Molten salt reactor

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A molten salt reactor (MSR) is a type of nuclear reactor where the primary coolant is a molten salt.

There have been many designs put forward for use of this type of reactor as a nuclear power plant and a few prototypes built. The concept is one of those proposed for development as a generation IV reactor.

In many designs the nuclear fuel is dissolved in the molten fluoride salt coolant as uranium tetrafluoride (UF₄). The fluid becomes critical in a graphite core which serves as the moderator. Many modern designs rely on ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high temperature cooling.

Contents 1 History 1.1 The aircraft reactor experiment 1.2 The Molten-Salt Reactor Experiment 1.3 Oak Ridge National Laboratory reactor 1.4 Liquid salt very high temperature reactor 1.5 The Fuji MSR 2 Technological issues 2.1 Molten-salt Fueled Reactors 2.2 Molten-salt cooled reactors 2.3 Fused salt selection 2.4 Fused salt purification and reprocessing 3 Political issues 4 Comparison to ordinary light water reactors 5 References 6 See also 7 External links

[edit] History

[edit] The aircraft reactor experiment

Main article: Aircraft Reactor Experiment

Extensive research into molten salt reactors started with the US Aircraft Reactor Experiment (ARE). The US Aircraft Reactor Experiment was a 2.5 MW_th nuclear reactor experiment designed to attain a high power density for use as an engine in a nuclear powered bomber. The project resulted in several experiments, three of which resulted in engine tests collectively called the Heat Transfer Reactor Experiments: HTRE-l, HTRE-2, and HTRE-3. One experiment used the molten fluoride salt NaF-ZrF₄-UF₄ (53-41-6 mol%) as fuel, was moderated by beryllium oxide (BeO), used liquid sodium as a secondary coolant, and had a peak temperature of 860 °C. It operated for a 1000 hour cycle in 1954. This experiment used Inconel 600 alloy for the metal structure and piping.

[edit] The Molten-Salt Reactor Experiment

Main article: Molten-Salt Reactor Experiment

Oak Ridge National Laboratory took the lead in researching the MSR through 1960s, and 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 an inherently safe epithermal thorium breeder reactor. It tested molten salt fuels of uranium and plutonium. The tested ²³³UF₄ fluid fuel has a unique decay path that minimizes waste, with waste isotopes having half-lives under 50 years. The red-hot 650 °C temperature of the reactor could power high-efficiency heat engines such as gas turbines. 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 and its moderator was pyrolytic graphite. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF₂-ZrF₄-UF₄ (65-30-5-0.1), the graphite core moderated it, and 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. (For more information, see the main article.)

[edit] Oak Ridge National Laboratory reactor

The culmination of the Oak Ridge National Laboratory research during the 1970-76 timeframe resulted in an MSR design which would use LiF-BeF₂-ThF₄-UF₄ (72-16-12-0.4) as fuel, was to be moderated by graphite with a 4 year replacement schedule, use NaF-NaBF₄ as the secondary coolant, and have a peak operating temperature of 705 °C. However, to date the molten salt reactor remains a "paper design", that is, no molten salt reactors have been built other than the experimental MSRE.

[edit] Liquid salt very high temperature reactor

Main article: Very high temperature reactor

Research is currently picking up again for reactors that utilize molten salts for coolant. Both the traditional molten salt reactor and the Very High Temperature Reactor (VHTR) have been picked as potential designs to be studied under the Generation Four Initiative (GEN-IV). A version of the VHTR currently being studied is the Liquid Salt Very High Temperature Reactor (LS-VHTR). 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. The fuel graphite would be in the form of graphite rods that would be inserted in hexagonal moderating graphite blocks. The molten salt would pass through holes drilled in the graphite blocks. The LS-VHTR has many attractive features, including: the ability to work at very high temperatures (the boiling point of most molten salts being considered are >1400 °C), low pressure cooling that can be used to more easily match hydrogen production facility conditions (most thermo chemical 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.

[edit] The Fuji MSR

The FUJI mini-MSR is a 100 MWe molten-salt-fueled Thorium fuel cycle thermal breeder reactor, using technology similar to the Oak Ridge National Laboratory Reactor. It is being developed by a consortium including members from Japan, the U.S. and Russia. As a breeder reactor, it converts Thorium into nuclear fuels. As a thermal-spectrum reactor, its neutron regulation is inherently safe. Like all molten salt reactors, its core is chemically inert, under low pressures to prevent explosions and toxic releases.^[1]

[edit] Technological issues

[edit] Molten-salt Fueled Reactors

The classic MSFR has been very exciting to many nuclear engineers. Its most prominent champion was Alvin Weinberg, who patented the light-water reactor and was a director of the U.S.'s Oak Ridge National Laboratory, a prominent nuclear research center.

