Nuclear fusion–fission hybrid

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Hybrid nuclear fusion–fission (hybrid nuclear power) is a proposed means of generating power by use of a combination of nuclear fusion and fission processes. The basic idea is to use high-energy fast neutrons from a fusion reactor to trigger fission in otherwise nonfissile fuels like U-238 or Th-232. Each neutron can trigger several fission events, multiplying the energy released by each fusion reaction hundreds of times. This would not only make fusion designs more economical in power terms, but also be able to burn fuels that were not suitable for use in conventional fission plants, even their nuclear waste. 

 

The concept dates to the 1950s, and was strongly advocated by Hans Bethe during the 1970s. At that time the first powerful fusion experiments were being built, but it would still be many years before they could be economically competitive. Hybrids were proposed as a way of greatly accelerating their market introduction, producing energy even before the fusion systems reached break-even. However, detailed studies of the economics of the systems suggested they could not compete with existing fission reactors.

The idea was abandoned and lay dormant until the 2000s, when the continued delays in reaching break-even led to a brief revival around 2009, notably as the basis of the LIFE program. This program was cancelled when the underlying technology, from the National Ignition Facility, failed to reach its design performance goals. Apollo Fusion, a company founded by Google executive Mike Cassidy in 2017, was also reported to be focused on using the subcritical nuclear fusion-fission hybrid method.

In general terms, the hybrid is similar in concept to the fast breeder reactor, which uses a compact high-energy fission core in place of the hybrid's fusion core. Another similar concept is the accelerator-driven subcritical reactor, which uses a particle accelerator to provide the neutrons instead of nuclear reactions.

Fission basics

Conventional fission power plants rely on the chain reaction caused when nuclear fission events release neutrons that cause further fission events. Each fission event in uranium releases two or three neutrons, so by careful arrangement and the use of various absorber materials, you can balance the system so one of those neutrons causes another fission event while the other one or two are lost. This careful balance is known as criticality. 

Natural uranium is a mix of several isotopes, mainly a trace amount of U-235 and over 99% U-238. When they undergo fission, both of these elements release fast neutrons with an energy distribution peaking around 1 to 2 MeV. This energy is too low to cause fission in U-238, which means it cannot sustain a chain reaction. U-235 will undergo fission when struck by neutrons of this energy, so it is possible for U-235 to sustain a chain reaction, as is the case in a nuclear bomb. However, the probability of one neutron causing fission in another U-235 atom before it escapes the fuel is too low to maintain criticality in a mass of natural uranium, so the chain reaction can only occur in fuels with increased amounts of U-235. This is accomplished by concentrating, or enriching, the fuel, increasing the amount of U-235 to produce enriched uranium, while the leftover, now mostly U-238, is a waste product known as depleted uranium.

U-235 will undergo fission more easily if the neutrons are of lower energy, the so-called thermal neutrons. Neutrons can be slowed to thermal energies through collisions with a neutron moderator material, the easiest to use being the hydrogen atoms found in water. By placing the fission fuel in water, the probability that the neutrons will cause fission in another U-235 is greatly increased, which means the level of enrichment needed to reach criticality is greatly reduced. This leads to the concept of reactor-grade enriched uranium, with the amount of U-235 increased from just less than 1% to between 3 and 5% depending on the reactor design. This is in contrast to weapons-grade enrichment, which increases to the U-235 to at least 20%, and more commonly, over 90%.

In order to maintain criticality, the fuel has to retain that extra concentration of U-235. However, a typical fission reactor burns off enough of the U-235 to cause the reaction to stop over a period on the order of a few months. A combination of burnup of the U-235 along with the creation of neutron absorbers, or poisons, as part of the fission process eventually results in the fuel mass not being able to maintain criticality. This burned up fuel has to be removed and replaced with fresh fuel. The result is nuclear waste that is highly radioactive and filled with long lived radionuclides that present a safety concern. 

The waste contains most of the U-235 it started with, only 1% or so of the energy in the fuel is extracted by the time it reaches the point where it is no longer fissile. One solution to this problem is to reprocess the fuel, which uses chemical processes to separate the U-235 (and other non-poison elements) from the waste, and then uses that U-235 in fresh fuel loads. This reduces the amount of new fuel that needs to be mined, and also concentrates the unwanted portions of the waste into a smaller load. Reprocessing is expensive, however, and has generally been more expensive than simply buying fresh fuel from the mine. 

