Liquid fluoride thorium reactor
The liquid fluoride thorium reactor (acronym LFTR; pronounced lifter) is a type of molten salt reactor. LFTRs use the thorium fuel cycle with a fluoride-based, molten, liquid salt for fuel. It can achieve high operating temperatures at atmospheric pressure.
LFTR is a type of thorium molten salt reactor (TMSR). Molten-salt-fueled reactors (MSRs) supply the nuclear fuel in the form of a molten salt mixture. They should not be confused with molten salt-cooled high temperature reactors (fluoride high-temperature reactors, FHRs) that use a solid fuel. Molten salt reactors, as a class, include both burners and breeders in fast or thermal spectra, using fluoride or chloride salt-based fuels and a range of fissile or fertile consumables. LFTRs are defined by the use of fluoride fuel salts and the breeding of thorium into uranium-233 in the thermal spectrum.
In a LFTR, thorium and uranium-233 are dissolved in carrier salts, forming a liquid fuel. In a typical operation, the liquid is pumped between a critical core and an external heat exchanger where the heat is transferred to a nonradioactive secondary salt. The secondary salt then transfers its heat to a steam turbine or closed-cycle gas turbine. This technology was first investigated at the Oak Ridge National Laboratory Molten-Salt Reactor Experiment in the 1960s. It has recently been the subject of a renewed interest worldwide. Japan, China, the UK and private US, Czech, Canadian and Australian companies have expressed intent to develop and commercialize the technology. LFTRs differ from other power reactors in almost every aspect: they use thorium rather than uranium, operate at low pressure, receive fuel by pumping without shutdown, entail no risk of nuclear meltdown, use a salt coolant and produce higher operating temperatures. These distinctive characteristics give rise to many potential advantages, as well as design challenges.
- 1 Background
- 2 Breeding basics
- 3 Reactor primary system design variations
- 4 Power generation
- 5 Removal of fission products
- 6 Advantages
- 7 Disadvantages
- 8 Recent developments
- 9 See also
- 10 References
- 11 Further reading
- 12 External links
- Uranium-235, which is already fissile, and occurs as 0.72% of natural uranium
- Plutonium-239, which can be bred from non-fissile uranium-238 (>99% of natural uranium)
- Uranium-233, which can be bred from non-fissile thorium-232 (~100% of natural thorium; which has about four times greater abundance in the earth's crust than uranium)
Th-232, U-235 and U-238 are primordial nuclides, having existed in their current form for over 4.5 billion years, predating the formation of the Earth; they were forged in the cores of dying stars through the r-process and scattered across the galaxy by supernovas. Their radioactive decay produces about half of the earth's internal heat.
For technical and historical reasons, the three are each associated with different reactor types. U-235 is the world's primary nuclear fuel and is usually used in light water reactors. U-238/Pu-239 has found the most use in liquid sodium fast breeder reactors and CANDU Reactors. Th-232/U-233 is best suited to molten salt reactors (MSR).
Alvin M. Weinberg pioneered the use of the MSR at Oak Ridge National Laboratory. At ORNL, two prototype molten salt reactors were successfully designed, constructed and operated. These were the Aircraft Reactor Experiment in 1954 and Molten-Salt Reactor Experiment from 1965 to 1969. Both test reactors used liquid fluoride fuel salts. The MSRE notably demonstrated fueling with U-233 and U-235 during separate test runs.(pix) Weinberg was removed from his post and the MSR program closed down in the early 1970s, after which research stagnated in the United States. Today, the ARE and the MSRE remain the only molten salt reactors ever operated.
In a nuclear power reactor, there are two types of fuel. The first is fissile material, which splits when hit by neutrons, releasing a large amount of energy and also releasing two or three new neutrons. These can split more fissile material, resulting in a continued chain reaction. Examples of fissile fuels are U-233, U-235 and Pu-239. The second type of fuel is called fertile. Examples of fertile fuel are Th-232 (mined thorium) and U-238 (mined uranium). Often the amount of fertile fuel in the reactor is far bigger than the amount of fissile, but it cannot be fissioned directly. It must first absorb one of the 2 or 3 neutrons produced in the fission process, which is called neutron capture, then it becomes a fissile isotope by radioactive decay. This process is called breeding.
All reactors breed some fuel this way, but today's solid fueled thermal reactors don't breed enough new fuel from the fertile to make up for the amount of fissile they consume. This is because today's reactors use the mined uranium-plutonium cycle in a moderated neutron spectrum. Such a fuel cycle, using slowed down neutrons, gives back less than 2 new neutrons from fissioning the bred plutonium. Since 1 neutron is required to sustain the fission reaction, this leaves a budget of less than 1 neutron per fission to breed new fuel. In addition, the materials in the core such as metals, moderators and fission products absorb some neutrons, leaving too few neutrons to breed enough fuel to continue operating the reactor. As a consequence they must add new fissile fuel periodically and swap out some of the old fuel to make room for the new fuel.
In a reactor that breeds at least as much new fuel as it consumes, it is not necessary to add new fissile fuel. Only new fertile fuel is added, which breeds to fissile inside the reactor. In addition the fission products need to be removed. This type of reactor is called a breeder reactor. If it breeds just as much new fissile from fertile to keep operating indefinitely, it is called a break-even breeder or isobreeder. A LFTR is usually designed as a breeder reactor: thorium goes in, fission products come out.
Reactors that use the uranium-plutonium fuel cycle require fast reactors to sustain breeding, because only with fast moving neutrons does the fission process provide more than 2 neutrons per fission. With thorium, it is possible to breed using a thermal reactor. This was proven to work in the Shippingport Atomic Power Station, whose final fuel load bred slightly more fissile from thorium than it consumed, despite being a fairly standard light water reactor. Thermal reactors require less of the expensive fissile fuel to start.
There are two ways to configure a breeder reactor to do the required breeding. One can place the fertile and fissile fuel together, so breeding and splitting occurs in the same place. Alternatively, fissile and fertile can be separated. The latter is known as core-and-blanket, because a fissile core produces the heat and neutrons while a separate blanket does all the breeding.
Reactor primary system design variations
Oak Ridge investigated both ways to make a breeder for their molten salt breeder reactor. Because the fuel is liquid, they are called the "single fluid" and "two fluid" thorium thermal breeder molten salt reactors.
Single fluid reactor
The one-fluid design includes a large reactor vessel filled with fluoride salt containing thorium and uranium. Graphite rods immersed in the salt function as a moderator and to guide the flow of salt. In the ORNL MSBR design a reduced amount of graphite near the edge of the reactor core would make the outer region under-moderated, and increased the capture of neutrons there by the thorium. With this arrangement, most of the neutrons were generated at some distance from the reactor boundary, and reduced the neutron leakage to an acceptable level. Still, a single fluid design needs a considerable size to permit breeding.
In a breeder configuration, extensive fuel processing was specified to remove fission products from the fuel salt.(p181) In a converter configuration fuel processing requirement was simplified to reduce plant cost. The trade-off was the requirement of periodic uranium refueling.
The MSRE was a core region only prototype reactor. The MSRE provided valuable long-term operating experience. According to estimates of Japanese scientists, a single fluid LFTR program could be achieved through a relatively modest investment of roughly 300-400 million dollars over 5–10 years to fund research to fill minor technical gaps and build a small reactor prototype comparable to the MSRE.
Two fluid reactor
The two-fluid design is mechanically more complicated compared to the "single fluid" reactor design. The "two fluid" reactor has a high-neutron-density core that burns uranium-233 from the thorium fuel cycle. A separate blanket of thorium salt absorbs the neutrons and its thorium is converted to protactinium-233. Protactinium-233 can be left in the blanket region where neutron flux is lower, so that it slowly decays to U-233 fissile fuel, rather than capture neutrons. This bred fissile U-233 can be recovered by simple fluorination, and placed in the core to fission. The core's salt is also purified, first by fluorination to remove uranium, then vacuum distillation to remove and reuse the carrier salts. The still bottoms left after the distillation are the fission products waste of a LFTR.
The advantages of separating the core and blanket fluid include:
- Simplified fuel processing. Thorium is chemically similar to several fission products, called lanthanides. With thorium in a separate blanket, thorium is kept isolated from the lanthanides. Without thorium in the core fluid, removal of lanthanide fission products is simplified.
- Low fissile inventory. Because the fissile fuel is concentrated in a small core fluid, the actual reactor core is more compact. There is no fissile material in the outer blanket that contains the fertile fuel for breeding. Because of this, the 1968 ORNL design required just 315 kilograms of fissile materials to start up a 250 MW(e) two fluid MSBR reactor.(p35) This reduces the cost of the initial fissile startup charge, and allows more reactors to be started up on any given amount of fissile material.
- More efficient breeding. The thorium blanket can effectively capture leaked neutrons from the core region. There is nearly zero fission occurring in the blanket, so the blanket itself does not leak significant numbers of neutrons. This results in a high efficiency of neutron use (neutron economy), and a higher breeding ratio, especially with small reactors.
One design weakness of the two-fluid design was the necessity for a barrier wall between the core and the blanket region, a wall that would have to be replaced periodically because of fast neutron damage.(p29) Graphite was the chosen material by ORNL because of its low neutron absorption, compatibility with the molten salts, high temperature resistance, and sufficient strength and integrity to separate the fuel and blanket salts. The effect of neutron radiation on graphite is to slowly shrink and then swell the graphite to cause an increase in porosity and a deterioration in physical properties.(p13) Graphite pipes would change length, and may crack and leak. ORNL chose not to pursue the two-fluid design, and no examples of the two-fluid reactor were ever constructed.