Two concepts were investigated. The "two fluid" reactor had a high-neutron-density core that burned U233 from the thorium fuel cycle. A blanket of thorium salts absorbed the neutrons and was eventually transmuted to U233 fuel. The engineers discovered that by carefully sculpting the moderator rods (to get neutron densities similar to a core and blanket), and modifying the fuel reprocessing chemistry, both thorium and uranium salts could coexist in a simpler, less expensive yet efficient "one fluid" reactor.

The power reactor design produced by Weinberg's research group was similar to the MSRE above, which was designed to validate the risky hot, high-neutron-density "kernel" part of the "kernel and blanket" thorium breeder.

The advantages cited by Weinberg and his associates at Oak Ridge National Laboratory include:

• It's safe to operate and maintain: Molten fluoride salts are mechanically and chemically stable at sea-level pressures at intense heats and radioactivity. Fluoride combines ionically with almost any transmutation product, keeping it out of circulation. Even radioactive noble gases come out in a predictable, containable place, where the fuel is coolest and most dispersed, the pump bowl.
• There's no high pressure steam in the core, just low-pressure molten salt. This means that the MSR's core cannot have a steam explosion, and does not need the most expensive item in a light water reactor, a high-pressure steam vessel for the core. Instead, there is a vat and low-pressure pipes (for molten salt) constructed of thick sheet metal. The metal is an exotic nickel alloy that resists heat and corrosion, Hastelloy-N, but there is much less of it, and the thin metal is less expensive to form and weld.
• With fuel reprocessing, the Thorium fuel cycle, so impractical in other types of reactors, produces 0.1% of the long-term high-level radioactive waste of a light-water reactor without reprocessing (all modern reactors in the U.S.). As thorium captures neutrons, it first becomes Th233, which quickly decays to protactinium (Pa233). Pa233 in turn decays to U233 with a half-life of 27 days. U233 is an excellent reactor fuel. As U233 is bombarded by neutrons with a thermal spectrum of speeds, each absorbed neutron either splits the uranium or produces a heavier isotope of uranium, which fission to elements similar to those from U233. These fission products almost all have half-lives less than 30 years. The only source of high-radioactivity long-lived transuranic elements is that a tiny bit of neptunium is produced from the tiny fraction of U236 produced at the tail-end of this process. The neptunium can be separated by the fuel-salt reprocessing, or it is such a tiny fraction that it can be left in the salt and fissioned by excess neutrons. If the transuranics are left in, the separated wastes are pure Uranium fission wastes. All have half-lives less than 30 years. Reprocessed waste from thorium is therefore less radioactive than natural ores in 300 years.
• The thorium breeder reactor uses low-energy "thermal" neutrons very similar to light water reactors. It is therefore much safer than the touchy fast-neutron breeder reactors that the uranium-to-plutonium fuel cycle requires. The thorium fuel cycle therefore combines safe reactors, a long-term source of abundant fuel, and no need for expensive fuel-enrichment facilities.
• A molten salt reactor's fuel can be continuously reprocessed with a small adjacent chemical plant. Weinberg's groups at Oak Ridge National Laboratory found that a very small reprocessing facility can service a large 1 GW power plant: All the salt has to be reprocessed, but only every ten days. Society's total inventory of expensive, poisonous radioactives is therefore much less than in a conventional light-water-reactor's fuel cycle, which moves entire cores to recycling plants. Also, everything except fuel and waste stays inside the plant. The reprocessing cycle is:
  • A sparge of fluorine to remove U233 fuel from the salt. This has to be done before the next step.
  • A 4-meter-tall molten bismuth column separates protactinium from the fuel salt.
  • A small storage facility to let the protactinium from the bismuth column decay to U233. With a 27 day half life, ten months of storage assures that 99.9% decays to U233 fuel.
  • A small vapor-phase fluoride-salt distillation system distills the salts. Each salt has a distinct temperature of vaporization. The light carrier salts evaporate at low temperatures, and form the bulk of the salt. The thorium salts must be separated from the fission wastes at higher temperatures. The amounts involved are about 800 kg of waste per year per GW generated, so the equipment is very small. Salts of long-lived transuranic metals go back into the reactor as fuel.
• With continuous reprocessing, a molten-salt-fueled reactor has more than 97% burn-up of fuel. This is very efficient, compared to any system, anywhere. Light water reactors burn up about 2% of fuel on a once-through fuel cycle (current practice, 2007).
• With salt distillation, an MSFR can burn Plutonium, or even fluoridated nuclear waste from light water reactors.
• The molten-salt-fueled reactor operates much hotter than LWR reactors, from 650 °C on conservative designs, to as hot as 950 °C on aggressive designs. So very efficient Brayton cycle (gas turbine) generators are possible. This is also very efficient, a goal of "generation IV reactors" that has already been achieved by MSRs. This reduces fuel use and auxiliary equipment (major capital expenses) by 50% or more.
• MSRs work in small sizes, as well as large, so a utility could easily build several small reactors (say 100 MWe) from income, reducing interest expense and business risks.
• Molten salt fuel reactors are not experimental. Several have been constructed and operated at 650 °C temperatures for extended times, with simple, practical validated designs. There's no need for new science at all, and very little risk in engineering new, larger or modular designs.
• The reactor, like all nuclear plants, has little effect on biomes. In particular, it uses only small amounts of land, relatively small amounts of construction, and the waste is separated from the biome, unlike both fossil and renewable energy projects.

