Breeder reactor

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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 breed 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 thaBreeder reactors could, in principle, extract almost all of the energy contained in uranium or thorium, decreasing fuel n 1% of the energy in the uranium mined from the earth.^[8] 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.^[9][10]

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.^[11]

With increased concerns about nuclear waste, breeding fuel cycles became interesting again because they can reduce actinide wastes, particularly plutonium and minor actinides.^[12] 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.^[13]

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.^[14] 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.^[15]

Conversion ratio, breakeven, breeding ratio, doubling time, and burnup

One measure of a reactor's performance is the "conversion ratio" (the average number of new fissile atoms created per fission event). All proposed nuclear reactors except specially designed and operated actinide burners^[16] 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.

The ratio of new fissile material in spent fuel to fissile material consumed from the fresh fuel is known as the "conversion ratio" or "breeding ratio" of a reactor.

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.^[17] In a breeder reactor, the conversion ratio is higher than 1. "Breakeven" is achieved when the conversion ratio reaches 1.0 and the reactor produces as much fissile material as it uses.

"Doubling time" is the amount of time it would take for a breeder reactor to produce enough new fissile material to create a starting fuel load 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, and given the amount of plutonium available in spent reactor fuel, doubling time has become a less-important metric in modern breeder-reactor design.^[18][19]

"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.^[20]

In the past, breeder-reactor development focused on reactors with low breeding-ratios, from 1.01 for the Shippingport Reactor^[21][22] running on thorium fuel and cooled by conventional light water to over 1.2 for the Soviet BN-350 liquid-metal-cooled reactor.^[23] 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.^[24] The Soviet BR-1 test reactor achieved a breeding ratio of 2.5 under non-commercial conditions.^[25]

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^[26] 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:

• 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.
• 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.^[27] 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.^[28][29]

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.^[8] 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.^[30][31][32]

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.^[33]

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).^[34]

Waste reduction

Actinides and fission products by half-life
Actinides[35] by decay chain				Half-life range (y)	Fission products of 235U by yield[36]
4n	4n+1	4n+2	4n+3
						4.5–7%	0.04–1.25%	<0 .001="" td="">
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
240Puƒ№	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		242Puƒ		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[39]
232Th№		238U№	235Uƒ№	0.7 G – 14.1 G
Legend for superscript symbols ₡  has thermal neutron capture cross section in the range of 8–50 barns ƒ  fissile m  metastable isomer №  naturally occurring radioactive material (NORM) þ  neutron poison (thermal neutron capture cross section greater than 3k barns) †  range 4–97 y: Medium-lived fission product ‡  over 200,000 y: Long-lived fission product

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.^[12] 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.^[40]

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.^[13]

In principle, breeder fuel cycles can recycle and consume all actinides,^[9] 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.^[8]

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.^[41]
• 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.^[42][43]

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.^[16]

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) – 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:^[1]

• 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.^{[further explanation needed]}

Four of the proposed generation IV reactor types are FBRs:^[44]

• 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.
• 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.^[16]

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.^[26]

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).^[45][46]

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.^[47] 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.^[48] Such self-contained breeders are currently envisioned as the final self-contained and self-supporting ultimate goal of nuclear reactor designers.^[8][16] The project was canceled in 1994 by United States Secretary of Energy Hazel O'Leary.^[49][50]

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.^[51]

India is also developing FBR technology using both uranium and thorium feedstocks.^{[citation needed]}

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.^[52] 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.^[53] 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.^[54][55]

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.^[56] Japan, India, China, the UK, as well as private US, Czech and Australian companies have expressed intent to develop and commercialize the technology.^{[citation needed]}

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.^[57][58] The rationale for pursuing breeder reactors—sometimes explicit and sometimes implicit—was based on the following key assumptions:^[58][59]

• 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.^[60]
• 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.^[61] 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.^[62]

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.^[63][64]

FBRs have been built and operated in the United States, the United Kingdom, France, the former USSR, India and Japan.^[1] 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.^[65][66][67]