Another possibility is to breed Pu-239 from the U-238 through neutron capture, or various other means. In order to do this, higher energy neutrons are required, which means they cannot be moderated as in a conventional reactor. The simplest way to achieve this is to further enrich the original fuel well beyond what is needed for use in a moderated reactor, to the point where the U-235 maintains criticality even with the fast neutrons. The extra fast neutrons escaping the fuel load can then be used to breed fuel in a U-238 assembly surrounding the reactor core, most commonly taken from the stocks of depleted uranium. The Pu-239 is then chemically separated and mixed into fresh fuel for conventional reactors, in the same fashion as normal reprocessing, but the total volume of fuel created in this process is much greater. In spite of this, like reprocessing, the economics of breeder reactors has proven unattractive, and commercial breeder plants have ceased operation.

Fusion basics

Fusion reactors typically burn a mixture of deuterium (D) and tritium (T). When heated to millions of degrees, the kinetic energy in the fuel begins to overcome the natural electrostatic repulsion between nuclei, the so-called coulomb barrier, and the fuel begins to undergo fusion. This reaction gives off an alpha particle and a high energy neutron of 14 MeV. A key requirement to the economic operation of a fusion reactor is that the alphas deposit their energy back into the fuel mix, heating it so that additional fusion reactions take place. This leads to a condition not unlike the chain reaction in the fission case, known as ignition. 

Deuterium can be obtained by the separation of hydrogen isotopes in sea water (see heavy water production). Tritium has a short half life of just over a decade, so only trace amounts are found in nature. To fuel the reactor, the neutrons from the reaction are used to breed more tritium through a reaction in a blanket of lithium surrounding the reaction chamber. Tritium breeding is key to the success of a D-T fusion cycle, and to date this technique has not been demonstrated. Predictions based on computer modeling suggests that the breeding ratios are quite small and a fusion plant would barely be able to cover its own use. Many years would be needed to breed enough surplus to start another reactor.

Hybrid concepts

Fusion–fission designs essentially replace the lithium blanket with a blanket of fission fuel, either natural uranium ore or even nuclear waste. The fusion neutrons have more than enough energy to cause fission in the U-238, as well as many of the other elements in the fuel, including some of the transuranic waste elements. The reaction can continue even when all of the U-235 is burned off; the rate is controlled not by the neutrons from the fission events, but the neutrons being supplied by the fusion reactor. 

Fission occurs naturally because each event gives off more than one neutron capable of producing additional fission events. Fusion, at least in D-T fuel, gives off only a single neutron, and that neutron is not capable of producing more fusion events. When that neutron strikes fissile material in the blanket, one of two reactions may occur. In many cases, the kinetic energy of the neutron will cause one or two neutrons to be struck out of the nucleus without causing fission. These neutrons still have enough energy to cause other fission events. In other cases the neutron will be captured and cause fission, which will release two or three neutrons. This means that every fusion neutron in the fusion–fission design can result in anywhere between two and four neutrons in the fission fuel.

This is a key concept in the hybrid concept, known as fission multiplication. For every fusion event, several fission events may occur, each of which gives off much more energy than the original fusion, about 11 times. This greatly increases the total power output of the reactor. This has been suggested as a way to produce practical fusion reactors in spite of the fact that no fusion reactor has yet reached break-even, by multiplying the power output using cheap fuel or waste. However, a number of studies have repeatedly demonstrated that this only becomes practical when the overall reactor is very large, 2 to 3 GWt, which makes it expensive to build.

These processes also have the side-effect of breeding Pu-239 or U-233, which can be removed and used as fuel in conventional fission reactors. This leads to an alternate design where the primary purpose of the fusion–fission reactor is to reprocess waste into new fuel. Although far less economical than chemical reprocessing, this process also burns off some of the nastier elements instead of simply physically separating them out. This also has advantages for non-proliferation, as enrichment and reprocessing technologies are also associated with nuclear weapons production. However, the cost of the nuclear fuel produced is very high, and is unlikely to be able to compete with conventional sources.

Neutron economy

A key issue for the fusion–fission concept is the number and lifetime of the neutrons in the various processes, the so-called neutron economy. 

In a pure fusion design, the neutrons are used for breeding tritium in a lithium blanket. Natural lithium consists of about 92% Li-7 and the rest is mostly Li-6. Li-7 requires neutron energies even higher than those released by fission, around 5 MeV, well within the range of energies provided by fusion. This reaction produces Tritium and Helium-4, and another slow neutron. Li-6 can react with high or low energy neutrons, including those released by the Li-7 reaction. This means that a single fusion reaction can produce several tritiums, which is a requirement if the reactor is going to make up for natural decay and losses in the fusion processes. 