Combining the above, some form of molten-salt thorium breeder could be the most efficient well-developed energy source known, whether measured by cost per kW, capital cost or social costs.

There are some design and social advantages:

• Thorium's fuel cycle resists proliferation in two ways:
  • It is verifiable because the epithermal thorium breeder produces only at most 9% more fuel than it burns in each year. Building bombs quickly will take power plants out of operation.
  • Also, an easy variation of the thorium fuel cycle would contaminate the Th232 breeding material with chemically inseparable Th230. The Th230 breeds into U232, which has a powerful gamma emitter in its decay chain (Tl-208) that makes the reactor fuel U233/U232 impractical in a bomb, because it harms electronics.
• Thorium is more abundant than uranium. The Earth's crust has about three times as much.
• Thorium is cheap. Currently, it costs US$ 30/kg.
• Control of the salt's corrosivity is easy. The uranium buffers the salt, forming more UF4 from UF3 as more F is present. UF3 can be regenerated by adding small amounts of metallic beryllium to absorb F. In the MSRE, a beryllium rod was inserted into the salt until the Uf3 was the correct concentration.^[2]
• Extensive validation (fuel rod design validation normally takes years and prevents effective deployment of new nuclear technologies) is not needed. The fuel is molten, chemical reprocessing eliminates reaction products, and there are tested fuel mixtures, notably FLi7BeU.
• There is no need for fuel fabrication. This greatly reduces the MSR's fuel expenses. It poses a business challenge, because reactor manufacturers customarily get their long-term profits from fuel fabrication. A government agency could, however, type-license a design, which utilities could replicate.
• Molten-fuel reactors can be made inherently safe: Tested fuel-salt mixtures have negative reactivity coefficients, so that they decrease power generation as they get too hot. Most fuel-salt reactor vessels also have a freeze plug at the bottom that has to be actively cooled. If the cooling fails, the fuel drains to a subcritical storage facility.
• Continuous reprocessing simplifies numerous reactor design and operating issues. For example, the poisoning effects from xenon-135 are not present. Neutron poisoning from fission products is continuously mitigated. Transuranics, the frighteningly long-lived "wastes" of light water reactors, are burned as fuel.
• A fuel-salt reactor is mechanically and neutronically simpler than light-water reactors. There are only two items in the core: fuel salts and moderators. This reduces concerns with moderating interactions with positive void coefficients as water boils, chemical interactions, etc.
• Coolant and piping need never enter the high-neutron-flux zone, because the fuel is used to cool the core. The fuel is cooled in low-neutron-flux heat-exchangers outside the core. This reduces worries about neutron effects on pipes, testing, development issues, etc.
• The salt distillation process means that chemical separation and recycling of fission products, say for nuclear batteries, is actually cheap. Xenon and other valuable transmuted noble gases separate out of the molten fuel in the pump-bowl. Any transuranics go right back into the fuel for burn-up.