Development and notable breeder reactors

Notable breeder reactors

Reactor	Country when built	Started	Shutdown	Design MWe	Final MWe	Thermal Power MWt	Capacity factor	No of leaks	Neutron temperature	Coolant	Reactor class
DFR	UK	1962	1977	14	11	65	34%	7	Fast	NaK	Test
BN-350	Soviet Union	1973	1999	135	52	750	43%	15	Fast	Sodium	Prototype
Rapsodie	France	1967	1983	0	–	40	–	2	Fast	Sodium	Test
Phénix	France	1975	2010	233	130	563	40.5%	31	Fast	Sodium	Prototype
PFR	UK	1976	1994	234	234	650	26.9%	20	Fast	Sodium	Prototype
KNK II	Germany	1977	1991	18	17	58	17.1%	21	Fast	Sodium	Research/Test
SNR-300	Germany	1985 (partial operation)	1991	327	–	–	–	–	Fast	Sodium	Prototype/Commercial
BN-600	Soviet Union	1981	operating	560	560	1470	74.2%	27	Fast	Sodium	Prototype/Commercial(Gen2)
FFTF	US	1982	1993	0	–	400	–	1	Fast	Sodium	Test
Superphénix	France	1985	1998	1200	1200	3000	7.9%	7	Fast	Sodium	Prototype/Commercial(Gen2)
FBTR	India	1985	operating	13	–	40	–	6	Fast	Sodium	Test
PFBR	India	commissioning	commissioning	500	–	1250	–	–	Fast	Sodium	Prototype/Commercial(Gen3)
Jōyō	Japan	1977	operating	0	–	150	–	–	Fast	Sodium	Test
Monju	Japan	1995	2017	246	246	714	trial only	1	Fast	Sodium	Prototype
BN-800	Russia	2015	operating	789	880	2100	–	–	Fast	Sodium	Prototype/Commercial(Gen3)
MSRE	US	1965	1969	0	–	7.4	–	–	Epithermal	Molten Salt(FLiBe)	Test
Clementine	US	1946	1952	0	–	0.025	–	–	Fast	Mercury	World's First Fast Reactor
EBR-1	US	1951	1964	0.2	0.2	1.4	–	–	Fast	NaK	World's First Power Reactor
Fermi-1	US	1963	1972	66	66	200	–	–	Fast	Sodium	Prototype
EBR-2	US	1964	1994	19	19	62.5	–	–	Fast	Sodium	Experimental/Test
Shippingport	US	1977 as breeder	1982	60	60	236	–	–	Thermal	Light Water	Experimental-Core3

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.^[71]
BOR-60 (first criticality 1969) was 60 MW, with construction started in 1965.^[72]
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.^{[73][74][needs update]} 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.^{[75][needs update]}

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.^{[citation needed][needs update]}

The China Experimental Fast Reactor (CEFR) is a 25 MW(e) prototype for the planned China Prototype Fast Reactor (CFRP).^[76] It started generating power on 21 July 2011.^[77]

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.^{[78][79][needs update]}

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.^{[80][81][82][83]}

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.^[84] Working at 35% of nominal efficiency, the reactor contributed to the energy network on 10 December 2015.^[85] It reached its full power production in August 2016.^[86]

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.^[87] 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.^[88]

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.^[89]

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.^[90] 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.^[91]

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.^[92][93] 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.^[94] 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.^[95]

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.

Molten salt reactor

From Wikipedia, the free encyclopedia

Example of a molten salt reactor scheme

A molten salt reactor (MSR) is a class of generation IV nuclear fission reactor in which the primary nuclear reactor coolant, or even the fuel itself, is a molten salt mixture. MSRs can run at higher temperatures than water-cooled reactors for a higher thermodynamic efficiency, while staying at low vapour pressure.

The nuclear fuel may be solid or dissolved in the coolant. In many designs the nuclear fuel dissolved in the coolant is uranium tetrafluoride (UF₄). The fluid becomes critical in a graphite core that serves as the moderator. Some solid-fuel designs propose ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high temperature cooling. The salts are much more efficient than compressed helium (another potential coolant in Generation IV reactor designs) at removing heat from the core, reducing the need for pumping and piping and reducing the core size.

The concept was established in the 1950s. The early Aircraft Reactor Experiment was primarily motivated by the small size that the design could provide, while the Molten-Salt Reactor Experiment was a prototype for a thorium fuel cycle breeder nuclear power plant. The increased research into Generation IV reactor designs included a renewed interest in the technology.^[2]


History

Aircraft reactor experiment


Aircraft Reactor Experiment building at ORNL. It was later retrofitted for the MSRE.

Extensive research into molten salt reactors started with the U.S. aircraft reactor experiment (ARE) in support of the U.S. Aircraft Nuclear Propulsion program. The ARE was a 2.5 MW_th nuclear reactor experiment designed to attain a high energy density for use as an engine in a nuclear-powered bomber.