When the lithium blanket is replaced, or supplanted, by fission fuel in the hybrid design, neutrons that do react with the fissile material are no longer available for tritium breeding. The new neutrons released from the fission reactions can be used for this purpose, but only in Li-6. One could process the lithium to increase the amount of Li-6 in the blanket, making up for these losses, but the downside to this process is that the Li-6 reaction only produces one tritium atom. Only the high-energy reaction between the fusion neutron and Li-7 can create more than one tritium, and this is essential for keeping the reactor running. 

To address this issue, at least some of the fission neutrons must also be used for tritium breeding in Li-6. Every one that does is no longer available for fission, reducing the reactor output. This requires a very careful balance if one wants the reactor to be able to produce enough tritium to keep itself running, while also producing enough fission events to keep the fission side energy positive. If these cannot be accomplished simultaneously, there is no reason to build a hybrid. Even if this balance can be maintained, it might only occur at a level that is economically infeasible.

Overall economy

Through the early development of the hybrid concept the question of overall economics appeared difficult to handle. A series of studies starting in the late 1970s provided a much clearer picture of the hybrid in a complete fuel cycle, and allowed the economics to be better understood. These studies appeared to indicate there was no reason to build a hybrid.

One of the most detailed of these studies was published in 1980 by Los Alamos National Laboratory (LANL). Their study noted that the hybrid would produce most of its energy indirectly, both through the fission events in its own reactor, and much more by providing Pu-239 to fuel conventional fission reactors. In this overall picture, the hybrid is essentially identical to the breeder reactor, which uses fast neutrons from plutonium fission to breed more fuel in a fission blanket in largely the same fashion as the hybrid. Both require chemical processing to remove the bred Pu-239, both presented the same proliferation and safety risks as a result, and both produced about the same amount of fuel. Since that fuel is the primary source of energy in the overall cycle, the two systems were almost identical in the end.

What was not identical, however, was the technical maturity of the two designs. The hybrid would require considerable additional research and development before it would be known if it could even work, and even if that were demonstrated, the end result would be a system essentially identical to breeders which were already being built at that time. The report concluded:

The investment of time and money required to commercialize the hybrid cycle could only be justified by a real or perceived advantage of the hybrid over the classical FBR. Our analysis leads us to conclude that no such advantage exists. Therefore, there is not sufficient incentive to demonstrate and commercialize the fusion–fission hybrid.

Rationale

The fusion process alone currently does not achieve sufficient gain (power output over power input) to be viable as a power source. By using the excess neutrons from the fusion reaction to in turn cause a high-yield fission reaction (close to 100%) in the surrounding subcritical fissionable blanket, the net yield from the hybrid fusion–fission process can provide a targeted gain of 100 to 300 times the input energy (an increase by a factor of three or four over fusion alone). Even allowing for high inefficiencies on the input side (i.e. low laser efficiency in ICF and Bremsstrahlung losses in Tokamak designs), this can still yield sufficient heat output for economical electric power generation. This can be seen as a shortcut to viable fusion power until more efficient pure fusion technologies can be developed, or as an end in itself to generate power, and also consume existing stockpiles of nuclear fissionables and waste products. 

In the LIFE project at the Lawrence Livermore National Laboratory LLNL, using technology developed at the National Ignition Facility, the goal is to use fuel pellets of deuterium and tritium surrounded by a fissionable blanket to produce energy sufficiently greater than the input (laser) energy for electrical power generation. The principle involved is to induce inertial confinement fusion (ICF) in the fuel pellet which acts as a highly concentrated point source of neutrons which in turn converts and fissions the outer fissionable blanket. In parallel with the ICF approach, the University of Texas at Austin is developing a system based on the tokamak fusion reactor, optimising for nuclear waste disposal versus power generation. The principles behind using either ICF or tokamak reactors as a neutron source are essentially the same (the primary difference being that ICF is essentially a point-source of neutrons while Tokamaks are more diffuse toroidal sources).

Use to dispose of nuclear waste

The surrounding blanket can be a fissile material (enriched uranium or plutonium) or a fertile material (capable of conversion to a fissionable material by neutron bombardment) such as thorium, depleted uranium or spent nuclear fuel. Such subcritical reactors (which also include particle accelerator-driven neutron spallation systems) offer the only currently-known means of active disposal (versus storage) of spent nuclear fuel without reprocessing. Fission by-products produced by the operation of commercial light water nuclear reactors (LWRs) are long-lived and highly radioactive, but they can be consumed using the excess neutrons in the fusion reaction along with the fissionable components in the blanket, essentially destroying them by nuclear transmutation and producing a waste product which is far safer and less of a risk for nuclear proliferation. The waste would contain significantly reduced concentrations of long-lived, weapons-usable actinides per gigawatt-year of electric energy produced compared to the waste from a LWR. In addition, there would be about 20 times less waste per unit of electricity produced. This offers the potential to efficiently use the very large stockpiles of enriched fissile materials, depleted uranium, and spent nuclear fuel.