Molten salt reactors, nevertheless, present a number of design challenges. Known issues include:

• Since it uses unfabricated fuel, basically just a mixture of chemicals, current reactor vendors don't want to develop it. They derive their long-term profits from sales of fabricated fuel assemblies.
• Uncooled graphite moderators can cause some geometries of this reactor to increase in reactivity with higher temperatures, making such designs unsafe. Careful design may fix this, however.
• High neutron fluxes and temperatures in a compact MSR core can rapidly change the shape of a graphite moderator element, to require refurbishing in as little as four years. Eliminating graphite from sealed piping was a major incentive to switch to a single-fluid design.^[3] Most MSR designs do not use graphite as a structural material, and arrange for it to be easy to replace. At least one design used graphite balls floating in salt, which could be removed and inspected continuously without shutting down the reactor.^[4]
• A safe thorium breeder reactor using slow thermal-energy neutrons also has a low breeding rate. Each year it can only breed thorium into about 109% of the U233 fuel it consumes. This means that obtaining enough U233 for a new reactor can take eight years or more, which would slow deployment of this type of reactor. Most practical, fast deployment plans would start the new Thorium reactors with Plutonium from existing light-water reactor wastes or decommissioned nuclear weapons. This scheme also decreases society's stock of high-level wastes.
• The high neutron density in the core rapidly transmutes most isotopes of lithium to tritium, a radioactive isotope of hydrogen. In an MSR, the tritium forms hydrogen fluoride (HF). Tritium HF is a corrosive, chemically poisonous, radiotoxic gas. All MSR designs used very expensive isotopically purified lithium-7 for their carrier salts in order to reduce tritium formation as far as possible. The MSRE proved that this worked.
• Some slow corrosion occurs even in the exotic nickel alloy, Hastelloy-N used for the reactor. The corrosion is more extreme if the reactor is exposed to hydrogen which forms corrosive HF gas. Mere exposure to water-vapor causes uptake of corrosive amounts of hydrogen, so practical MSRs operate the salt under a blanket of dry inert gas, usually helium.
• When cold, the fuel salts radiogenically produce poisonous fluorine gas. The salts should be defueled and wastes removed before extended shutdowns. Unfortunately, this was discovered the unpleasant way, while the MSRE was shut-down over a 20-year period.
• The salt mixture is toxic, and water-soluble. The reactor design must therefore isolate the salt from the biome. This is a normal reactor safety requirement anyway.

An MSR based on chloride salts has many of the same advantages. However, the larger, less-dense atoms of chlorine causes the reactor to be a fast breeder. Theoretically, it wastes even fewer neutrons and breeds more efficiently, though it may be less safe. It would require a salt with an isotopically-separated chlorine, Cl37, to prevent neutronic activation of the chlorine into sulfur which would form corrosive sulfur chloride.

[edit] Molten-salt cooled reactors

Molten-salt-fueled reactors (MSFR) are quite different from molten-salt-cooled reactors (MSCR), a Gen IV proposal. The MSCR cannot reprocess fuel easily and has fuel rods that need to be fabricated and validated, delaying deployment by up to twenty years from project inception. However, since it uses fabricated fuel, reactor manufacturers can still proft by selling fuel assemblies. Also, the reactor's core retains many safety and cost advantages. Notably, there's no steam in the core to cause an explosion, and no large, expensive steel pressure vessel. Since it can operate at high temperatures, the conversion of the heat to electricity can also use an efficient, light weight Brayton cycle gas turbine.

Much of the current research on MSCRs is focused on small compact heat exchangers. By using smaller heat exchangers, less molten salt needs to be used and therefore significant cost savings could be achieved.

Molten salts can be highly corrosive, more so as temperatures rise. For the primary cooling loop of the MSR, a material is needed that can withstand corrosion at high temperatures and intense radiation. Experiments show that Hastelloy-N and similar alloys are quite suited to the tasks at operating temperatures up to about 700 °C. However, long-term experience with a production scale reactor has yet to be gained. Higher operating temperatures would be desirable, especially since at 850 °C thermo chemical production of hydrogen becomes possible. Materials for this temperature range have not yet been found, though carbon composites, carbides, and refractory metal based or ODS alloys might be feasible.