The project included experiments, including high temperature reactor and engine tests collectively called the Heat Transfer Reactor Experiments: HTRE-1, HTRE-2 and HTRE-3 at the National Reactor Test Station (now Idaho National Laboratory) as well as an experimental high-temperature molten salt reactor at Oak Ridge National Laboratory – the ARE.

The ARE used molten fluoride salt NaF-ZrF₄-UF₄ (53-41-6 mol%) as fuel, moderated by beryllium oxide (BeO). Liquid sodium was a secondary coolant.

The experiment had a peak temperature of 860 °C. It produced 100 MWh over nine days in 1954. This experiment used Inconel 600 alloy for the metal structure and piping.^[3]

After ARE, another reactor was operated at the Critical Experiments Facility of the Oak Ridge National Laboratory in 1957. It was part of the circulating-fuel reactor program of the Pratt & Whitney Aircraft Company (PWAC). This was called the PWAR-1, the Pratt and Whitney Aircraft Reactor-1. The experiment was run for only a few weeks and at essentially zero nuclear power, but it reached criticality. The operating temperature was held constant at approximately 675 °C (1,250 °F). The PWAR-1 used NaF-ZrF₄-UF₄ as the primary fuel and coolant, making it one of the three critical molten salt reactors ever built.^[4]


Molten-salt reactor experiment


MSRE plant diagram

Oak Ridge National Laboratory (ORNL) took the lead in researching the MSR through the 1960s. Much of their work culminated with the Molten-Salt Reactor Experiment (MSRE). The MSRE was a 7.4 MW_th test reactor simulating the neutronic "kernel" of a type of epithermal thorium molten salt breeder reactor called the liquid fluoride thorium reactor. The large (expensive) breeding blanket of thorium salt was omitted in favor of neutron measurements.  The MSRE was located at ORNL. Its piping, core vat and structural components were made from Hastelloy-N, moderated by pyrolytic graphite. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF₂-ZrF₄-UF₄ (65-29-5-1). The graphite core moderated it. Its secondary coolant was FLiBe (2LiF-BeF₂). It reached temperatures as high as 650 °C and operated for the equivalent of about 1.5 years of full power operation.


Oak Ridge National Laboratory molten salt breeder reactor

The culmination of the Oak Ridge National Laboratory research during the 1970–1976 timeframe resulted in a proposed molten salt breeder reactor (MSBR) design which would use LiF-BeF₂-ThF₄-UF₄ (72-16-12-0.4) as fuel. It was to be moderated by graphite with a 4-year replacement schedule. The secondary coolant was to be NaF-NaBF_4. Its peak operating temperature was to be 705 °C.^[5] Despite the success, the MSR program closed down in the early 1970s in favor of the liquid metal fast-breeder reactor (LMFBR),^[6] after which research stagnated in the United States.^[7][8] As of 2011, the ARE and the MSRE remained the only molten-salt reactors ever operated.

The MSBR project received funding until 1976. Inflation-adjusted to 1991 dollars, the project received $38.9 million from 1968 to 1976.^[9]

Officially, the program was cancelled because:


• The political and technical support for the program in the United States was too thin geographically. Within the United States, only in Oak Ridge, Tennessee, was the technology well understood.^[6]
• The MSR program was in competition with the fast breeder program at the time, which got an early start and had copious government development funds allocated to many parts of the United States. When the MSR development program had progressed far enough to justify an expanded program leading to commercial development, the AEC could not justify the diversion of substantial funds from the LMFBR to a competing program.^[6]

Oak Ridge National Laboratory denatured molten salt reactor (DMSR)

In 1980, the engineering technology division at Oak Ridge National Laboratory published a paper entitled "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor with Once-Through Fueling." In it, the authors "examine the conceptual feasibility of a molten-salt power reactor fueled with denatured uranium-235 (i.e. with low-enriched uranium) and operated with a minimum of chemical processing." The main priority behind the design characteristics is proliferation resistance.^[10] Lessons learned from past projects and research at ORNL were considered. Although the DMSR can theoretically be fueled partially by thorium or plutonium, fueling solely with low enriched uranium (LEU) helps maximize proliferation resistance.

Another important goal of the DMSR was to minimize R&D and to maximize feasibility. The Generation IV international Forum (GIF) includes "salt processing" as a technology gap for molten salt reactors.^[11] The DMSR requires minimal chemical processing because it is a burner rather than a breeder. Both reactors built at ORNL were burner designs. In addition, the choices to use graphite for neutron moderation and enhanced Hastelloy-N for piping simplify the design and reduce R&D.