One additional design weakness of the two-fluid design was its complex plumbing. ORNL thought it necessary to use complex interleaving of the core and blanket piping in order to get a high reactor power level with acceptably low power density.(p4) More recent research has put into question the need for complex interleaving graphite tubing, suggesting a simple elongated tube-in-shell reactor would allow high total reactor power without complex tubing.(p6)
The recovery of high-purity uranium-233 has been raised as a potential nuclear proliferation concern.(p99) A design with no protactinium separation would ensure that any U-233 is contaminated with U-232 whose decay chain emits 2 MeV gamma rays too hazardous for weapons workers.
Hybrid "one and a half fluid" reactor
A two fluid reactor that has thorium in the fuel salt is sometimes called a "one and a half fluid" reactor, or 1.5 fluid reactor. This is a hybrid, with some of the advantages and disadvantages of both 1 fluid and 2 fluid reactors. Like the 1 fluid reactor, it has thorium in the fuel salt, which complicates the fuel processing. And yet, like the 2 fluid reactor, it can use a highly effective separate blanket to absorb neutrons that leak from the core. The added disadvantage of keeping the fluids separate using a barrier remains, but with thorium present in the fuel salt there are fewer neutrons that must pass through this barrier into the blanket fluid. This results in less damage to the barrier. Any leak in the barrier would also be of lower consequence, as the processing system must already deal with thorium in the core.
The main design question when deciding between a 1/1.5 fluid or two fluid LFTR is whether a more complicated reprocessing or a more demanding structural barrier will be easier to solve.
|Design concept||Breeding ratio||Fissile inventory|
|Single-fluid, 30 year graphite life, fuel processing||1.06||2300 kg|
|Single-fluid, 4 year graphite life, fuel processing||1.06||1500 kg|
|1.5 fluid, replaceable core, fuel processing||1.07||900 kg|
|Two-fluid, replaceable core, fuel processing||1.07||700 kg|
The LFTR with a high operating temperature of 700 degrees Celsius can operate at a thermal efficiency to electrical of 45%. This is higher than today's light water reactors (LWRs) that are at 32-36% thermal to electrical efficiency.
The Rankine cycle is the most basic thermodynamic power cycle. The simplest cycle consists of a steam generator, a turbine, a condenser, and a pump. The working fluid is usually water. A Rankine power conversion system coupled to a LFTR could take advantage of increased steam temperature to improve its thermal efficiency. The subcritical Rankine steam cycle is currently used in commercial power plants, with the newest plants utilizing the higher temperature, higher pressure, supercritical Rankine steam cycles. The work of ORNL from the 1960s and 1970s on the MSBR assumed the use of a standard supercritical steam turbine with an efficiency of 44%,(p74) and had done considerable design work on developing molten fluoride salt – steam generators.
The working gas of a Brayton cycle can be helium, nitrogen, or carbon dioxide. The high-pressure working gas is expanded in a turbine to produce power. The low-pressure warm gas is cooled in an ambient cooler. The low-pressure cold gas is compressed to the high-pressure of the system. Often the turbine and the compressor are mechanically connected through a single shaft. High pressure Brayton cycles are expected to have a smaller generator footprint compared to lower pressure Rankine cycles. A Brayton cycle heat engine can operate at lower pressure with wider diameter piping. The world's first commercial Brayton cycle solar power module (100 kW) was built and demonstrated in Israel's Arava Desert in 2009.
- Industrial process heat for many uses, such as ammonia production with the Haber process.
- Desalination of water
- Hydrogen production by water splitting
- Combined heat and power
- Nuclear marine propulsion[unreliable source?]
Removal of fission products
||This section may be too technical for most readers to understand. (April 2015)|
The LFTR needs a mechanism to remove the fission products from the fuel salt and recover at least the fissile material. Some fission products in the salt absorb neutrons and reduce the production of new fissile fuel. Especially the concentrations of some of the rare earth elements need to be kept low, as they have a large cross section for neutron capture. Some other elements with a small cross section like Cs or Zr can be tolerated in much higher concentrations, so they may accumulate over years of operation.
Removal of fission products is similar to reprocessing of solid fuel elements, without the need to remove and rebuild the fuel cladding. As the fuel of a LFTR is a molten salt mixture, it is attractive to use pyroprocessing, high temperature methods working directly from the hot molten salt. Pyroprocessing does not use radiation sensitive solvents and is not easily disturbed by decay heat. It can be used on the highly radioactive fuel directly from the reactor. Having the chemical separation on site, close to the reactor avoids transport and keeps the total inventory of the fuel cycle low. Ideally everything except new fuel (thorium) and waste (fission products) stays inside the plant.
On site processing is planned to work continuously, cleaning a small fraction of the salt every day and sending it back to the reactor. There is no need to make the fuel salt very clean; the purpose is to keep the concentration of fission products and other impurities (e.g. oxygen) low enough.
The more noble metals (Pd, Ru, Ag, Mo, Nb, Sb, Tc) do not form fluorides in the normal salt, but form fine metallic particles in the salt. They can plate out at metal surfaces like the heat exchanger or some kinds of high surface area filters that are easier to remove. Still there is some uncertainty where these noble elements end up, as the MSRE only provided a relatively short operating experience and independent laboratory experiments are difficult.
Some elements like Xe and Kr come out easily as gas, assisted by a sparge of helium. In addition a part of the "noble" metals are removed together with the gas as a fine mist. Especially the fast removal of Xe-135 is important, as this is a very strong neutron poison and makes reactor control more difficult if left in the reactor. Removal of Xe also improves neutron economy. The gas (mainly He, Xe and Kr) is held up for about 2 days until a large fraction of the Xe-135 and other short lived isotopes have decayed. Most of the gas can then be recycled. After an additional hold up of several months, radioactivity is low enough to separate the gas at low temperatures into helium (for reuse), xenon (for sale) and krypton. The krypton needs storage (e.g. in compressed form) for an extended time (several decades) to wait for the decay of Kr-85.(p274)
For cleaning the salt mixture several methods of chemical separation were proposed. Compared to classical PUREX reprocessing, pyroprocessing can be more compact and produce less secondary waste. The pyroprocesses of the LFTR salt already starts with a suitable liquid form, so it may be less expensive than using solid oxide fuels. However, because no complete molten salt reprocessing plant has been built, all testing has been limited to the laboratory, and with only a few elements. There is still more research and development needed to improve separation and make reprocessing more economically viable.
Uranium and some other elements can be removed from the salt by a process called fluorine volatility: A sparge of fluorine removes volatile high-valence fluorides as a gas. This is mainly uranium hexafluoride, containing the uranium-233 fuel, but also neptunium hexafluoride, technetium hexafluoride and selenium hexafluoride, as well as fluorides of various highly radioactive short-lived fission products such as iodine-131, 99molybdenum, and 132tellurium. The volatile fluorides can be further separated by adsorption and distillation. Handling uranium hexafluoride is well established in enrichment. The higher valence fluorides are quite corrosive at high temperatures and require more resistant materials than Hastelloy. One suggestion in the MSBR program at ORNL was using solidified salt as a protective layer. At the MSRE reactor fluorine volatility was used to remove uranium from the fuel salt. Also for use with solid fuel elements fluorine volatility is quite well developed and tested.
Another simple method, tested during the MSRE program, is high temperature vacuum distillation. The lower boiling point fluorides like uranium tetrafluoride and the LiF and BeF carrier salt can be removed by distillation. Under vacuum the temperature can be lower than the ambient pressure boiling point. So a temperature of about 1000 °C is sufficient to recover most of the FLiBe carrier salt. However, while possible in principle, separation of thorium fluoride from the even higher boiling point lanthanide fluorides would require very high temperatures and new materials. The chemical separation for the 2-fluid designs, using uranium as a fissile fuel can work with these two relatively simple processes: Uranium from the blanket salt can be removed by fluorine volatility, and transferred to the core salt. To remove the fissile products from the core salt, first the uranium is removed via fluorine volatility. Then the carrier salt can be recovered by high temperature distillation. The fluorides with a high boiling point, including the lanthanides stay behind as waste.
The early Oak Ridge's chemistry designs were not concerned with proliferation and aimed for fast breeding. They planned to separate and store protactinium-233, so it could decay to uranium-233 without being destroyed by neutron capture in the reactor. With a half-life of 27 days, 2 months of storage would assure that 75% of the 233Pa decays to 233U fuel. The protactinium removal step is not required per se for a LFTR. Alternate solutions are operating at a lower power density and thus a larger fissile inventory (for 1 or 1.5 fluid) or a larger blanket (for 2 fluid). Also a harder neutron spectrum helps to achieve acceptable breeding without protactinium isolation.
If Pa separation is specified, this must be done quite often (for example, every 10 days) to be effective. For a 1 GW, 1-fluid plant this means about 10% of the fuel or about 15 t of fuel salt need to go through reprocessing every day. This is only feasible if the costs are much lower than current costs for reprocessing solid fuel.
Newer designs usually avoid the Pa removal and send less salt to reprocessing, which reduces the required size and costs for the chemical separation. It also avoids proliferation concerns due to high purity U-233 that might be available from the decay of the chemical separated Pa.
Separation is more difficult if the fission products are mixed with thorium, because thorium, plutonium and the lanthanides (rare earth elements) are chemically similar. One process suggested for both separation of protactinium and the removal of the lanthanides is the contact with molten bismuth. In a redox-reaction some metals can be transferred to the bismuth melt in exchange for lithium added to the bismuth melt. At low lithium concentrations U, Pu and Pa move to the bismuth melt. At more reducing conditions (more lithium in the bismuth melt) the lanthanides and thorium transfer to the bismuth melt too. The fission products are then removed from the bismuth alloy in a separate step, e.g. by contact to a LiCl melt. However this method is far less developed. A similar method may also be possible with other liquid metals like aluminum.
