protactinium-233, which then decays to uranium-233. If the protactinium remains in the reactor, small amounts of uranium-232 are also produced, which has the strong gamma emitter thallium-208 in its decay chain. Similar to uranium-fueled designs, the longer the fuel and fertile material remain in the reactor, the more of these undesirable elements build up. In the envisioned commercial thorium reactors, high levels of uranium-232 would be allowed to accumulate, leading to extremely high gamma-radiation doses from any uranium derived from thorium. These gamma rays complicate the safe handling of a weapon and the design of its electronics; this explains why uranium-233 has never been pursued for weapons beyond proof-of-concept demonstrations.

While the thorium cycle may be proliferation-resistant with regard to uranium-233 extraction from fuel (because of the presence of uranium-232), it poses a proliferation risk from an alternate route of uranium-233 extraction, which involves chemically extracting protactinium-233 and allowing it to decay to pure uranium-233 outside of the reactor. This process is an obvious chemical operation which is not required for normal operation of these reactor designs, but it could feasibly happen beyond the oversight of organizations such as the International Atomic Energy Agency (IAEA), and thus must be safeguarded against.

Waste reduction

Actinides and fission products by half-life
Actinides by decay chain				Half-life range (a)	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 a		155Euþ
244Cmƒ	241Puƒ	250Cf	227Ac№	10–29 a	90Sr	85Kr	113mCdþ
232Uƒ		238Puƒ	243Cmƒ	29–97 a	137Cs	151Smþ	121mSn
248Bk	249Cfƒ	242mAmƒ		141–351 a	No fission products have a half-life in the range of 100 a–210 ka ...
	241Amƒ		251Cfƒ	430–900 a
		226Ra№	247Bk	1.3–1.6 ka
240Pu	229Th	246Cmƒ	243Amƒ	4.7–7.4 ka
	245Cmƒ	250Cm		8.3–8.5 ka
			239Puƒ	24.1 ka
		230Th№	231Pa№	32–76 ka
236Npƒ	233Uƒ	234U№		150–250 ka	99Tc₡	126Sn
248Cm		242Pu		327–375 ka		79Se₡
				1.53 Ma	93Zr
	237Npƒ			2.1–6.5 Ma	135Cs₡	107Pd
236U			247Cmƒ	15–24 Ma		129I₡
244Pu				80 Ma	... nor beyond 15.7 Ma
232Th№		238U№	235Uƒ№	0.7–14.1 Ga
₡,  has thermal neutron capture cross section in the range of 8–50 barns ƒ,  fissile №,  primarily a naturally occurring radioactive material (NORM) þ,  neutron poison (thermal neutron capture cross section greater than 3k barns)

Nuclear waste became a greater concern by the 1990s. Breeding fuel cycles attracted renewed interest because of their potential to reduce actinide wastes, particularly various isotopes of plutonium and the minor actinides (neptunium, americium, curium, etc). Since breeder reactors on a closed fuel cycle would use nearly all of the isotopes of these actinides fed into them as fuel, their fuel requirements would be reduced by a factor of about 100. The volume of waste they generate would be reduced by a factor of about 100 as well. While there is a huge reduction in the volume of waste from a breeder reactor, the activity of the waste is about the same as that produced by a light-water reactor.

In addition, the waste from a breeder reactor has a different decay behavior, because it is made up of different materials. Breeder reactor waste is mostly fission products, while light-water reactor waste is mostly un-used uranium isotopes and a large quantity of transuranics. After spent nuclear fuel has been removed from a light-water reactor for longer than 100,000 years, the transuranics would be the main source of radioactivity. Eliminating them would eliminate much of the long-term radioactivity from the spent fuel.

In principle, breeder fuel cycles can recycle and consume all actinides, leaving only fission products. As the graphic in this section indicates, fission products have a peculiar 'gap' in their aggregate half-lives, such that no fission products have a half-life between 91 years and two hundred thousand years. As a result of this physical oddity, after several hundred years in storage, the activity of the radioactive waste from a Fast Breeder Reactor would quickly drop to the low level of the long-lived fission products. However, to obtain this benefit requires the highly efficient separation of transuranics from spent fuel. If the fuel reprocessing methods used leave a large fraction of the transuranics in the final waste stream, this advantage would be greatly reduced.