United Kingdom

The UK's Atomic Energy Research Establishment (AERE) were developing an alternative MSR design across its National Laboratories at Harwell, Culham, Risley and Winfrith. AERE opted to focus on a lead-cooled 2.5 GWe Molten Salt Fast Reactor (MSFR) concept using a chloride.^[12] They also researched the option of helium gas as an alternative coolant.

The UK MSFR would be fuelled by plutonium, a fuel considered to be 'free' by the program's research scientists, because of the UK's plutonium stockpile.

Despite their different designs, ORNL and AERE maintained contact during this period with information exchange and expert visits. Theoretical work on the concept was conducted between 1964 and 1966, while experimental work was ongoing between 1968 and 1973. The program received annual government funding of around £100,000-£200,000 (equivalent to £2m-£3m in 2005). This funding came to an end in 1974, partly due to the success of the Prototype Fast Reactor at Dounreay which was considered a priority for funding as it went critical in the same year.^[12]

AERE reports and findings from its MSR Program conducted in the 1960s and 1970s are available for public viewing at the UK National Archives in Kew, London.^[12]


Soviet Union

In the USSR, a molten-salt reactor research program was started in the second half of the 1970s at the Kurchatov Institute. It included theoretical and experimental studies, particularly the investigation of mechanical, corrosion and radiation properties of the molten salt container materials. The main findings supported the conclusion that there were no physical nor technological obstacles to the practical implementation of MSRs.^[15] A reduction in activity occurred after 1986 due to the Chernobyl accident, along with a general stagnation of nuclear power and the nuclear industry.^[16](p381)


Twenty-first century

Canada

Terrestrial Energy Inc. (TEI), a Canadian-based company, is developing a DMSR design called the Integral Molten Salt Reactor (IMSR). The IMSR is designed to be deployable as a small modular reactor (SMR). Their design currently undergoing licensing is 400MW thermal (190MW electrical). With high operating temperatures, the IMSR has applications in industrial heat markets as well as traditional power markets. The main design features include neutron moderation from graphite, fueling with low-enriched uranium and a compact and replaceable Core-unit. Decay heat is removed passively using nitrogen (with air as an emergency alternative). The latter feature permits the operational simplicity necessary for industrial deployment.^[17]

Terrestrial has completed the first phase of a prelicensing review by the Canadian Nuclear Safety Commission, which provides a regulatory opinion that the design features are generally safe enough to eventually obtain a license to construct the reactor.^[18]


China

Under Jiang Mianheng's direction, China initiated a thorium molten-salt reactor research project. It was formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011.^[19] A 100-MW demonstrator of the solid fuel version (TMSR-SF), based on pebble bed technology, was to be ready by 2024. Initially, a 10-MW pilot and a larger demonstrator of the liquid fuel (TMSR-LF) variant were targeted for 2024 and 2035 respectively.^[20][21] However, China has accelerated its program to build two 12 MW reactors underground at Wuwei research facilities in Gansu Province by 2020.^[22] Heat from the reaction will be used to produce electricity, hydrogen, industrial chemicals, drinking water through desalination, and minerals.^[22] The project also seeks to test new corrosion resistant materials.^[22]

In 2017, ANSTO/Shanghai Institute Of Applied Physics announced the creation of a NiMo-SiC alloy for use in molten salt reactors.^[23][24]


Denmark

Seaborg Technologies, a company based in Denmark, is developing the core for a Molten Salt Waste-burner (MSW). The MSW is a high temperature, single salt, thermal MSR designed to go critical on a combination of thorium and nuclear waste from conventional nuclear reactors. The MSW design is modular. The reactor core is estimated to be replaced every 6–10 years. However, the fuel will not be replaced and will burn for the entire power plant lifetime. The first version of the Seaborg core is planned to produce 50 MW_th power and could consume approximately 1 ton (not considering natural decays) of transuranic waste over its 60 years power plant lifetime. After 60 years the ²³³U concentration in the fuel salt is high enough to initiate a closed thorium fuel cycle in the next generation power plant.^[25]


France

The CNRS project EVOL (Evaluation and viability of liquid fuel fast reactor system) project, with the objective of proposing a design of the MSFR (Molten Salt Fast Reactor),^[26] released its final report in 2014.^[27] The various molten salt reactor projects like FHR, MOSART, MSFR, and TMSR have common themes in basic R&D areas, according to a 2014 paper giving an overview of the MSR in a GenV context.^[28] Another paper gives an overview of the MSFR.^[29] More resources are available in the MSFR bibliography.^[30]

The EVOL project will be continued by the EU-funded SAMOFAR (Safety Assessment of the Molten Salt Fast Reactor) project, in which several European research institutes and universities collaborate^[31].