- Inherent safety. LFTR designs use a strong negative temperature coefficient of reactivity to achieve passive inherent safety against excursions of reactivity. The temperature dependence comes from 3 sources. The first is that thorium absorbs more neutrons if it overheats, the so-called Doppler effect. This leaves fewer neutrons to continue the chain reaction, reducing power. The second part is heating the graphite moderator, that usually causes a positive contribution to the temperature coefficient. The third effect has to do with thermal expansion of the fuel. If the fuel overheats, it expands considerably, which, due to the liquid nature of the fuel, will push fuel out of the active core region. In a small (e.g. the MSRE test reactor) or well moderated core this reduces the reactivity. However, in a large, under-moderated core (e.g. the ORNL MSBR design), less fuel salt means better moderation and thus more reactivity and an undesirable positive temperature coefficient.
- Stable coolant. Molten fluorides are chemically stable and impervious to radiation. The salts do not burn, explode, or decompose, even under high temperature and radiation. There are no rapid violent reactions with water and air that sodium coolant has. There is no combustible hydrogen production that water coolants have. However the salt is not stable to radiation at low (less than 100 C) temperatures due to radiolysis.
- Low pressure operation. Because the coolant salts remain liquid at high temperatures, LFTR cores are designed to operate at low pressures, like 0.6 MPa (comparable to the pressure in the drinking water system) from the pump and hydrostatic pressure. Even if the core fails, there is little increase in volume. Thus the containment building cannot blow up. LFTR coolant salts are chosen to have very high boiling points. Even a several hundred degree heatup during a transient or accident does not cause a meaningful pressure increase. There is no water or hydrogen in the reactor that can cause a large pressure rise or explosion as happened during the Fukushima Daiichi nuclear accident.[unreliable source?]
- Leak Resistance.Due to the low pressure operation and low pressure differences through the primary heat exchangers, the potential for large leaks is also greatly reduced.
- No pressure buildup from fission. LFTRs are not subject to pressure buildup of gaseous and volatile fission products. The liquid fuel allows for online removal of gaseous fission products, such as xenon, for processing, thus these decay products would not be spread in a disaster. Further, fission products are chemically bonded to the fluoride-salt, including iodine,[dubious ] cesium, and strontium, capturing the radiation and preventing the spread of radioactive material to the environment.
- Easier to control. A molten fuel reactor has the advantage of easy removal of xenon-135. Xenon-135, an important neutron absorber, makes solid fueled reactors difficult to control. In a molten fueled reactor, xenon-135 can be removed. In solid-fuel reactors, xenon-135 remains in the fuel and interferes with reactor control.
- Slow heatup. Coolant and fuel are inseparable, so any leak or movement of fuel will be intrinsically accompanied by a large amount of coolant. Molten fluorides have high volumetric heat capacity, some such as FLiBe, even higher than water. This allows them to absorb large amounts of heat during transients or accidents.
- Passive decay heat cooling. Many reactor designs (such as that of the Molten-Salt Reactor Experiment) allow the fuel/coolant mixture to escape to a drain tank, when the reactor is not running (see "Fail safe core" below). This tank is planned to have some kind (details are still open) of passive decay heat removal, thus relying on physical properties (rather than controls) to operate.
- Fail safe core. LFTRs can include a freeze plug at the bottom that has to be actively cooled, usually by a small electric fan. If the cooling fails, say because of a power failure, the fan stops, the plug melts, and the fuel drains to a subcritical passively cooled storage facility. This not only stops the reactor, also the storage tank can more easily shed the decay heat from the short-lived radioactive decay of irradiated nuclear fuels. Even in the event of a major leak from the core such as a pipe breaking, the salt will spill onto the kitchen-sink-shaped room the reactor is in, which will drain the fuel salt by gravity into the passively cooled dump tank.
- Less activated waste. LFTRs have very little structural material inside the core. Only the fuel salt, graphite, and small amounts of metals or composites are inside the actual reactor core. This reduces the amount of neutrons lost to structural components, improving the neutron economy, and reducing the amount of activated structural waste. Fluorine, lithium and beryllium do not have significant long term neutron activation.
- Less long lived waste. LFTRs can dramatically reduce the long-term radiotoxicity of their reactor wastes. Light water reactors with uranium fuel have fuel that is more than 95% U-238. These reactors normally transmute part of the U-238 to Pu-239, a long lived isotope. Almost all of the fuel is therefore only one step away from becoming a transuranic long lived element. Plutonium-239 has a half life of 24,000 years, and is the most common transuranic in spent nuclear fuel from light water reactors. Transuranics like Pu-239 cause the perception that reactor wastes are an eternal problem. In contrast, the LFTR uses the thorium fuel cycle, which transmutes thorium to U-233. Because thorium is a lighter element, more neutron captures are required to produce the transuranic elements. U-233 has two chances to fission in a LFTR. First as U-233 (90% will fission) and then the remaining 10% has another chance as it transmutes to U-235 (80% will fission). The fraction of fuel reaching neptunium-237, the most likely transuranic element, is therefore only 2%, about 15 kg per GWe-year. This is a transuranic production 20x smaller than light water reactors, which produce 300 kg of transuranics per GWe-year. Importantly, because of this much smaller transuranic production, it is much easier to recycle the transuranics. That is, they are sent back to the core to eventually fission. Reactors operating on the U238-plutonium fuel cycle produce far more transuranics, making full recycle difficult on both reactor neutronics and the recycling system. In the LFTR, only a fraction of a percent, as reprocessing losses, goes to the final waste. When these two benefits of lower transuranic production, and recycling, are combined, a thorium fuel cycle reduces the production of transuranic wastes by more than a thousand-fold compared to a conventional once-through uranium-fueled light water reactor. The only significant long lived waste is the uranium fuel itself, but this can be used indefinitely by recycling, always generating electricity. If the thorium stage ever has to be shut down, part of the reactors can be shut down and their uranium fuel inventory burned out in the remaining reactors, allowing a burndown of even this final waste to as small a level as society demands. The LFTR does still produce radioactive fission products in its waste, but they don't last very long - the radiotoxicity of these fission products is dominated by cesium-137 and strontium-90. The longer half-life is cesium: 30.17 years. So, after 30.17 years, decay reduces the radioactivity by a half. Ten half-lives will reduce the radioactivity by two raised to a power of ten, a factor of 1,024. Fission products at that point, in about 300 years, are less radioactive than natural uranium. What's more, the liquid state of the fuel material allows separation of the fission products not only from the fuel, but from each other as well, which enables them to be sorted by the length of each fission product's half-life, so that the ones with shorter half-lives can be brought out of storage sooner than those with longer half-lives.
- Destruction of existing long lived wastes. LFTRs can use existing transuranic wastes for their initial fissile startup charge better than any solid fueled reactor for various technical and physical reasons. Because the fuel is a liquid homogeneous solution, it is always perfectly mixed, impervious to radiation damage and can accept any composition of plutonium, neptunium, americium and curium up to the solubility limit. Solid fueled reactors, such as solid fueled fast reactors, while theoretically outperforming the LFTR in burning of these higher actinides, can only accept limited amounts of these higher actinides (neptunium, americium and curium are often called minor actinides). This is because the fuel is not perfectly mixed, as it is confined in solid fuel elements, and also because the coolant void coefficient (coolant overheating) can become positive for too high levels of minor actinides. In addition, manufacturing solid fuels with high amounts of americium and curium is also difficult due to decay heat generation and helium production rates. As a result, solid fueled reactors usually only use reprocessed plutonium but do not use the americium and curium, which constitute a large portion of the radiotoxicity of the long lived waste.
- Proliferation resistance. The LFTR resists diversion of its fuel to nuclear weapons in four ways: first, the 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 U-232 are also produced. U-232 has a decay chain product (thallium-208) that emits powerful, dangerous gamma rays. These are not a problem inside a reactor, but in a bomb, they complicate bomb manufacture, harm electronics and reveal the bomb's location. The second proliferation resistant feature comes from the fact that LFTRs produce very little plutonium, around 15 kg per gigawatt-year of electricity (this is the output of a single large reactor over a year). This plutonium is also mostly Pu-238, which makes it unsuitable for fission bomb building, due to the high heat and spontaneous neutrons emitted. The third track, a LFTR doesn't make much spare fuel. It produces at most 9% more fuel than it burns each year, and it's even easier to design a reactor that makes only 1% more fuel. With this kind of reactor, building bombs quickly will take power plants out of operation, and this is an easy indication of national intentions. And finally, use of thorium can reduce and eventually eliminate the need to enrich uranium. Uranium enrichment is one of the two primary methods by which states have obtained bomb making materials.
Economy and efficiency
- Thorium abundance. A LFTR breeds thorium into uranium-233 fuel. The Earth's crust contains about three to four times as much thorium as U-238 (thorium is about as abundant as lead). It is a byproduct of rare-earth mining, normally discarded as waste. Using LFTRs, there is enough affordable thorium to satisfy the global energy needs for hundreds of thousands of years. Thorium is more common in the earth’s crust than tin, mercury, or silver. A cubic meter of average crust yields the equivalent of about four sugar cubes of thorium, enough to supply the energy needs of one person for more than ten years if completely fissioned. Lemhi Pass on the Montana-Idaho border is estimated to contain 1,800,000 tons of high-grade thorium ore. Five hundred tons could supply all U.S. energy needs for one year. Due to lack of current demand, the U.S. government has returned about 3,200 metric tons of refined thorium nitrate to the crust, burying it in the Nevada desert.