[edit] Fused salt selection

The types of fused salts that are chosen come from an optimization of salt characteristics. Fused fluorides are generally chosen over other salts because of the usefulness of the elements without isotope separation, better neutron economy and moderating efficiency, lower vapor pressure and better chemical stability. Chlorides have also been considered for molten salt reactors, but not nearly as much work has been done on reactor designs that utilize them. Additionally, whenever lithium fluoride is used as part of the salt composition, the lithium must be enriched to a very high purity (99.999%?) in lithium-7 to get tritium production under control.

Due to the high "redox window" available for fused fluoride salts, allowing for the chemical potential of the fused salt system to be manipulated, the following types of salts are the most promising. FLiBe can be used in conjunction with beryllium additions to drive down the electrochemical potential and virtually eliminate corrosion issues. However, beryllium is extremely toxic to humans. Many other salts have potential corrosion issues, especially at the elevated temperatures being talked about for future hydrogen production facilities.

To date, most research has focused on FLiBe for the nuclear heat transport system, for the obvious reasons that 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 has a measurable benefit to the neutron economy of the reactor, due to neutron doubling. This process occurs when the Beryllium nucleus re-emit two neutrons after absorbing a single neutron. For the fuel carrying salts, generally 1% or 2% by mole fraction of UF₄ is added, however thorium and plutonium fluorides have also been used. The MSFR is the only system that has run a single reactor, the MSRE, from all three known nuclear fuels.

Material	Total Neutron Capture Relative to Graphite (per unit volume)	Moderating Ratio (Avg. 0.1 to 10 eV)
Heavy Water	0.2	11,449
Light Water	75	246
Graphite	1	863
Sodium	47	2
UCO	285	2
UO2	3,583	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

Above is a table comparing the neutron capture and moderating efficiency of several materials. Red are Be bearing salts, blue are ZrF₄ bearing salts, and green are LiF bearing salts. (Source: ORNL/TM-2005/218, Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR), December 2005, D. T. Ingersoll)

[edit] Fused salt purification and reprocessing

Salts must be extremely pure initially, and would most likely be continuously cleaned in a large-scale molten salt reactor. Any water vapor in the salt will form hydrofluoric acid (HF) which is extremely corrosive. Other impurities can cause non-beneficial chemical reactions and would most likely have to be cleansed from the system. It should be noted that most power plants have to ensure that the primary coolant they are using is extremely pure; otherwise, they would encounter corrosion issues as well.

The possibility of online reprocessing can be an advantage of the MSR design. Continuous reprocessing ensures a low inventory of fission products at all times, which improves neutron economy. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. To allow breeding from thorium, the intermediate product protactinium has to be removed from the reactor and stored for some months while it decays into uranium 233. Left in the fuel it would absorb too many neutrons to make breeding with a graphite moderator and thermal spectrum possible (though with alternate designs in which the thorium is kept in a separate fluid from the fuel, the protactinium can simply be diluted with a larger volume of thorium fluid which proportionately reduces the neutron absorption; also some heavy water moderated reactor designs could overcome this, albeit at a lower thermal efficiency). The necessary reprocessing technology, which has to process the complete fuel every 10 days, has only been demonstrated at laboratory scale. For a power reactor such a large reprocessing facility is currently deemed uneconomic.

[edit] Political issues

To exploit the molten salt reactor's breeding potential to the fullest, the reactor must be co-located with a reprocessing facility. Any kind of nuclear reprocessing is still illegal in many countries. Some people fear that operating an MSR could pave the way to the plutonium economy with its associated proliferation dangers. (A similar argument lead to the shutdown of the Integral Fast Reactor project in 1994.)

Because of the recent lack of orders for new nuclear reactors, the nuclear industry business involved the selling fuel bundles to reload reactors and providing services for the reactor operators to keep the reactors running.^{[verification needed]} The fuel fabrication and servicing business is highly competitive, and only a few vendors who have the depth of experience have survived. The business model for molten-salt fueled reactors would be similar to the model for light water reactors, although it would not involve fabricating fuel assemblies. The nuclear industry would need to develop experience and confidence in the viability of molten salt reactors before it is commercialized. This would involve building demonstration plants and developing decades of operating experience.^{[verification needed]}

[edit] Comparison to ordinary light water reactors

Molten salt reactors are an immature technology. No large-scale reactor has been built and operated for a long period, and unexpected problems are likely. Whether an MSR will be economically and technologically viable is unknown. Small, experimental MSRs have operated as long as several years, and the problems were fixed.