Both types of breeding cycles can reduce actinide wastes:

• The fast breeder reactor's fast neutrons can fission actinide nuclei with even numbers of both protons and neutrons. Such nuclei usually lack the low-speed "thermal neutron" resonances of fissile fuels used in LWRs.
• The thorium fuel cycle inherently produces lower levels of heavy actinides. The fertile material in the thorium fuel cycle has an atomic weight of 232, while the fertile material in the uranium fuel cycle has an atomic weight of 238. That mass difference means that thorium-232 requires six more neutron capture events per nucleus before the transuranic elements can be produced. In addition to this simple mass difference, the reactor gets two chances to fission the nuclei as the mass increases: First as the effective fuel nuclei U233, and as it absorbs two more neutrons, again as the fuel nuclei U235.

A reactor whose main purpose is to destroy actinides, rather than increasing fissile fuel-stocks, is sometimes known as a burner reactor. Both breeding and burning depend on good neutron economy, and many designs can do either. Breeding designs surround the core by a breeding blanket of fertile material. Waste burners surround the core with non-fertile wastes to be destroyed. Some designs add neutron reflectors or absorbers.

Breeder reactor concepts

There are several concepts for breeder reactors; the two main ones are:

• Reactors with a fast neutron spectrum are called fast breeder reactors (FBR's) – these typically utilize uranium-238 as fuel.
• Reactors with a thermal neutron spectrum are called thermal breeder reactors – these typically utilize thorium-232 as fuel.

Fast breeder reactor

Schematic diagram showing the difference between the Loop and Pool types of LMFBR.

In 2006 all large-scale fast breeder reactor (FBR) power stations were liquid metal fast breeder reactors (LMFBR) 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 outside the reactor tank (but inside the biological shield due to radioactive ²⁴
  Na in the primary coolant)

Experimental Breeder Reactor II, which served as the prototype for the Integral Fast Reactor

• Pool type, in which the primary heat exchangers and pumps are immersed in the reactor tank

All current fast neutron reactor designs use liquid metal as the primary coolant, to transfer heat from the core to steam used to power the electricity generating turbines. FBRs have been built cooled by liquid metals other than sodium—some early FBRs used mercury, other experimental reactors have used a sodium-potassium alloy called NaK. Both 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. Lead and lead-bismuth alloy have also been used.

Three of the proposed generation IV reactor types are FBRs:

• Gas-cooled fast reactor (GFR) cooled by helium.
• Sodium-cooled fast reactor (SFR) based on the existing liquid-metal FBR (LMFBR) and integral fast reactor designs.
• Lead-cooled fast reactor (LFR) based on Soviet naval propulsion units.

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 (used because it is "transparent" to neutrons). Enriched uranium can also be used on its own.

Many designs surround the core in a blanket of tubes that contain non-fissile uranium-238, which, by capturing fast neutrons from the reaction in the core, converts to fissile plutonium-239 (as is some of the uranium in the core), which is then reprocessed and used as nuclear fuel. Other FBR designs rely on the geometry of the fuel itself (which also contains uranium-238), arranged to attain sufficient fast neutron capture. The plutonium-239 (or the fissile uranium-235) fission cross-section is much smaller in a fast spectrum than in a thermal spectrum, as is the ratio between the ²³⁹Pu/²³⁵U fission cross-section and the ²³⁸U absorption cross-section. This increases the concentration of ²³⁹Pu/²³⁵U needed to sustain a chain reaction, as well as the ratio of breeding to fission. On the other hand, a fast reactor needs no moderator to slow down the neutrons at all, taking advantage of the fast neutrons producing a greater number of neutrons per fission than slow neutrons. For this reason ordinary liquid water, being a moderator and neutron absorber, is an undesirable primary coolant for fast reactors. Because large amounts of water in the core are required to cool the reactor, the yield of neutrons and therefore breeding of ²³⁹Pu are strongly affected. Theoretical work has been done on reduced moderation water reactors, which may have a sufficiently fast spectrum to provide a breeding ratio slightly over 1. This would likely result in an unacceptable power derating and high costs in a liquid-water-cooled reactor, but the supercritical water coolant of the supercritical water reactor (SCWR) has sufficient heat capacity to allow adequate cooling with less water, making a fast-spectrum water-cooled reactor a practical possibility.