India

Ratan Kumar Sinha, Chairman of Atomic Energy Commission of India, stated in 2013: "India is also investigating Molten Salt Reactor (MSR) technology. We have molten salt loops operational at BARC."^[32]


Japan

The Fuji Molten Salt Reactor is a 100 to 200 MW_e LFTR, using technology similar to the Oak Ridge project. A consortium including members from Japan, the U.S. and Russia are developing the project. The project would likely take 20 years to develop a full size reactor,^[33] but the project seems to lack funding.^[34]


United Kingdom

The Alvin Weinberg Foundation is a British non-profit organization founded in 2011, dedicated to raising awareness about the potential of thorium energy and LFTR. It was formally launched at the House of Lords on 8 September 2011.^[35][36][37] It is named after American nuclear physicist Alvin M. Weinberg, who pioneered thorium molten salt reactor research.

A study on MSRs completed in July 2015 by Energy Process Developments, funded by Innovate UK, summarizes MSR activity internationally. It looks at the feasibility of developing a pilot scale demonstration MSR in the UK. A review of potential UK sites is given along with an insight into the UK regulatory process for innovative reactor technology. The technical review of six MSR designs led to the selection of the Stable Salt Reactor, designed by Moltex Energy, as most suitable for UK implementation.^[38] However, Moltex "failed to get engagement from the UK government quickly enough" and so will target Canada instead.^[39]


United States

Idaho National Laboratory designed a molten-salt-cooled, molten-salt-fuelled reactor with a prospective output of 1000 MW_e.^[40]

Kirk Sorensen, former NASA scientist and chief nuclear technologist at Teledyne Brown Engineering, is a long-time promoter of the thorium fuel cycle, coining the term liquid fluoride thorium reactor. In 2011, Sorensen founded Flibe Energy, a company aimed at developing 20–50 MW LFTR reactor designs to power military bases. (It is easier to approve novel military designs than civilian power station designs in today's US nuclear regulatory environment).

Transatomic Power was created by Ph.D. students from MIT including CEO Leslie Dewan and Mark Massie, and Russ Wilcox of E Ink.^[45] They are pursuing what they term a Waste-Annihilating Molten Salt Reactor (acronym WAMSR), intending to consume existing spent nuclear fuel.^[46] Transatomic received venture capital funding in early 2015.^[47]

In January 2016, the United States Department of Energy announced a $80m award fund to develop Generation IV reactor designs.^[48] One of the two beneficiaries, Southern Company will use the funding to develop a Molten Chloride Fast Reactor (MCFR), a type of MSR developed earlier by British scientists.^[12]


Variants

Liquid-salt very-high-temperature reactor

It is essentially a standard VHTR design that uses liquid salt as a coolant in the primary loop, rather than a single helium loop. It relies on "TRISO" fuel dispersed in graphite. Early AHTR research focused on graphite would be in the form of graphite rods that would be inserted in hexagonal moderating graphite blocks, but current studies focus primarily on pebble-type fuel. The LS-VHTR has many attractive features, including the ability to work at very high temperatures (the boiling point of most molten salt candidates is >1400 °C); low-pressure cooling that can be used to more easily match hydrogen production facility conditions (most thermochemical cycles require temperatures in excess of 750 °C); better electric conversion efficiency than a helium-cooled VHTR operating at similar conditions; passive safety systems and better retention of fission products in the event of an accident.^{[citation needed]} This concept is now referred to as "fluoride salt-cooled high-temperature reactor" (FHR).^[49]


Liquid fluoride thorium reactor (LFTR)

Advocates estimate that five hundred metric tons of thorium could supply all U.S. energy needs for one year.^[50] The U.S. Geological Survey estimates that the largest known U.S. thorium deposit, the Lemhi Pass district on the Montana-Idaho border, contains thorium reserves of 64,000 metric tons.^[51]


Molten-salt fueling options

The LFTR design was strongly supported by Alvin Weinberg, who patented the light-water reactor and was a director of the U.S.'s Oak Ridge National Laboratory. In 2016 Nobel prize winning physicist Carlo Rubbia, former Director General of CERN, claimed that one of the main reasons why research was cut is that thorium is difficult to turn into a nuclear weapon.^[52]