Safety

In contrast to current commercial fission reactors, hybrid reactors potentially demonstrate what is considered inherently safe behavior because they remain deeply subcritical under all conditions and decay heat removal is possible via passive mechanisms. The fission is driven by neutrons provided by fusion ignition events, and is consequently not self-sustaining. If the fusion process is deliberately shut off or the process is disrupted by a mechanical failure, the fission damps out and stops nearly instantly. This is in contrast to the forced damping in a conventional reactor by means of control rods which absorb neutrons to reduce the neutron flux below the critical, self-sustaining, level. The inherent danger of a conventional fission reactor is any situation leading to a positive feedback, runaway, chain reaction such as occurred during the Chernobyl disaster. In a hybrid configuration the fission and fusion reactions are decoupled, i.e. while the fusion neutron output drives the fission, the fission output has no effect whatsoever on the fusion reaction, completely eliminating any chance of a positive feedback loop.

Fuel cycle

There are three main components to the hybrid fusion fuel cycle: deuterium, tritium, and fissionable elements. Deuterium can be derived by separation of hydrogen isotopes in sea water. Tritium may be generated in the hybrid process itself by absorption of neutrons in lithium bearing compounds. This would entail an additional lithium bearing blanket and a means of collection. The third component is externally derived fissionable materials from demilitarized supplies of fissionables, or commercial nuclear fuel and waste streams. Fusion driven fission also offers the possibility of using Thorium as a fuel, which would greatly increase the potential amount of fissionables available. The extremely energetic nature of the fast neutrons emitted during the fusion events (up to 0.17 the speed of light) can allow normally non-fissioning U-238 to undergo fission directly (without conversion first to Pu-239), enabling refined natural Uranium to be used with very low enrichment, while still maintaining a deeply subcritical regime.

Engineering considerations

Practical engineering designs must first take into account safety as the primary goal. All designs should incorporate passive cooling in combination with refractory materials to prevent melting and reconfiguration of fissionables into geometries capable of un-intentional criticality. Blanket layers of Lithium bearing compounds will generally be included as part of the design to generate Tritium to allow the system to be self-supporting for one of the key fuel element components. Tritium, because of its relatively short half-life and extremely high radioactivity, is best generated on site to obviate the necessity of transportation from a remote location. D-T fuel can be manufactured on site using Deuterium derived from heavy water production and Tritium generated in the hybrid reactor itself. Nuclear spallation to generate additional neutrons can be used to enhance the fission output, with the caveat that this is a tradeoff between the number of neutrons (typically 20-30 neutrons per spallation event) against a reduction of the individual energy of each neutron. This is a consideration if the reactor is to use natural Thorium as a fuel. While high energy (0.17c) neutrons produced from fusion events are capable of directly causing fission in both Thorium and U-238, the lower energy neutrons produced by spallation generally cannot. This is a tradeoff which affects the mixture of fuels against the degree of spallation used in the design.

Breeder reactor (updated)

From Wikipedia, the free encyclopedia

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 devices achieve this because their neutron economy is high enough to create more fissile fuel than they use from fertile material, such as uranium-238 or thorium-232. Breeders were at first found attractive because their fuel economy was better 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 efficiency and types of nuclear waste

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 (non-fissile)	0.350 barn	3 (non-fissile)
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 (non-fissile)	1.136 barn	11
Np-237	0.02249 barn	3 (non-fissile)	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 (non-fissile)	2.253 barn	55
Pu-241	1012 barn	75	2.298 barn	87
Pu-242	0.002557 barn	1 (non-fissile)	2.027 barn	53
Am-241	600.4 barn	1 (non-fissile)	0.2299 microbarn	21
Am-242m	6409 barn	75	2.550 barn	94
Am-243	0.1161 barn	1 (non-fissile)	2.140 barn	23
Cm-242	5.064 barn	1 (non-fissile)	2.907 barn	10
Cm-243	617.4 barn	78	2.500 barn	94
Cm-244	1.037 barn	4 (non-fissile)	0.08255 microbarn	33

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 or energy used in mining. Adherents claim that with seawater uranium extraction, there would be enough fuel for breeder reactors to satisfy our energy needs for 5 billion years at 1983's total energy consumption rate, thus making nuclear energy effectively a renewable energy.