- No shortage of natural resources. Sufficient other natural resources such as beryllium, lithium, nickel and molybdenum are available to build thousands of LFTRs.
- Reactor efficiency. Conventional reactors consume less than one percent of the mined uranium, leaving the rest as waste. With well working reprocessing LFTR may consume about 99% of its thorium fuel. The improved fuel efficiency means that 1 ton of natural thorium in a LFTR produces as much energy as 35 t of enriched uranium in conventional reactors (requiring 250 t of natural uranium), or 4,166,000 tons of black coal in a coal power plant.
- Thermodynamic efficiency. LFTRs operating with modern supercritical steam turbines would operate at 45% thermal to electrical efficiency. With future closed gas Brayton cycles, which could be used in a LFTR power plant due to its high temperature operation, the efficiency could be up to 54%. This is 20 to 40% higher than today's light water reactors (33%), resulting in the same 20 to 40% reduction in fissile and fertile fuel consumption, fission products produced, waste heat rejection for cooling, and reactor thermal power.
- No enrichment and fuel element fabrication. Since 100% of natural thorium can be used as a fuel, and the fuel is in the form of a molten salt instead of solid fuel rods, expensive fuel enrichment and solid fuel rods' validation procedures and fabricating processes are not needed. This greatly decreases LFTR fuel costs. Even if the LFTR is started up on enriched uranium, it only needs this enrichment once just to get started. After startup, no further enrichment is required.
- Lower fuel cost. The salts are fairly inexpensive compared to solid fuel production. For example, while beryllium is quite expensive per kg, the amount of beryllium required for a large 1 GWe reactor is quite small. ORNL's MSBR required 5.1 tons of beryllium metal, as 26 tons of BeF2. At a price of $147/kg BeF2,(p44) this inventory would cost less than $4 million, a modest cost for a multi billion dollar power plant. Consequently, a beryllium price increase over the level assumed here has little effect in the total cost of the power plant. The cost of enriched lithium-7 is less certain, at $120–800/kg LiF. and an inventory (again based on the MSBR system) of 17.9 tons lithium-7 as 66.5 tons LiF makes between $8 million and $53 million for the LiF. Adding the 99.1 tons of thorium @ $30/kg adds only $3 million. Fissile material is more expensive, especially if expensively reprocessed plutonium is used, at a cost of $100 per gram fissile plutonium. With a startup fissile charge of only 1.5 tons, made possible through the soft neutron spectrum this makes $150 million. Adding everything up brings the total cost of the one time fuel charge at $165 to $210 million. This is similar to the cost of a first core for a light water reactor. Depending on the details of reprocessing the salt inventory once can last for decades, whereas the LWR needs a completely new core every 4 to 6 years (1/3 is replaced every 12 to 24 months). ORNL's own estimate for the total salt cost of even the more expensive 3 loop system was around $30 million, which is less than $100 million in today's money.
- LFTRs are cleaner: as a fully recycling system, the discharge wastes from a LFTR are predominantly fission products, most of which have relatively short half lives (83% in hours or days compared to longer-lived actinide wastes of conventional nuclear power plants. This results in a significant reduction in the needed waste containment period in a geologic repository. The remaining 17% of waste products require only 300 years until reaching background levels. The radiotoxicity of the thorium fuel cycle waste is 10,000 times less than that of the uranium/plutonium fuel lifecycle.
- Less fissile fuel needed. Because LFTRs are thermal spectrum reactors, they need much less fissile fuel to get started. Only 1-2 tons of fissile are required to start up a single fluid LFTR, and potentially as low as 0.4 ton for a two fluid design. In comparison, solid fueled fast breeder reactors need at least 8 tons of fissile fuel to start the reactor. While fast reactors can theoretically start up very well on the transuranic waste, their high fissile fuel startup makes this very expensive.
- No downtime for refueling. LFTRs have liquid fuels, and therefore there is no need to shut down and take apart the reactor just to refuel it. LFTRs can thus refuel without causing a power outage (online refueling).
- Load following. As the LFTR does not have xenon poisoning, there is no problem reducing the power in times of low demand for electricity and turn back on at any time.
- No high pressure vessel. Since the core is not pressurized, it does not need the most expensive item in a light water reactor, a high-pressure reactor vessel for the core. Instead, there is a low-pressure vessel and pipes (for molten salt) constructed of relatively thin materials. Although the metal is an exotic nickel alloy that resists heat and corrosion, Hastelloy-N, the amount needed is relatively small.
- Excellent heat transfer. Liquid fluoride salts, especially LiF based salts, have good heat transfer properties. Fuel salt such as LiF-ThF4 has a volumetric heat capacity that is around 22% higher than water, FLiBe has around 12% higher heat capacity than water. In addition, the LiF based salts have a thermal conductivity around twice that of the hot pressurized water in a pressurized water reactor. This results in efficient heat transfer and a compact primary loop. Compared to helium, a competing high temperature reactor coolant, the difference is even bigger. The fuel salt has over 200 times higher volumetric heat capacity as hot pressurized helium and over 3 times the thermal conductivity. A molten salt loop will use piping of 1/5 the diameter, and pumps 1/20 the power, of those required for high-pressure helium, while staying at atmospheric pressure
- Smaller, low pressure containment. By using liquid salt as the coolant instead of pressurized water, a containment structure only slightly bigger than the reactor vessel can be used. Light water reactors use pressurized water, which flashes to steam and expands a thousandfold in the case of a leak, necessitating a containment building a thousandfold bigger in volume than the reactor vessel. The LFTR containment can not only be smaller in physical size, its containment is also inherently low pressure. There are no sources of stored energy that could cause a rapid pressure rise (such as Hydrogen or steam) in the containment.[unreliable source?] This gives the LFTR a substantial theoretical advantage not only in terms of inherent safety, but also in terms of smaller size, lower materials use, and lower construction cost.
- Air cooling. A high temperature power cycle can be air-cooled at little loss in efficiency, which is critical for use in many regions where water is scarce. No need for large water cooling towers used in conventional steam-powered systems would also decrease power plant construction costs.[unreliable source?]
- From waste to resource. There are suggestions that it might be possible to extract some of the fission products so that they have separate commercial value. However, compared to the produced energy, the value of the fission products is low, and chemical purification is expensive.
- Efficient mining. The extraction process of thorium from the earth's crust is a much safer and efficient mining method than that of uranium. Thorium’s ore, monazite, generally contains higher concentrations of thorium than the percentage of uranium found in its respective ore. This makes thorium a more cost efficient and less environmentally damaging fuel source. Thorium mining is also easier and less dangerous than uranium mining, as the mine is an open pit, which doesn't require ventilation such as the underground uranium mines, where radon levels are potentially harmful.
LFTRs are quite unlike today's operating commercial power reactors. These differences create design difficulties and trade-offs:
- Mothballed technology - Only a few MSRs have actually been built. Those experimental reactors were constructed more than 40 years ago. This leads some technologists[who?] to say that it is difficult to critically assess the concept.
- Startup fuel - Unlike mined uranium, mined thorium does not have a fissile isotope. Thorium reactors breed fissile uranium-233 from thorium, but require a considerable amount of U-233 for initial start up. There is very little of this material available. This raises the problem of how to start the reactors in a reasonable time frame. One option is to produce U-233 in today's solid fuelled reactors, then reprocess it out of the solid waste. A LFTR can also be started by other fissile isotopes, enriched uranium or plutonium from reactors or decommissioned bombs. For enriched uranium startup, high enrichment is needed. Decommissioned uranium bombs have enough enrichment, but not enough is available to start many LFTRs. It is difficult to separate plutonium fluoride from lanthanide fission products. One option for a two-fluid reactor is to operate with plutonium or enriched uranium in the fuel salt, breed U-233 in the blanket, and store it instead of returning it to the core. Instead, add plutonium or enriched uranium to continue the chain reaction, similar to today's solid fuel reactors. When enough U-233 is bred, replace the fuel with new fuel, retaining the U-233 for other startups. A similar option exists for a single-fluid reactor operating as a converter. Such a reactor would not reprocess fuel while operating. Instead the reactor would start on plutonium with thorium as the fertile and add plutonium. The plutonium eventually burns out and U-233 is produced in situ. At the end of the reactor fuel life, the spent fuel salt can be reprocessed to recover the bred U-233 to start up new LFTRs.
- Salts freezing - Fluoride salt mixtures have melting points ranging from 300 to over 600 degrees Celsius. The salts, especially those with beryllium fluoride, are very viscous near their freezing point. This requires careful design and freeze protection in the containment and heat exchangers. Freezing must be prevented in normal operation, during transients, and during extended downtime. The primary loop salt contains the decay heat-generating fission products, which help to maintain the required temperature. For the MSBR, ORNL planned on keeping the entire reactor room (the hot cell) at high temperature. This avoided the need for individual electric heater lines on all piping and provided more even heating of the primary loop components.(p311) One "liquid oven" concept developed for molten salt-cooled, solid-fueled reactors employs a separate buffer salt pool containing the entire primary loop. Because of the high heat capacity and considerable density of the buffer salt, the buffer salt prevents fuel salt freezing and participates in the passive decay heat cooling system, provides radiation shielding and reduces deadweight stresses on primary loop components. This design could also be adopted for LFTRs.