Thorium is not for tomorrow but unless you do any development, it will not get there. — Dr Carlo Rubbia, Nobel Laureate and former Director General of CERN, January 2016^[52]

Alternatives to thorium include enriched uranium-235 or fissile material from dismantled nuclear weapons.^[53]


Molten-salt-cooled reactors

Molten-salt-cooled solid-fuel reactors are quite different from molten-salt-fueled reactors. They are called "molten salt reactor system" in the Generation IV proposal, also called Molten Salt Converter Reactor (MSCR). These reactors were additionally referred to as advanced high-temperature reactors (AHTRs), but since about 2010 the preferred DOE designation is fluoride high-temperature reactors (FHR).^[54]

The FHR concept cannot reprocess fuel easily and has fuel rods that need to be fabricated and validated, delaying deployment by up to twenty years^{[citation needed]} from project inception. However, since it uses fabricated fuel, reactor manufacturers can still profit by selling fuel assemblies.

The FHR retains the safety and cost advantages of a low-pressure, high-temperature coolant, also shared by liquid metal cooled reactors. Notably, steam is not created in the core (as is present in BWRs), and no large, expensive steel pressure vessel (as required for PWRs). Since it can operate at high temperatures, the conversion of the heat to electricity can use an efficient, lightweight Brayton cycle gas turbine.

Much of the current research on FHRs is focused on small, compact heat exchangers that reduce molten salt volumes and associated costs.^[55]

Molten salts can be highly corrosive and corrosivity increases with temperature. For the primary cooling loop, a material is needed that can withstand corrosion at high temperatures and intense radiation. Experiments show that Hastelloy-N and similar alloys are suited to these tasks at operating temperatures up to about 700 °C. However, operating experience is limited. Still higher operating temperatures are desirable – at 850 °C thermochemical production of hydrogen becomes possible. Materials for this temperature range have not been validated, though carbon composites, molybdenum alloys (e.g. TZM), carbides, and refractory metal based or ODS alloys might be feasible.


Dual-fluid molten salt reactors

A prototypical example of a dual fluid reactor is the lead-cooled, salt-fueled reactor.


Fused salt selection


Molten FLiBe

The salt mixtures are chosen to make the reactor safer and more practical. Fluoride salts are favored, because fluorine has only one stable isotope (F-19), and does not easily become radioactive under neutron bombardment. Both of these make fluorine better than chlorine, which has two stable isotopes (Cl-35 and Cl-37), as well as a slow-decaying isotope between them which facilitates neutron absorption by Cl-35. Compared to chlorine and other halides, fluorine also absorbs fewer neutrons and slows ("moderates") neutrons better. Low-valence fluorides boil at high temperatures, though many pentafluorides and hexafluorides boil at low temperatures. They also must be very hot before they break down into their constituent elements. Such molten salts are "chemically stable" when maintained well below their boiling points.

On the other hand, some salts are so useful that isotope separation of the halide is worthwhile. Chlorides permit fast breeder reactors to be constructed using molten salts. Much less research has been done on reactor designs using chloride salts. Chlorine, unlike fluorine, must be purified to isolate the heavier stable isotope, chlorine-37, thus reducing production of sulfur tetrafluoride that occurs when chlorine-35 absorbs a neutron to become chlorine-36, then degrades by beta decay to sulfur-36.

Similarly, any lithium present in a salt mixture must be in the form of purified lithium-7, because lithium-6 effectively captures neutrons and produces tritium. Even if pure ⁷Li is used, salts containing lithium will cause significant tritium production, comparable with heavy water reactors.

Reactor salts are usually close to eutectic mixtures to reduce their melting point. A low melting point simplifies melting the salt at startup and reduces the risk of the salt freezing as it is cooled in the heat exchanger.

Due to the high "redox window" of fused fluoride salts, the redox potential of the fused salt system can be changed. Fluorine-Lithium-Beryllium ("FLiBe") can be used with beryllium additions to lower the redox potential and almost eliminate corrosion. However, since beryllium is extremely toxic, special precautions must be engineered into the design to prevent its release into the environment. Many other salts can cause plumbing corrosion, especially if the reactor is hot enough to make highly reactive hydrogen.