The type of coolants, temperatures and fast neutron spectrum puts the fuel cladding material (normally austenitic stainless or ferritic-martensitic steels) under extreme conditions. The understanding of the radiation damage, coolant interactions, stresses and temperatures are necessary for the safe operation of any reactor core. All materials used to date in sodium-cooled fast reactors have known limits, as explored in ONR-RRR-088 review. Oxide Dispersion Strengthened (ODS) steel is viewed as the long-term radiation resistant fuel-cladding material that overcome the shortcomings of today's material choices.

There are only two commercially operating breeder reactors as of 2017: the BN-600 reactor, at 560 MWe, and the BN-800 reactor, at 880 MWe. Both are Russian sodium-cooled reactors.

Integral fast reactor

Main article: Integral Fast Reactor

One design of fast neutron reactor, specifically conceived to address the waste disposal and plutonium issues, was the integral fast reactor (IFR, also known as an integral fast breeder reactor, although the original reactor was designed to not breed a net surplus of fissile material).

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. The IFR pyroprocessing system uses molten cadmium cathodes and electrorefiners to reprocess metallic fuel directly on-site at the reactor. Such systems not only co-mingle all the minor actinides with both uranium and plutonium, they are compact and self-contained, so that no plutonium-containing material needs to be transported away from the site of the breeder reactor. Breeder reactors incorporating such technology would most likely be designed with breeding ratios very close to 1.00, so that after an initial loading of enriched uranium and/or plutonium fuel, the reactor would then be refueled only with small deliveries of natural uranium metal. A quantity of natural uranium metal equivalent to a block about the size of a milk crate delivered once per month would be all the fuel such a 1 gigawatt reactor would need. Such self-contained breeders are currently envisioned as the final self-contained and self-supporting ultimate goal of nuclear reactor designers. The project was canceled in 1994 by United States Secretary of Energy Hazel O'Leary.

Other fast reactors

The graphite core of the Molten Salt Reactor Experiment

Another proposed fast reactor is a fast molten salt reactor, 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).

Several prototype FBRs have been built, ranging in electrical output from a few light bulbs' equivalent (EBR-I, 1951) to over 1,000 MWe. As of 2006, the technology is not economically competitive to thermal reactor technology, but India, Japan, China, South Korea and Russia are all committing substantial research funds to further development of fast breeder reactors, anticipating that rising uranium prices will change this in the long term. Germany, in contrast, abandoned the technology due to safety concerns. The SNR-300 fast breeder reactor was finished after 19 years despite cost overruns summing up to a total of €3.6 billion, only to then be abandoned.

India is also developing FBR technology using both uranium and thorium feedstocks.

Thermal breeder reactor

The Shippingport Reactor, used as a prototype light water breeder for five years beginning in August 1977

The advanced heavy water reactor (AHWR) is one of the few proposed large-scale uses of thorium. India is developing this technology, motivated by substantial thorium reserves; almost a third of the world's thorium reserves are in India, which lacks significant uranium reserves.

The third and final core of the Shippingport Atomic Power Station 60 MWe reactor was a light water thorium breeder, which began operating in 1977. It used pellets made of thorium dioxide and uranium-233 oxide; initially, the U-233 content of the pellets was 5–6% in the seed region, 1.5–3% in the blanket region and none in the reflector region. It operated at 236 MWt, generating 60 MWe and ultimately produced over 2.1 billion kilowatt hours of electricity. After five years, the core was removed and found to contain nearly 1.4% more fissile material than when it was installed, demonstrating that breeding from thorium had occurred.

The liquid fluoride thorium reactor (LFTR) is also planned as a thorium thermal breeder. Liquid-fluoride reactors may have attractive features, such as inherent safety, no need to manufacture fuel rods and possibly simpler reprocessing of the liquid fuel. This concept was first investigated at the Oak Ridge National Laboratory Molten-Salt Reactor Experiment in the 1960s. From 2012 it became the subject of renewed interest worldwide. Japan, India, China, the UK, as well as private US, Czech and Australian companies have expressed intent to develop and commercialize the technology.