- Beryllium toxicity - The proposed salt mixture FLiBe, contains large amounts of beryllium, which is toxic to humans. The salt in the primary cooling loops must be isolated from workers and the environment to prevent beryllium poisoning. This is routinely done in industry.(pp52–66) Based on this industrial experience, the added cost of beryllium safety is expected to cost only $0.12/MWh.(p61) After start up, the fission process in the primary fuel salt produces highly radioactive fission products with a high gamma and neutron radiation field. Effective containment is therefore a primary requirement. It is possible to operate instead using lithium fluoride-thorium fluoride eutectic without beryllium, as the French LFTR design, the "TMSR", has chosen. This comes at the cost of a somewhat higher melting point, but has the additional advantages of simplicity (avoiding BeF
2 in the reprocessing systems), increased solubility for plutonium-trifluoride, reduced tritium production (beryllium produces lithium-6, which in turn produces tritium) and improved heat transfer (BeF
2 increases the viscosity of the salt mixture). Alternative solvents such as the fluorides of sodium, rubidium and zirconium allow lower melting points at a tradeoff in breeding.
- Loss of delayed neutrons - In order to be predictably controlled, nuclear reactors rely on delayed neutrons. They require additional slowly-evolving neutrons from fission product decay to continue the chain reaction. Because the delayed neutrons evolve slowly, this makes the reactor very controllable. In a LFTR, the presence of fission products in the heat exchanger and piping means a portion of these delayed neutrons are also lost. They do not participate in the core's critical chain reaction, which in turn means the reactor behaves less gently during changes of flow, power, etc. Approximately up to half of the delayed neutrons can be lost. In practice, it means that the heat exchanger must be compact so that the volume outside the core is as small as possible. The more compact (higher power density) the core is, the more important this issue becomes. Having more fuel outside the core in the heat exchangers also means more of the expensive fissile fuel is needed to start the reactor. This makes a fairly compact heat exchanger an important design requirement for a LFTR.
- Waste management - About 83% of the radioactive waste has a half-life in hours or days, with the remaining 17% requiring 300 year storage in geologically stable confinement to reach background levels. Because some of the fission products, in their fluoride form, are highly water-soluble, fluorides are less suited to long-term storage. For example, cesium fluoride has a very high solubility in water. For long term storage, conversion to an insoluble form such as a glass, could be desirable.
- Uncertain decommissioning costs - Cleanup of the Molten-Salt Reactor Experiment was about $130 million, for a small 8 MW(th) unit. Much of the high cost was caused by the unexpected evolution of fluorine and uranium hexafluoride from cold fuel salt in storage that ORNL did not defuel and store correctly, but this has now been taken into consideration in MSR design. In addition, decommissioning costs don't scale strongly with plant size based on previous experience, and costs are incurred at the end of plant life, so a small per kilowatthour fee is sufficient. For example, a GWe reactor plant produces over 300 billion kWh of electricity over a 40-year lifetime, so a $0.001/kWh decommissioning fee delivers $300 million plus interest at the end of the plant lifetime.
- Noble metal buildup - Some radioactive fission products, such as noble metals, deposit on pipes. Novel equipment, such as nickel-wool sponge cartridges, must be developed to filter and trap the noble metals to prevent build up.
- Limited graphite lifetime - Compact designs have a limited lifetime for the graphite moderator and fuel / breeding loop separator. Under the influence of fast neutrons, the graphite first shrinks, then expands indefinitely until it becomes very weak and can crack, creating mechanical problems and causing the graphite to absorb enough fission products to poison the reaction. The 1960 two-fluid design had an estimated graphite replacement period of four years.(p3) Eliminating graphite from sealed piping was a major incentive to switch to a single-fluid design.(p3) Replacing this large central part requires remotely operated equipment. MSR designs have to arrange for this replacement. In a molten salt reactor, virtually all of the fuel and fission products can be piped to a holding tank. Only a fraction of one percent of the fission products end up in the graphite, primarily due to fission products slamming into the graphite. This makes the graphite surface radioactive, and without recycling/removal of at least the surface layer, creates a fairly bulky waste stream. Removing the surface layer and recycling the remainder of the graphite would solve this issue.[original research?] Several techniques exist to recycle or dispose of nuclear moderator graphite. Graphite is inert and immobile at low temperatures, so it can be readily stored or buried if required. At least one design used graphite balls (pebbles) floating in salt, which could be removed and inspected continuously without shutting down the reactor. Reducing power density increases graphite lifetime.(p10) By comparison, solid-fueled reactors typically replace 1/3 of the fuel elements, including all of the highly radioactive fission products therein, every 12 to 24 months. This is routinely done under a protecting and cooling column layer of water.
- Graphite-caused positive reactivity feedback - When graphite heats up, it increases U-233 fission, causing an undesirable positive feedback. The LFTR design must avoid certain combinations of graphite and salt and certain core geometries. If this problem is addressed by employing adequate graphite and thus a well-thermalized spectrum, it is difficult to reach break-even breeding. The alternative of using little or no graphite results in a faster neutron spectrum. This requires a large fissile inventory and radiation damage increases.
- Limited plutonium solubility - Fluorides of plutonium, americium and curium occur as trifluorides, which means they have three fluorine atoms attached (PuF
3). Such trifluorides have a limited solubility in the FLiBe carrier salt. This complicates startup, especially for a compact design that uses a smaller primary salt inventory. Of course, leaving plutonium carrying wastes out of the startup process is an even better solution, making this a non issue. Solubility can be increased by operating with less or no beryllium fluoride (which has no solubility for trifluorides) or by operating at a higher temperature(as with most other liquids, solubility rises with temperature). A thermal spectrum, lower power density core does not have issues with plutonium solubility.
- Proliferation risk from reprocessing - Effective reprocessing implies a proliferation-risk. LFTRs could be used to handle plutonium from other reactors as well. However, as stated above, plutonium is chemically difficult to separate from thorium and plutonium cannot be used in bombs if diluted in large amounts of thorium. In addition, the plutonium produced by the thorium fuel cycle is mostly Pu-238, which produces high levels of spontaneous neutrons and decay heat that make it impossible to construct a fission bomb with this isotope alone, and extremely difficult to construct one containing even very small percentages of it. The heat production rate of 567 W/kg means that a bomb core of this material would continuously produce several kilowatts of heat. The only cooling route is by conduction through the surrounding high explosive layers, which are poor conductors. This creates unmanageably high temperatures that would destroy the assembly. The spontaneous fission rate of 1204 kBq/g is over twice that of Pu-240. Even very small percentages of this isotope would reduce bomb yield drastically by "predetonation" due to neutrons from spontaneous fission starting the chain reaction causing a "fizzle" rather than an explosion. Reprocessing itself involves automated handling in a fully closed and contained hot cell, which complicates diversion. Compared to today's extraction methods such as PUREX, the pyroprocesses are inaccessible and produce impure fissile materials, often with large amounts of fission product contamination. While not a problem for an automated system, it poses severe difficulties for would-be proliferators.
- Proliferation risk from protactinium separation - Compact designs can breed only using rapid separation of protactinium, a proliferation-risk, since this potentially gives access to high purity 233-U. This is difficult as the 233-U from these reactors will be contaminated with 232-U, a high gamma radiation emitter, requiring a protective hot enrichment facility as a possible path to weapons-grade material. Because of this, commercial power reactors may have to be designed without separation. In practice, this means either not breeding, or operating at a lower power density. A two-fluid design might operate with a bigger blanket and keep the high power density core (which has no thorium and therefore no protactinium).
- Proliferation of neptunium-237 - In designs utilizing a fluorinator, Np-237 appears with uranium as gaseous hexafluoride and can be easily separated using solid fluoride pellet absorption beds. No one has produced such a bomb, but Np-237's considerable fast fission cross section and low critical mass imply the possibility. When the Np-237 is kept in the reactor, it transmutes to Pu-238, a high value fuel for space radioisotope thermal generators. A single gram is worth thousands of dollars. Pu-238 is itself an excellent proliferation deterrent. Because of this, Np-237 could be returned to the reactor and transmuted. Vacuum distillation does not separate neptunium. All reactors produce considerable neptunium, which is always present in high (mono)isotopic quality, and is easily extracted chemically. However, this is a byproduct of traditional uranium fission, along with elements like plutonium, americium, and curium. Np-239 or isotopes thereof are not necessarily produced in a LFTR.
- Neutron poisoning and tritium production from lithium-6 - Lithium-6 is a strong neutron poison; using LiF with natural lithium, with its 7.5% lithium-6 content, prevents reactors from starting. The high neutron density in the core rapidly transmutes lithium-6 to tritium, losing neutrons that are required to sustain break-even breeding. Tritium is a radioactive isotope of hydrogen, which is nearly identical, chemically, to ordinary hydrogen. In the MSR the tritium is quite mobile because, in its elemental form, it rapidly diffuses through metals at high temperature. If the lithium is isotopically enriched in lithium-7, and the isotopic separation level is high enough (99.995% lithium-7), the amount of tritium produced is only a few hundred grams per year for a 1 GWe reactor. This much smaller amount of tritium comes mostly from the lithium-7 - tritium reaction and from beryllium, which can produce tritium indirectly by first transmuting to tritium-producing lithium-6. LFTR designs that use a lithium salt, choose the lithium-7 isotope. In the MSRE, lithium-6 was successfully removed from the fuel salt via isotopic enrichment. Since lithium-7 is at least 16% heavier than lithium-6, and is the most common isotope, lithium-6 is comparatively easy and inexpensive to extract. Vacuum distillation of lithium achieves efficiencies of up to 8% per stage and requires only heating in a vacuum chamber. However, about one fission in 90,000 produces helium-6, which quickly decays to lithium-6 and one fission in 12,500 produces an atom of tritium directly (in all reactor types). Practical MSRs operate under a blanket of dry inert gas, usually helium. LFTRs offer a good chance to recover the tritium, since it is not highly diluted in water as in CANDU reactors. Various methods exist to trap tritium, such as hydriding it to titanium, oxidizing it to less mobile (but still volatile) forms such as sodium fluoroborate or molten nitrate salt, or trapping it in the turbine power cycle gas and offgasing it using copper oxide pellets.(p41) ORNL developed a secondary loop coolant system that would chemically trap residual tritium so that it could be removed from the secondary coolant rather than diffusing into the turbine power cycle. ORNL calculated that this would reduce Tritium emissions to acceptable levels.