Nuclear waste became a greater concern by the 1990s. In broad terms, spent nuclear fuel has two 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 physical behavior of the fission products is markedly different from that of the transuranics. In particular, fission products do not themselves undergo fission, and therefore cannot be used for nuclear weapons. 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 became interesting again because 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.

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. An example of this process is the evolution of the Light Water Reactor, a very heavily moderated thermal design, into the Super Fast Reactor concept, using light water in an extremely low-density supercritical form to increase the neutron economy high 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:

1. Fast breeder reactor (FBR) 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.
2. Thermal breeder reactor which use thermal-spectrum (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 produces neutron-absorbing fission products. Because of this unavoidable physical process, it is necessary to 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 extracts 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 co-processes.

All these systems have modestly 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 could happen beyond the oversight of organizations such as the International Atomic Energy Agency (IAEA).

Waste reduction

Actinides and fission products by half-life
Actinides by decay chain				Half-life range (y)	Fission products of 235U by yield
4n	4n+1	4n+2	4n+3
						4.5–7%	0.04–1.25%	<0 .001="" span="">
228Ra№				4–6	†		155Euþ
244Cmƒ	241Puƒ	250Cf	227Ac№	10–29		90Sr	85Kr	113mCdþ
232Uƒ		238Puƒ	243Cmƒ	29–97		137Cs	151Smþ	121mSn
248Bk[37]	249Cfƒ	242mAmƒ		141–351	No fission products have a half-life in the range of 100–210 k years ...
	241Amƒ		251Cfƒ[38]	430–900
		226Ra№	247Bk	1.3 k – 1.6 k
240Pu	229Th	246Cmƒ	243Amƒ	4.7 k – 7.4 k
	245Cmƒ	250Cm		8.3 k – 8.5 k
			239Puƒ	24.1 k
		230Th№	231Pa№	32 k – 76 k
236Npƒ	233Uƒ	234U№		150 k – 250 k	‡	99Tc₡	126Sn
248Cm		242Pu		327 k – 375 k			79Se₡
				1.53 M		93Zr
	237Npƒ			2.1 M – 6.5 M		135Cs₡	107Pd
236U			247Cmƒ	15 M – 24 M			129I₡
244Pu№				80 M	... nor beyond 15.7 M years
232Th№		238U№	235Uƒ№	0.7 G – 14.1 G
Legend for superscript symbols ₡  has thermal neutron capture cross section in the range of 8–50 barns ƒ  fissile m  metastable isomer №  primarily a naturally occurring radioactive material (NORM) þ  neutron poison (thermal neutron capture cross section greater than 3k barns) †  range 4–97 y: Medium-lived fission product ‡  over 200,000 y: Long-lived fission product

Nuclear waste became a greater concern by the 1990s. Breeding fuel cycles attracted renewed interest because of their potential to reduce actinide wastes, particularly plutonium and minor actinides. Since breeder reactors on a closed fuel cycle would use nearly all of the 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 has a large quantity of transuranics. After spent nuclear fuel has been removed from a light-water reactor for longer than 100,000 years, these 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 longer than 91 years and shorter than 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:

1. Reactors with a fast neutron spectrum are called fast breeder reactors (FBR) – these typically utilize uranium-238 as fuel.
2. 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 sodium-24 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. The relative merits of lead vs sodium are discussed here.

Four of the proposed generation IV reactor types are FBRs:

1. Gas-cooled fast reactor (GFR) cooled by helium.
2. Sodium-cooled fast reactor (SFR) based on the existing liquid-metal FBR (LMFBR) and integral fast reactor designs.
3. Lead-cooled fast reactor (LFR) based on Soviet naval propulsion units.
4. Supercritical water reactor (SCWR) based on existing LWR and supercritical boiler technology.

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.

Scaled-down model of TOPAZ nuclear reactor

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 an 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.

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

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 alpha-emitter, 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. At present time, two FBRs are being operated for power generation in Russia. Several reactors are planned, many for research related to the Generation IV reactor initiative.

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 an early leader in the FBR segment. 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 amongst 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 February 16, 2006, the U.S., France and Japan signed an "arrangement" to research and develop sodium-cooled fast reactors in support of the Global Nuclear Energy Partnership. 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) and possible future site of the ASTRID Gen-IV reactor.

 

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 operational 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 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.