Discussion

Like many aspects of nuclear power, fast breeder reactors have been subject to much controversy over the years. In 2010 the International Panel on Fissile Materials said "After six decades and the expenditure of the equivalent of tens of billions of dollars, the promise of breeder reactors remains largely unfulfilled and efforts to commercialize them have been steadily cut back in most countries". In Germany, the United Kingdom, and the United States, breeder reactor development programs have been abandoned. The rationale for pursuing breeder reactors—sometimes explicit and sometimes implicit—was based on the following key assumptions:

• It was expected that uranium would be scarce and high-grade deposits would quickly become depleted if fission power were deployed on a large scale; the reality, however, is that since the end of the cold war, uranium has been much cheaper and more abundant than early designers expected.
• It was expected that breeder reactors would quickly become economically competitive with the light-water reactors that dominate nuclear power today, but the reality is that capital costs are at least 25% more than water-cooled reactors.
• It was thought that breeder reactors could be as safe and reliable as light-water reactors, but safety issues are cited as a concern with fast reactors that use a sodium coolant, where a leak could lead to a sodium fire.
• It was expected that the proliferation risks posed by breeders and their "closed" fuel cycle, in which plutonium would be recycled, could be managed. But since plutonium-breeding reactors produce plutonium from U238, and thorium reactors produce fissile U233 from thorium, all breeding cycles could theoretically pose proliferation risks. However U232, which is always present in U233 produced in breeder reactors, is a strong gamma-emitter via its daughter products, and would make weapon handling extremely hazardous and the weapon easy to detect.

There are some past anti-nuclear advocates that have become pro-nuclear power as a clean source of electricity since breeder reactors effectively recycle most of their waste. This solves one of the most-important negative issues of nuclear power. In the documentary Pandora's Promise, a case is made for breeder reactors because they provide a real high-kW alternative to fossil fuel energy. According to the movie, one pound of uranium provides as much energy as 5,000 barrels of oil.

FBRs have been built and operated in the United States, the United Kingdom, France, the former USSR, India and Japan. The experimental FBR SNR-300 was built in Germany but never operated and eventually shut down amid political controversy following the Chernobyl disaster. As of 2019, two FBRs are being operated for power generation in Russia. Several reactors are planned, many for research related to the Generation IV reactor initiative.

Development and notable breeder reactors

The Soviet Union (comprising Russia and other countries, dissolved in 1991) 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-600 (1981), followed by Russia's BN-800 (2016)

Future plants

The Chinese Experimental Fast Reactor is a 65 MW (thermal), 20 MW (electric), sodium-cooled, pool-type reactor with a 30-year design lifetime and a target burnup of 100 MWd/kg.

India has been trying to develop fast breeder reactors for decades but suffered repeated delays. In 2012 an FBR called the Prototype Fast Breeder Reactor was due to be completed and commissioned. The program is intended to use fertile thorium-232 to breed fissile uranium-233. India is also pursuing thorium thermal breeder reactor technology. India's focus on thorium is due to the nation's large reserves, though known worldwide reserves of thorium are four times those of uranium. India's Department of Atomic Energy (DAE) said in 2007 that it would simultaneously construct four more breeder reactors of 500 MWe each including two at Kalpakkam.

BHAVINI, an Indian nuclear power company, was established in 2003 to construct, commission and operate all stage II fast breeder reactors outlined in India's three stage nuclear power programme. To advance these plans, the Indian FBR-600 is a pool-type sodium-cooled reactor with a rating of 600 MWe.

The China Experimental Fast Reactor (CEFR) is a 25 MW(e) prototype for the planned China Prototype Fast Reactor (CFRP). It started generating power on 21 July 2011.

China also initiated a research and development project in thorium molten-salt thermal breeder-reactor technology (liquid fluoride thorium reactor), formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011. Its ultimate target was to investigate and develop a thorium-based molten salt nuclear system over about 20 years.

Kirk Sorensen, former NASA scientist and chief nuclear technologist at Teledyne Brown Engineering, has long been a promoter of thorium fuel cycle and particularly liquid fluoride thorium reactors. In 2011, Sorensen founded Flibe Energy, a company aimed to develop 20–50 MW LFTR reactor designs to power military bases.