- Corrosion from tellurium - The reactor makes small amounts of tellurium as a fission product. In the MSRE, this caused small amounts of corrosion at the grain boundaries of the special nickel alloy, Hastelloy-N. Metallurgical studies showed that adding 1 to 2% niobium to the Hastelloy-N alloy improves resistance to corrosion by tellurium.(pp81–87) Maintaining the ratio of UF
3 to less than 60 reduced corrosion by keeping the fuel salt slightly reducing. The MSRE continually contacted the flowing fuel salt with a beryllium metal rod submerged in a cage inside the pump bowl. This caused a fluorine shortage in the salt, reducing tellurium to a less aggressive (elemental) form. This method is also effective in reducing corrosion in general, because the fission process produces more fluorine atoms that would otherwise attack the structural metals.(pp3–4)
- Radiation damage to nickel alloys - The standard Hastelloy N alloy was found to be embrittled by neutron radiation. Neutrons reacted with nickel to form helium. This helium gas concentrated at specific points inside the alloy, where it increased stresses. ORNL addressed this problem by adding 1–2% titanium or niobium to the Hastelloy N. This changed the alloy's internal structure so that the helium would be finely distributed. This relieved the stress and allowed the alloy to withstand considerable neutron flux. However the maximum temperature is limited to about 650 °C. Other alloys also showed promise. The outer vessel wall that contains the salt can have neutronic shielding, such as boron carbide, to effectively protect it from neutron damage.
- Long term fuel salt storage - If the fluoride fuel salts are stored in solid form over many decades, radiation can cause the release of corrosive fluorine gas and uranium hexafluoride. The salts must be defueled and wastes removed before extended shutdowns and stored above 100 degrees Celsius. Fluorides are less suitable for long term storage because some have high water solubility unless vitrified in insoluble borosilicate glass.
- Business model - Today's solid fuelled reactor vendors make long term revenues by fuel fabrication.[dubious ] Without any fuel to fabricate and sell, a LFTR would adopt a different business model.
- Development of the power cycle - Developing a large helium or supercritical carbon dioxide turbine is needed for highest efficiency. These gas cycles offer numerous potential advantages for use with molten salt-fueled or molten salt-cooled reactors. These closed gas cycles face design challenges and engineering upscaling work for a commercial turbine-generator set. A standard supercritical steam turbine could be used at a small penalty in efficiency (the net efficiency of the MSBR was designed to be approximately 44%, using an old 1970s steam turbine). A molten salt to steam generator would still have to be developed. Currently, molten nitrate salt steam generators are used in concentrated solar thermal power plants such as Andasol in Spain. Such a generator could be used for an MSR as a third circulating loop, where it would also trap any tritium that diffuses through the primary and secondary heat exchanger
The Fuji MSR
The FUJI MSR was a design for a 100 to 200 MWe molten-salt-fueled thorium fuel cycle thermal breeder reactor, using technology similar to the Oak Ridge National Laboratory Reactor Experiment. It was being developed by a consortium including members from Japan, the United States, and Russia. As a breeder reactor, it converts thorium into nuclear fuels. Like all molten salt reactors, its core is chemically inert, under low pressures to prevent explosions and toxic releases. An industry group presented updated plans about FUJI MSR in July 2010. The projected cost is 2.85 cents per kilowatt hour.
Chinese thorium MSR project
The People's Republic of China has initiated a research and development project in thorium molten-salt reactor technology. It was formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011. Its ultimate target is to investigate and develop a thorium based molten salt nuclear system in about 20 years. An expected intermediate outcome of the TMSR research program is to build a 2 MW pebble bed fluoride salt cooled research reactor in 2015, and a 2 MW molten salt fueled research reactor in 2017. This would be followed by a 10 MW demonstrator reactor and a 100 MW pilot reactors. The project is spearheaded by Jiang Mianheng, with a start-up budget of $350 million, and has already recruited 140 PhD scientists, working full-time on thorium molten salt reactor research at the Shanghai Institute of Applied Physics. An expansion to 750 staff is planned by 2015.
Kirk Sorensen, former NASA scientist and Chief Nuclear Technologist at Teledyne Brown Engineering, has been a long time promoter of thorium fuel cycle and particularly liquid fluoride thorium reactors. He first researched thorium reactors while working at NASA, while evaluating power plant designs suitable for lunar colonies. Material about this fuel cycle was surprisingly hard to find, so in 2006 Sorensen started "energyfromthorium.com", a document repository, forum, and blog to promote this technology. In 2006, Sorensen coined the liquid fluoride thorium reactor and LFTR nomenclature to describe a subset of molten salt reactor designs based on liquid fluoride-salt fuels with breeding of thorium into uranium-233 in the thermal spectrum. In 2011, Sorensen founded Flibe Energy, a company that initially intends to develop 20-50 MW LFTR small modular reactor designs to power military bases. (It is easier to promote novel military designs than civilian power station designs in today's US nuclear regulatory environment).
Thorium Energy Generation Pty. Limited (TEG)
Thorium Energy Generation Pty. Limited (TEG) was an Australian research and development company dedicated to the worldwide commercial development of LFTR reactors, as well as thorium accelerator-driven systems. As of June 2015, TEG had ceased operations.
Alvin Weinberg Foundation
The Alvin Weinberg Foundation is a British charity founded in 2011, dedicated to act as a communications, debate and lobbying hub to raise awareness about the potential of thorium energy and LFTR. It was formally launched at the House of Lords on 8 September 2011. It is named after American nuclear physicist Alvin M. Weinberg, who pioneered the thorium molten salt reactor research.
Thorcon is a proposed molten salt converter reactor by Martingale, Florida. It features a simplified design with no reprocessing and swappable cans for ease of equipment replacement, in lieu of higher nuclear breeding efficiency.
- Generation IV reactor
- List of small nuclear reactor designs
- Small modular reactor
- Thorium Energy Alliance
- Accelerator-driven sub-critical reactor
- Greene, Sherrel (May 2011). Fluoride Salt-cooled High Temperature Reactors - Technology Status and Development Strategy. ICENES-2011. San Francisco, CA.
- LeBlanc, David (2010). "Molten salt reactors: A new beginning for an old idea" (PDF). Nuclear Engineering and Design (Elsevier) 240 (6): 1644. doi:10.1016/j.nucengdes.2009.12.033.
- Stenger, Victor (12 January 2012). "LFTR: A Long-Term Energy Solution?". Huffington Post.
- Warmflash, David (16 January 2015). "Thorium Power Is the Safer Future of Nuclear Energy". Discover Magazine. Retrieved 22 January 2015.
- UP (29 September 1946). "Atomic Energy 'Secret' Put into Language That Public Can Understand". Pittsburgh Press. Retrieved 18 October 2011.
- UP (21 October 1946). "Third Nuclear Source Bared". The Tuscaloosa News. Retrieved 18 October 2011.
- Hargraves, Robert; Moir, Ralph (July 2010). "Liquid Fluoride Thorium Reactors" (PDF). American Scientist 98 (4): 304–313. doi:10.1511/2010.85.304.
- Synthesis of heavy elements. Gesellschaft für Schwerionenforschung. gsi.de
- The KamLAND Collaboration; Gando, Y.; Ichimura, K.; Ikeda, H.; Inoue, K.; Kibe, Y.; Kishimoto, Y.; Koga, M.; Minekawa, Y. et al. (17 July 2011). "Partial radiogenic heat model for Earth revealed by geoneutrino measurements". Nature Geoscience 4 (9): 647–651. Bibcode:2011NatGe...4..647T. doi:10.1038/ngeo1205.
- "Lab's early submarine reactor program paved the way for modern nuclear power plants". Argonne's Nuclear Science and Technology Legacy. Argonne National Laboratory. 1996.
- Sorensen, Kirk (2 July 2009). "Lessons for the Liquid-Fluoride Thorium Reactor" (PDF). Mountain View, CA: Google.
- Rosenthal, M.; Briggs, R.; Haubenreich, P. "Molten-Salt Reactor Program: Semiannual Progress Report for Period Ending August 31, 1971" (PDF). ORNL-4728. Oak Ridge National Laboratory.
- MacPherson, H. G. (1 August 1985). "The Molten Salt Reactor Adventure". Nuclear Science and Engineering 90: 374–380.
- Weinberg, Alvin (1997). The First Nuclear Era: The Life and Times of a Technological Fixer. Springer. ISBN 978-1-56396-358-2.
- "ORNL: The First 50 Years - Chapter 6: Responding to Social Needs". Retrieved 12 November 2011.
- "Plutonium". World Nuclear Association. March 2012. Retrieved 28 June 2012.
The most common isotope formed in a typical nuclear reactor is the fissile Pu-239 isotope, formed by neutron capture from U-238 (followed by beta decay), and which yields much the same energy as the fission of U-235. Well over half of the plutonium created in the reactor core is consumed in situ and is responsible for about one third of the total heat output of a light water reactor (LWR).(Updated)
- Rosenthal; M. W. et al. (August 1972). "The Development Status of Molten-Salt Breeder Reactors" (PDF). ORNL-4812. Oak Ridge National Laboratory.
- Rosenthal, M. W.; Kasten, P. R.; Briggs, R. B. (1970). "Molten Salt Reactors - History, Status, and Potential" (PDF). Nuclear Applications and Technology 8.