MSRs may be safer. Molten salts trap fission products chemically, and react slowly or not at all in air. Also, the fuel salt does not burn in air or water. The core and primary cooling loop is operated at atmospheric pressure, and has no steam, so a pressure explosion is impossible. Even in the unlikely case of an accident, most radioactive fission products would stay in the salt instead of dispersing into the atmosphere. A molten core is meltdown-proof, so the worst possible accident would be a leak. In this case, the fuel salt can be drained into passively cooled storage, managing the accident. Neutron-producing accelerators have even been proposed for some super-safe subcritical experimental designs.

Some types of molten salt reactors are very efficient. Since the core and primary coolant loop are low pressure, it can be constructed of thin, relatively inexpensive weldments. So, it can be far less expensive than the massive pressure vessel required by the core of a light water reactor. Also, some form of fluid-fueled thorium breeder could use less fissile material per megawatt than any other reactor. Molten salt reactors can run at extremely high temperatures, with extremely high efficiencies when producing electricity. The temperatures are high enough to produce process heat for hydrogen production or other chemical reactions. Because of this, they have been included in the GEN-IV roadmap for further study.

Molten-salt-fueled thorium breeders close the nuclear fuel cycle and potentially eliminate the need for both fuel enrichment and fuel fabrication, both major expenses.

The MSR also has far better neutron economy and, depending on the design, a harder neutron spectrum. So, it can operate with less reactive fuels. Some designs (such as the MSRE) can operate a single design from all three common nuclear fuels. For example, it can breed from uranium-238, thorium or even burn the transuranic spent nuclear fuel from light water reactors. In contrast, a water-cooled reactor cannot completely consume the plutonium it produces, because the increasing impurities from the fission wastes capture too many neutrons, "poisoning" the reaction.

Molten salt-fueled thorium breeders can operate for extended periods, possibly decades, without refueling, by chemically precipitating neutronic poisons.

MSRs scale over a wide range of powers. Reactors as small as several megawatts have been constructed and operated. Theoretical designs up to several gigawatts have been proposed^[5].

Because of their low structures and compact cores, MSRs weigh less per watt (that is, they have a greater "specific power") than other proven reactor designs. So, in small sizes, with long refueling intervals, they are an excellent choice to power vehicles, including ships, aircraft and spacecraft.

[edit] References

1. ^ FUJI Reactor, in the MSR article of the Encyclopedia of the Earth
2. ^ Briant, Raymond C.; Alvin M. Weinberg (1957). Molten Fluorides as Power Reactor Fuels (PDF). Nuclear Science and Engineering;  2, 797–803. Retrieved on 2008-05-18.
3. ^ Energy from Thorium blog->Reactor Design->Graphite and Two-Fluid vs. One-Fluid LFRs Viewed 6/2007
4. ^ ORNL-4548: Molten-Salt Reactor Program: Semiannual Progress Report for Period Ending February 28, 1970, pg. 57
5. ^ Weinberg et al. WASH 1080, ORNL

• Energy from Thorium's Document Repository Contains scanned versions of many of the U.S. government engineering reports. This repository is the main reference for the molten-salt fueled reactor's technical discussion.

• Moir, R.W; Cost of Electricity from MSRs; Lawrence Livermore National Laboratory, U.S.

• J.H. Devan et al. (unknown date). Material Considerations for Molten Salt Accelerator-based Plutonium Conversion Systems, pg. 475-486

• W.D. Manely et al. (1960). Metallurgical Problems in Molten Fluoride Systems. Progress in Nuclear Energy, VOl. 2, pg. 164-179

• http://energy.inel.gov/gen-iv/msr.shtml

• Bruce Hoglund's Eclectic Interests Home Page Nuclear Power, Thorium, Molten Salt reactors, etc.

• "The First Nuclear Era : The Life and Times of a Technological Fixer", by Alvin Martin Weinberg (1994). Book by a former director of the Oak Ridge National Laboratory, and a promoter of nuclear power and molten salt reactors.

• Generation IV International Forum MSR website

• INL MSR workshop summary

• Molten Salt Chemistry Plays a Prominant (sic) Role in Accelerator-Driven Transmutation Systems

[edit] See also

• Generation IV reactor
• nuclear power
• nuclear reactor
• nuclear fuel
• nuclear material
• atomic physics
• Liquid fluoride reactor

[edit] External links

• Energy from Thorium Blog
• Nuclear Engineering
• Wikibooks

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