To date, most research has focused on FLiBe, because lithium and beryllium are reasonably effective moderators and form a eutectic salt mixture with a lower melting point than each of the constituent salts. Beryllium also performs neutron doubling, improving the neutron economy. This process occurs when the beryllium nucleus re-emits two neutrons after absorbing a single neutron. For the fuel carrying salts, generally 1% or 2% (by mole) of UF₄ is added. Thorium and plutonium fluorides have also been used.



Comparison of the neutron capture and moderating efficiency of several materials. Red are Be-bearing, blue are ZrF₄-bearing and green are LiF-bearing salts.^[56]

Material	Total neutron capture relative to graphite (per unit volume)	Moderating ratio (Avg. 0.1 to 10 eV)
Heavy water	0.2	11449
ZrH[57][58][59]	~0.2	~0 if <0 .14="" ev="" if="">0.14 eV
Light water	75	246
Graphite	1	863
Sodium	47	2
UCO	285	2
UO2	3583	0.1
2LiF–BeF2	8	60
LiF–BeF2–ZrF4 (64.5–30.5–5)	8	54
NaF–BeF2 (57–43)	28	15
LiF–NaF–BeF2 (31–31–38)	20	22
LiF–ZrF4 (51–49)	9	29
NaF–ZrF4 (59.5–40.5)	24	10
LiF-NaF–ZrF4 (26–37–37)	20	13
KF–ZrF4 (58–42)	67	3
RbF–ZrF4 (58–42)	14	13
LiF–KF (50–50)	97	2
LiF–RbF (44–56)	19	9
LiF–NaF–KF (46.5–11.5–42)	90	2
LiF–NaF–RbF (42–6–52)	20	8

Fused salt purification

Techniques for preparing and handling molten salt were first developed at Oak Ridge National Lab.^[60] The purpose of salt purification was to eliminate oxides, sulfur and metal impurities. Oxides could result in the deposition of solid particles in reactor operation. Sulfur had to be removed because of its corrosive attack on nickel-based alloys at operational temperature. Structural metal such as chromium, nickel, and iron had to be removed for corrosion control.

A water content reduction purification stage using HF and helium sweep gas was specified to run at 400 °C. Oxide and sulfur contamination in the salt mixtures were removed using gas sparging of HF – H₂ mixture, with the salt heated to 600 °C.^[60](p8) Structural metal contamination in the salt mixtures were removed using hydrogen gas sparging, at 700 °C.^[60](p26) Solid ammonium hydrofluoride was proposed as a safer alternative for oxide removal.^[61]


Fused salt processing

The possibility of online processing can be an MSR advantage. Continuous processing would reduce the inventory of fission products, control corrosion and improve neutron economy by removing fission products with high neutron absorption cross-section, especially xenon. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. Online fuel processing can introduce risks of fuel processing accidents,^[62](p15) which can trigger release of radio isotopes.

In some thorium breeding scenarios, the intermediate product protactinium-233 would be removed from the reactor and allowed to decay into highly pure uranium-233, an attractive bomb-making material. More modern designs propose to use a lower specific power or a separate large thorium breeding blanket. This dilutes the protactinium to such an extent that few protactinium atoms absorb a second neutron or, via a (n, 2n) reaction (in which an incident neutron is not absorbed but instead knocks a neutron out of the nucleus), generate uranium-232. Because U-232 has a short half-life and its decay chain contains hard gamma emitters, it makes the isotopic mix of uranium less attractive for bomb-making. This benefit would come with the added expense of a larger fissile inventory or a 2-fluid design with a large quantity of blanket salt.

The necessary fuel salt reprocessing technology has been demonstrated, but only at laboratory scale. A prerequisite to full-scale commercial reactor design is the R&D to engineer an economically competitive fuel salt cleaning system.


Fissile fuel reprocessing issues


Changes in the composition of a MSR fast neutron (kg/GW)

Reprocessing refers to the chemical separation of fissionable uranium and plutonium from spent nuclear fuel.^[63] The recovery of uranium or plutonium could increase the risk of nuclear proliferation. In the United States the regulatory regime has varied dramatically in different administrations.^[63]

In the original 1971 Molten Salt Breeder Reactor proposal, uranium reprocessing was scheduled every ten days as part of reactor operation.^[64](p181) Subsequently, a once-through fueling design was proposed that limited uranium reprocessing to every thirty years at the end of useful salt life.^[65](p98) A mixture of uranium-238 was called for to make sure recovered uranium would not be weapons grade. This design is referred to as denatured molten salt reactor.^[66] If reprocessing were to be prohibited then the uranium would be disposed with other fission products.