- Section 5.3, WASH 1097 "The Use of Thorium in Nuclear Power Reactors", available as a PDF from Liquid-Halide Reactor Documents Accessed 11/23/09
- Briggs, R. B. (November 1964). "Molten-Salt Reactor Program Semiannual Progress Report For Period Ending July 31, 1964" (PDF). ORNL-3708. Oak Ridge National Laboratory.
- Furukawa; K. A. et al. (2008). "A road map for the realization of global-scale Thorium breeding fuel cycle by single molten-fluoride flow". Energy Conversion and Management 49 (7): 1832. doi:10.1016/j.enconman.2007.09.027.
- Hargraves, Robert; Moir, Ralph (January 2011). "Liquid Fuel Nuclear Reactors". Forum on Physics & Society (American Physical Society) 41 (1): 6–10.
- Robertson, R. C.; Briggs, R. B.; Smith, O. L.; Bettis, E. S. (1970). "Two-Fluid Molten-Salt Breeder Reactor Design Study (Status as of January 1, 1968)". ORNL-4528. Oak Ridge National Laboratory. doi:10.2172/4093364.
- Robertson, R. C. (June 1971). "Conceptual Design Study of a Single-Fluid Molten-Salt Breeder Reactor" (PDF). ORNL-4541. Oak Ridge National Laboratory.
- Engel; J. R. et al. (1980). "Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling" (PDF). ORNL/TM-7207. Oak Ridge National Lab, TN.
- LeBlanc, David (May 2010). "Too Good to Leave on the Shelf". Mechanical Engineering (American Society of Mechanical Engineers).
- Hough, Shane (4 July 2009) Supercritical Rankine Cycle. if.uidaho.edu
- "Oak Ridge National Laboratory: A New Approach to the Design of Steam Generators for Molten Salt Reactor Power Plants" (PDF). Moltensalt.org. Retrieved 24 October 2012.
- Sabharwall, Piyush; Kim, Eung S.; McKellar, Michael; Anderson, Nolan (April 2011). Process Heat Exchanger Options for Fluoride Salt High Temperature Reactor (PDF) (Report) (INL/EXT-11-21584). Idaho National Laboratory.
- ""Flower power" has been inaugurated in Israel" (News). Enel Green Power. 10 July 2009.
- "Nuclear Desalination". World Nuclear Association. March 2012.
- Hill, Robert; Hodge, C. G.; Gibbs, T. (May 2012). The potential of the molten salt reactor for warship propulsion (PDF). INEC 2012. Edinburgh, UK.
- "Pyrochemical Separations in Nuclear Applications: A Status Report" (PDF). Retrieved 24 October 2012.
- Forsberg, Charles W. (2006). "Molten-Salt-Reactor Technology Gaps" (PDF). Proceedings of the 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (American Nuclear Society). Retrieved 7 April 2012.
- "LIFE Materials: Molten-Salt Fuels Volume 8" (PDF). E-reports-ext.11nl.gov. Retrieved 24 October 2012.
- "Low-Pressure Distillation of Molten Fluoride Mixtures: Nonradioactive Tests for the MSRE Distillation Experiment;1971, ORNL-4434" (PDF). Retrieved 24 October 2012.
- "Design Studies of 1000-Mw(e) Molten-Salt Breeder Reactors; 1966, ORNL-3996" (PDF). Retrieved 24 October 2012.
- "Engineering Tests of the Metal Transfer Process for Extraction of Rare-Earth Fission Products from a Molten-Salt Breeder Reactor Fuel Salt; 1976, ORNL-5176" (PDF). Retrieved 24 October 2012.
- Conocar, Olivier; Douyere, Nicolas; Glatz, Jean-Paul; Lacquement, Jérôme; Malmbeck, Rikard and Serp, Jérôme (2006). "Promising pyrochemical actinide/lanthanide separation processes using aluminium". Nuclear Science and Engineering 153 (3): 253–261.
- "Molten Salt Reactors: A New Beginning for an Old Idea" (PDF). Retrieved 24 October 2012.
- "Potential of Thorium Fueled Molten Salt Reactors" (PDF). Retrieved 24 October 2012.
- "6th Int'l Summer Student School on Nuclear Physics Methods and Accelerators in Biology and Medicine (July 2011, JINR Dubna, Russia)" (PDF). Uc2.jinr.ru. Retrieved 24 October 2012.
- Cooper, N.; Minakata, D.; Begovic, M.; Crittenden, J. (2011). "Should We Consider Using Liquid Fluoride Thorium Reactors for Power Generation?". Environmental Science & Technology 45 (15): 6237. doi:10.1021/es2021318.
- Mathieu, L.; Heuer, D.; Brissot, R.; Garzenne, C.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Loiseaux, J.-M.; Méplan, O. et al. (2006). "The Thorium molten salt reactor: Moving on from the MSBR" (PDF). Progress in Nuclear Energy 48 (7): 664. arXiv:nucl-ex/0506004. doi:10.1016/j.pnucene.2006.07.005.
- "Engineering Database of Liquid Salt Thermophysical and Thermochemical Properties" (PDF). Inl.gov. Retrieved 24 October 2012.
- "Chapter 13: Construction Materials for Molten-Salt Reactors" (PDF). Moltensalt.org. Retrieved 24 October 2012.
- "Thermal- and Fast Spectrum Molten Salt Reactors for Actinide Burning and Fuel Production" (PDF). Retrieved 24 October 2012.
- Devanney, Jack. "Simple Molten Salt Reactors: a time for courageous impatience" (PDF). C4tx.org. Retrieved 24 October 2012.
- Moir, R. W. (2008). "Recommendations for a restart of molten salt reactor development" (PDF). Energy Convers. Management 49 (7): 1849–1858. doi:10.1016/j.enconman.2007.07.047.
- Leblanc, D. (2010). "Molten salt reactors: A new beginning for an old idea". Nuclear Engineering and Design 240 (6): 1644. doi:10.1016/j.nucengdes.2009.12.033.
- "The Influence of Xenon-135 on Reactor Operation" (PDF). C-n-t-a.com. Retrieved 24 October 2012.
- "Assessment of Candidate Molten Salt Coolants for the Advanced High-Temperature Reactor (AHTR)- ORNL-TM-2006-12" (PDF). Retrieved 24 October 2012.
- "A Modular Radiant Heat-Initiated Passive Decay-Heat-Removal System for Salt-Cooled Reactors" (PDF). Ornl.gov. Retrieved 24 October 2012.
- Thorium Fuel Cycle, AEC Symposium Series, 12, USAEC, Feb. 1968
- "Using LTFR to Minimize Actinide Wastes" (PDF). Thoriumenergyaslliance.com. Retrieved 24 October 2012.
- Hargraves, Robert and Moir, Ralph (27 July 2011). "Liquid Fuel Nuclear Reactors". Aps.org. Retrieved 3 August 2012.
- Engel, J. R.; Grimes, W. R.; Bauman, H. F.; McCoy, H. E.; Dearing, J. F.; Rhoades, W. A. (1980) Conceptual Design Characteristics of Denatured Molten-Salt Breeder Reactor with Once-through Fueling; ORNL/TM-7207; Oak Ridge National Laboratory: Oak Ridge, TN.
- "Transmutation Dynamics: Impacts of Multi-Recycling on Fuel Cycle Performances" (PDF). Inl.gov. September 2009. Retrieved 24 October 2012.
- "Long term Radiotoxicity" (PDF). Ictp.it. Retrieved 24 October 2012.
- Sylvain, David et al. (March–April 2007). "Revisiting the Thorium-Uranium nuclear fuel cycle" (PDF). Europhysics News 38 (2): 24–27. Bibcode:2007ENews..38...24D. doi:10.1051/EPN:2007007.
- "Image based on" (PDF). Thoriumenergyalliance.com. Retrieved 24 October 2012.
- Evans-Pritchard, Ambrose (29 August 2010) Obama could kill fossil fuels overnight with a nuclear dash for thorium. Telegraph. Retrieved on 24 April 2013.
- Hargraves, R., & Moir, R. (2010). Liquid fluoride thorium reactors. American Scientist, 98(4), 304-313.
- "Oak Ridge National Laboratory: Abstract" (PDF). Energyfromthorium. Retrieved 24 October 2012.
- "Denatured Molten Salt Reactors" (PDF). Coal2nuclear.com. Retrieved 24 October 2012.
- "Estimated Cost of Adding a Third Salt-Circulating System for Controlling Tritium Migration in the 1000-Mw(e) MSBR [Disc 5]" (PDF). Retrieved 24 October 2012.
- Bonometti, J. "LFTR Liquid Fluoride Thorium Reactor-What fusion wanted to be!" Presentation available in www.energyfromthorium.com (2011)
- "Critical issues of nuclear energy systems employing molten salt fluorides" (PDF). Retrieved 24 October 2012.
- "LIFE Materials: Molten-Salt Fuels Volume 8" (PDF). E-reports-ext.llnl.gov. Retrieved 24 October 2012.
- Peterson, Per F.; Zhao, H. and Fukuda, G. (5 December 2003). "Comparison of Molten Salt and High-Pressure Helium for the NGNP Intermediate Heat Transfer Fluid" (PDF). U.C. Berkeley Report UCBTH-03-004.
- Forsberg, Charles W.; Peterson, Per F; Zhao, Haihua (2007). "High-temperature liquid-fluoride-salt closed-brayton-cycle solar power towers" (PDF). Journal of solar energy engineering 129 (2): 141–146. doi:10.1115/1.2710245.
- "Products". Flibe Energy. Retrieved 24 October 2012.