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

A cutaway model of the BN-600 reactor, superseded by the BN-800 reactor family.

 

Construction of the BN-800 reactor

Russia has a plan for increasing its fleet of fast breeder reactors significantly. A BN-800 reactor (800 MWe) at Beloyarsk was completed in 2012, succeeding a smaller BN-600. In June 2014 the BN-800 was started in the minimum power mode. Working at 35% of nominal efficiency, the reactor contributed to the energy network on 10 December 2015. It reached its full power production in August 2016.

Plans for the construction of a larger BN-1200 reactor (1,200 MWe) was scheduled for completion in 2018, with two additional BN-1200 reactors built by the end of 2030. However, in 2015 Rosenergoatom postponed construction indefinitely to allow fuel design to be improved after more experience of operating the BN-800 reactor, and among cost concerns.

An experimental lead-cooled fast reactor, BREST-300 will be built at the Siberian Chemical Combine (SCC) in Seversk. The BREST (Russian: bystry reaktor so svintsovym teplonositelem, English: fast reactor with lead coolant) design is seen as a successor to the BN series and the 300 MWe unit at the SCC could be the forerunner to a 1,200 MWe version for wide deployment as a commercial power generation unit. The development program is as part of an Advanced Nuclear Technologies Federal Program 2010–2020 that seeks to exploit fast reactors for uranium efficiency while 'burning' radioactive substances that would otherwise be disposed of as waste. Its core would measure about 2.3 metres in diameter by 1.1 metres in height and contain 16 tonnes of fuel. The unit would be refuelled every year, with each fuel element spending five years in total within the core. Lead coolant temperature would be around 540 °C, giving a high efficiency of 43%, primary heat production of 700 MWt yielding electrical power of 300 MWe. The operational lifespan of the unit could be 60 years. The design is expected to be completed by NIKIET in 2014 for construction between 2016 and 2020.

On 16 February 2006, the United States, France and Japan signed an "arrangement" to research and develop sodium-cooled fast reactors in support of the Global Nuclear Energy Partnership. In April 2007 the Japanese government selected Mitsubishi Heavy Industries (MHI) as the "core company in FBR development in Japan". Shortly thereafter, MHI started a new company, Mitsubishi FBR Systems (MFBR) to develop and eventually sell FBR technology.

The Marcoule Nuclear Site in France, location of the Phénix (on the left).

In September 2010 the French government allocated €651.6 million to the Commissariat à l'énergie atomique to finalize the design of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a 600 MW fourth-generation reactor design to be finalized in 2020. As of 2013 the UK had shown interest in the PRISM reactor and was working in concert with France to develop ASTRID. In 2019, CEA announced this design would not be built before mid-century.

In October 2010 GE Hitachi Nuclear Energy signed a memorandum of understanding with the operators of the US Department of Energy's Savannah River Site, which should allow the construction of a demonstration plant based on the company's S-PRISM fast breeder reactor prior to the design receiving full Nuclear Regulatory Commission (NRC) licensing approval. In October 2011 The Independent reported that the UK Nuclear Decommissioning Authority (NDA) and senior advisers within the Department for Energy and Climate Change (DECC) had asked for technical and financial details of PRISM, partly as a means of reducing the country's plutonium stockpile.

The traveling wave reactor (TWR) proposed in a patent by Intellectual Ventures is a fast breeder reactor designed to not need fuel reprocessing during the decades-long lifetime of the reactor. The breed-burn wave in the TWR design does not move from one end of the reactor to the other but gradually from the inside out. Moreover, as the fuel's composition changes through nuclear transmutation, fuel rods are continually reshuffled within the core to optimize the neutron flux and fuel usage at any given point in time. Thus, instead of letting the wave propagate through the fuel, the fuel itself is moved through a largely stationary burn wave. This is contrary to many media reports, which have popularized the concept as a candle-like reactor with a burn region that moves down a stick of fuel. By replacing a static core configuration with an actively managed "standing wave" or "soliton" core, TerraPower's design avoids the problem of cooling a highly variable burn region. Under this scenario, the reconfiguration of fuel rods is accomplished remotely by robotic devices; the containment vessel remains closed during the procedure, and there is no associated downtime.