Comparison to light water reactors

MSRs, especially those with the fuel dissolved in the salt differ considerably from conventional reactors. Reactor core pressure can be low and the temperature much higher. In this respect an MSR is more similar to a liquid metal cooled reactor than to a conventional light water cooled reactor. MSRs are often planned as breeding reactors with a closed fuel cycle – as opposed to the once-through fuel currently used in U.S. nuclear reactors.

Safety concepts rely on a negative temperature coefficient of reactivity and a large possible temperature rise to limit reactivity excursions. As an additional method for shutdown, a separate, passively cooled container below the reactor can be included. In case of problems and for regular maintenance the fuel is drained from the reactor. This stops the nuclear reaction and acts as another second cooling system. Neutron-producing accelerators have been proposed for some super-safe subcritical experimental designs.^[67]

Cost estimates from the 1970s were slightly lower than for conventional light-water reactors.^[68]

The temperatures of some proposed designs are high enough to produce process heat for hydrogen production or other chemical reactions. Because of this, they are included in the GEN-IV roadmap for further study.^[69]


Advantages

MSR offers many potential advantages over current light water reactors:^[5]
 

• Inherently safe design (safety by passive components and the strong negative temperature coefficient of reactivity of some designs). In some designs, the fuel and the coolant are the same fluid, so a loss of coolant removes the reactor's fuel. Unlike steam, fluoride salts dissolve poorly in water, and do not form burnable hydrogen. Unlike steel and solid uranium oxide, molten salts are not damaged by the core's neutron bombardment.
• A low-pressure MSR lacks a LWR's high-pressure radioactive steam and therefore do not experience leaks of radioactive steam and cooling water, and the expensive containment, steel core vessel, piping and safety equipment needed to contain radioactive steam.
• MSRs make closed nuclear fuel cycles cheaper and more practical. If fully implemented, a closed nuclear fuel cycle reduces environmental impacts: The chemical separation makes long-lived actinides back into reactor fuel. The discharged wastes are mostly fission products (nuclear ashes) with short half-lives. This reduces the needed geologic containment to 300 years rather than the tens of thousands of years needed by a light-water reactor's spent nuclear fuel. It also permits society to use more-abundant nuclear fuels.
• The fuel's liquid phase might be pyroprocessed to separate fission products (nuclear ashes) from actinide fuels. This may have advantages over conventional reprocessing, though much development is still needed.
• Fuel rods are not required.
• In new solid-fueled reactor designs, the longest-lead item is the safety testing of fuel element designs. Fuel tests usually must cover several three-year refueling cycles. However, several molten salt fuels have already been validated.
• Some designs can "burn" problematic transuranic elements from traditional solid-fuel nuclear reactors.
• An MSR can react to load changes in less than 60 seconds (unlike "traditional" solid-fuel nuclear power plants that suffer from xenon poisoning).
• Molten salt reactors can run at high temperatures, yielding high production efficiency. This reduces the size, expense and environmental impacts of a power plant.
• MSRs can offer a high "specific power," that is high power at a low mass as demonstrated by the ARE.^[3] Simplified MSR power plants may be suitable for ships.
• A possibly good neutron economy makes the MSR attractive for the neutron poor thorium fuel cycle.
• LWR's (and most other solid-fuel reactors) have no fundamental "off switch", but once the initial criticality is overcome, an MSR is comparatively easy and fast to turn off by letting the freeze plug melt.

Disadvantages

• Little development compared to most Gen IV designs .
• Required onsite chemical plant to manage core mixture and remove fission products.
• Required regulatory changes to deal with radically different design features.
• MSR designs rely on nickel-based alloys to hold the molten salt. Alloys based on nickel and iron are prone to embrittlement under high neutron flux.^[65](p83)
• Corrosion risk.^[70]
• As a breeder reactor, a modified MSR might be able to produce weapons-grade nuclear material.^[71]
• The MSRE and aircraft nuclear reactors used enrichment levels so high that they approach the levels of nuclear weapons. These levels would be illegal in most modern regulatory regimes for power plants. Some modern designs avoid this issue.^[72]
• Neutron damage to solid moderator materials can limit the core lifetime of an MSR that makes moderately fast neutrons. For example, the MSRE was designed so that its graphite moderator sticks had very low tolerances, so neutron damage could change their size without damage. "Two fluid" MSR designs are unable to use graphite piping because graphite changes size when it is bombarded with neutrons, and graphite pipes would crack and leak.