- Bush, R. P. (1991). "Recovery of Platinum Group Metals from High Level Radioactive Waste" (PDF). Platinum Metals Review 35 (4): 202–208.
- "Thorium fuel cycle — Potential benefits and challenges" (PDF). International Atomic Energy Agency. Retrieved 27 October 2014.
- Peterson, Per F. and Zhao, Haihua (29 December 2005). "Preliminary Design Description for a First-Generation Liquid-Salt VHTR with Metallic Vessel Internals (AHTR-MI)" (PDF). U.C. Berkeley Report UCBTH-05-005.
- Fei, Ting et al. (16 May 2008). "A MODULAR PEBBLE-BED ADVANCE D HIGH TEMPERATURE REACTOR" (PDF). U.C. Berkeley Report UCBTH-08-001. Retrieved 24 October 2012.
- "The Thorium Molten Salt Reactor: Launching The Thorium Cycle While Closing The Current Fuel Cycle" (PDF). Retrieved 24 October 2012.
- "The Aircraft Reactor Experiment-Physics" (PDF). Moltensalt.org. Retrieved 24 October 2012.
- "Fluorine Production and Recombination in Frozen MSR Salts after Reactor Operation [Disc 5]" (PDF). Retrieved 24 October 2012.
- "Costs of decommissioning nuclear power plants" (PDF). Iaea.org. Retrieved 24 October 2012.
- "Oak Ridge National Laboratory: Graphite Behaviour and Its Effects on MSBR Performance" (PDF). Moltensalt.org. Retrieved 24 October 2012.
- "IAEA-TECDOC-1521" (PDF). Retrieved 24 October 2012.
- "Semiannual Progress Report for Period Ending February 28, 1970" (PDF). ORNL-4548: Molten-Salt Reactor Program. p. 57. Retrieved 6 June 2015.
- Rodriguez-Vieitez, E.; Lowenthal, M. D.; Greenspan, E.; Ahn, J. (7 October 2002). Optimization of a Molten-Salt Transmuting Reactor (PDF). PHYSOR 2002. Seoul, Korea.
- "Nuclear Weapons Archive - Useful Tables". Retrieved 2013-08-31.
- "Neptunium 237 and Americium: World Inventories and Proliferation Concerns" (PDF). Isis-online.org. Retrieved 24 October 2012.
- "Assessment of Plutonium-238 (Pu-238) Production Alternatives" (PDF). Retrieved 24 October 2012.
- "Distribution and Behavior of Tritium in the Coolant-Salt Technology Facility [Disc 6]" (PDF). Retrieved 24 October 2012.
- Manely; W. D. et al. (1960). "Metallurgical Problems in Molten Fluoride Systems". Progress in Nuclear Energy 2: 164–179.
- "Information Bridge: DOE Scientific and Technical Information - Sponsored by OSTI" (PDF). Osti.gov. 31 August 2012. Retrieved 24 October 2012.
- "Information Bridge: DOE Scientific and Technical Information - Sponsored by OSTI" (PDF). Osti.gov. 31 August 2012. Retrieved 24 October 2012.
- Moir; R. W. et al. (2002). "Deep-Burn Molten-Salt Reactors" (Application under Solicitation). LAB NE 2002-1. Department of Energy, Nuclear Energy Research Initiative.
- "Status of materials development for molten salt reactors" (PDF). Retrieved 24 October 2012.
-  (52 MB) Intergranular Cracking of INOR-8 in the MSRE,
- "Potential of Thorium Molten Salt Reactors: Detailed Calculations and Concept Evolutions in View of a Large Nuclear Energy Production" (PDF). Hal.archives-ouvertes.fr. Retrieved 24 October 2012.
- National Research Council (U.S.). Committee on Remediation of Buried and Tank Wastes. Molten Salt Panel (1997). Evaluation of the U.S. Department of Energy's alternatives for the removal and disposition of molten salt reactor experiment fluoride salts. National Academies Press. p. 15. ISBN 0-309-05684-5.
- Forsberg, C.; Beahm, E.; Rudolph, J. (2 December 1996). Direct Conversion of Halogen-Containing Wastes to Borosilicate Glass (PDF). Symposium II Scientific Basis for Nuclear Waste Management XX 465. Boston, Massachusetts: Materials Research Society. pp. 131–137.
- Zhao, H. and Peterson, Per F. (25 February 2004). "A Reference 2400 MW(t) Power Conversion System Point Design for Molten-Salt-Cooled Fission and Fusion Energy Systems" (PDF). U.C. Berkeley Report UCBTH-03-002.
- "Conceptual Design study of a Single Fluid Molten Salt Breeder Reactor" (PDF). Energyfromthorium.com. Retrieved 24 October 2012.
- "Heat Transfer Salt for High Temperature Steam Generation [Disc 5]" (PDF). Retrieved 24 October 2012.
- Fuji MSR pp. 821-856, Jan 2007 20MB PDF
- "Small nuclear power reactors". Eoearth.org. 6 January 2010. Retrieved 24 October 2012.
- "IThEO Presents International Thorium Energy & Molten-Salt Technology Inc." (news). International Thorium Energy Organisation. 20 July 2010.
- "Chapter X. MSR-FUJI General Information, Technical Features, and Operating Characteristics" (PDF).
- Martin, Richard (2011-02-01). "China Takes Lead in Race for Clean Nuclear Power". Wired Science.
- "未来核电站 安全"不挑食"". Whb.news365.com.cn. 26 January 2011. Retrieved 24 October 2012.
- Clark, Duncan (16 February 2011). "China enters race to develop nuclear energy from Thorium". The Guardian (London).
- "Kun Chen from Chinese Academy of Sciences on China Thorium Molten Salt Reactor TMSR Program". YouTube. 10 August 2012. Retrieved 24 October 2012.
- Halper, Mark (30 October 2012). "Completion date slips for China.s thorium molten salt reactor". Weinberg Foundation. Retrieved 17 April 2013.
- Evans-Pritchard, Ambrose (6 January 2013). "China blazes trail for 'clean' nuclear power from thorium". The Daily Telegraph.
- "Flibe Energy". Flibe Energy. Retrieved 24 October 2012.
- "New Huntsville company to build Thorium-based nuclear reactors". Huntsvillenewswire.com. 27 September 2011. Retrieved 24 October 2012.
- Halper, Mark. "Home". The Alvin Weinberg Foundation. Retrieved 24 October 2012.
- Martin, David. "Welcome to the Alvin Weinberg Foundation". The Alvin Weinberg Foundation. Retrieved 2014-05-22.
- Clark, Duncan (9 September 2011). "Thorium advocates launch pressure group". The Guardian (London).
- "The Weinberg Foundation - London: Weinberg Foundation to heat up campaign for safe, green,...". Mynewsdesk. 8 September 2011. Retrieved 24 October 2012.
- "New NGO to fuel interest in safe thorium nuclear reactors". BusinessGreen. 8 September 2011. Retrieved 24 October 2012.
- Hargraves, Robert (2009). Aim High!: Thorium energy cheaper than from coal solves more than just global warming (PDF). BookSurge Publishing. ISBN 1-4392-2538-9.
- Martin, Richard (2012). SuperFuel: Thorium, the Green Energy Source for the Future. Palgrave Macmillan. ISBN 0-230-11647-7.
- Cooper, N.; Minakata, D.; Begovic, M.; Crittenden, J. (2011). "Should We Consider Using Liquid Fluoride Thorium Reactors for Power Generation?". Environmental Science & Technology 45 (15): 6237. doi:10.1021/es2021318.
The Restoration of the Earth, Theodore B. Taylor and Charles C. Humpstone, 166 pages, Harper & Row (1973) isbn: 978-0060142315
Sustainable energy - Without the Hot Air, David J.C. MacKay, 384 pages, UIT Cambridge (2009) isbn: 978-0954452933
2081: A Hopeful Vision of the Human Future, Gerard K. O'Neill, 284 pages, Simon & Schuster (1981) isbn: 978-0671242572
The Second Nuclear Era: A New Start for Nuclear Power, Alvin M. Weinberg et al., 460 pages, Praeger Publishers (1985) ISBN 978-0275901837
Thorium Fuel Cycle - Potential Benefits and Challenges, IAEA, 105 pages (2005) ISBN 978-9201034052
The Nuclear Imperative: A Critical Look at the Approaching Energy Crisis (More Physics for Presidents), Jeff Eerkens, 212 pages, Springer (2010) ISBN 978-9048186662
- Uranium Is So Last Century - Enter Thorium, the New Green Nuke Wired Magazine article
- Is Thorium the Biggest Energy Breakthrough Since Fire? Possibly. Forbes article
- International Thorium Energy Organisation – IThEO.org
- Energy from Thorium - Web site about LFTR with a blog, ORNL molten salt reactor program reports, research papers repository and discussion forum
- Thorium Energy Alliance - advocacy and educational organisation dedicated to thorium energy
- International Thorium Molten-Salt Forum
- ThoriumMSR - comprehensive website and blog about thorium molten salt reactor technology
- The Weinberg Foundation website
- TEDxYYC - Kirk Sorensen - Thorium Sorensen's presentation about LFTR at TEDxYYC 2011
- Liquid Fluoride Thorium Reactor: What Fusion Wanted To Be Google TechTalk by Dr. Joe Bonometti NASA / Naval Postgraduate School
- The Thorium Molten-Salt Reactor: Why Didn't This Happen Google TechTalk by Kirk Sorensen examining the history of thorium molten salt reactor development at Oak Ridge, political climate and reasons responsible for the cancellation of the program
- Kirk Sorensen - A Global Alternative @ TEAC4 Kirk Sorensen's presentation at Thorium Energy Alliance Conference #4 in Chicago.