The liquid fluoride thorium reactor (LFTR; often pronounced lifter) is a type of molten salt reactor. LFTRs use the thorium fuel cycle with a fluoride-based molten (liquid) salt for fuel. In a typical design, 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. Molten-salt-fueled reactors (MSRs) supply the nuclear fuel mixed into a molten salt. They should not be confused with designs that use a molten salt for cooling only (fluoride high-temperature reactors) and still have 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 neutron spectrum. The LFTR concept was first investigated at the Oak Ridge National Laboratory Molten-Salt Reactor Experiment in the 1960s, though the MSRE did not use thorium. The LFTR has recently been the subject of a renewed interest worldwide. Japan, China, the UK and private US, Czech, Canadian and Australian companies have expressed the intent to develop, and commercialize the technology. LFTRs differ from other power reactors in almost every aspect: they use thorium that breeds into uranium-233, instead of using isotope-enhanced uranium-235 (enriched uranium); they are refueled by pumping without shutdown. Their liquid salt coolant allows higher operating temperature and much lower pressure in the primary cooling loop. These distinctive characteristics give rise to many potential advantages, as well as design challenges.
Background
By 1946, eight years after the discovery of nuclear fission, three fissile isotopes had been publicly identified for use as nuclear fuel: 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 (ARE) in 1954 and Molten-Salt Reactor Experiment (MSRE) 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. 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.
Breeding basics
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). In order to become fissile these nuclides must first absorb a neutron that's been produced in the process of fission, to become Th-233 and U-239 respectively. After two sequential beta decays, they transmute into fissile isotopes U-233 and Pu-239 respectively. This process is called breeding. All reactors breed some fuel this way, but today's solid-fueled thermal reactors do not 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 two new neutrons from fissioning the bred plutonium. Since one neutron is required to sustain the fission reaction, this leaves a budget of less than one 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, fissile 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, but are more sensitive to fission products left in the core. 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 (molten salt breeder reactor) 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. 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 than 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 neutrons and slowly converts its thorium 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 injecting additional fluorine to create uranium hexafluoride, a gas which can be captured as it comes out of solution. Once reduced again to uranium tetrafluoride, a solid, it can be mixed into the core salt medium 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:
Simpler 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, other than that which has been bred there. 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. 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 weakness of the two-fluid design is the necessity of periodically replacing the core-blanket barrier due to fast neutron damage. ORNL chose graphite for its barrier material 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 it, causing an increase in porosity and a deterioration in physical properties. Graphite pipes would change length, and may crack and leak. Another weakness of the two-fluid design is its complex plumbing. ORNL thought a complex interleaving of core and blanket tubes was necessary to achieve a high power level with acceptably low power density. ORNL chose not to pursue the two-fluid design, and no examples of the two-fluid reactor were ever constructed. However, more recent research has questioned the need for ORNL's complex interleaving graphite tubing, suggesting a simple elongated tube-in-shell reactor that would allow high power output without complex tubing, accommodate thermal expansion, and permit tube replacement. Additionally, graphite can be replaced with high molybdenum alloys, which are used in fusion experiments and have greater tolerance to neutron damage.
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 one and a half or two fluid LFTR is whether a more complicated reprocessing or a more demanding structural barrier will be easier to solve.
Power generation An LFTR with a high operating temperature of 700 degrees Celsius can operate at a thermal efficiency in converting heat to electricity of 45%. This is higher than today's light water reactors (LWRs) that are at 32–36% thermal to electrical efficiency. In addition to electricity generation, concentrated thermal energy from the high-temperature LFTR can be used as high-grade industrial process heat for many uses, such as ammonia production with the Haber process or thermal hydrogen production by water splitting, eliminating the efficiency loss of first converting to electricity.
Rankine cycle
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%, and had done considerable design work on developing molten fluoride salt – steam generators.
Brayton cycle
The Brayton cycle generator has a much smaller footprint than the Rankine cycle, lower cost and higher thermal efficiency, but requires higher operating temperatures. It is therefore particularly suitable for use with a LFTR. The working gas can be helium, nitrogen, or carbon dioxide. 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. The high-pressure working gas is expanded in a turbine to produce power. 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.
Removal of fission products
The LFTR needs a mechanism to remove the fission products from the fuel. Fission products left in the reactor absorb neutrons and thus reduce neutron economy. This is especially important in the thorium fuel cycle with few spare neutrons and a thermal neutron spectrum, where absorption is strong. The minimum requirement is to recover the valuable fissile material from used fuel. Removal of fission products is similar to reprocessing of solid fuel elements; by chemical or physical means, the valuable fissile fuel is separated from the waste fission products. Ideally the fertile fuel (thorium or U-238) and other fuel components (e.g. carrier salt or fuel cladding in solid fuels) can also be reused for new fuel. However, for economic reasons they may also end up in the waste. 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 concentrations of some of the rare earth elements must be especially kept low, as they have a large absorption cross section. Some other elements with a small cross section like Cs or Zr may accumulate over years of operation before they are removed. As the fuel of a LFTR is a molten salt mixture, it is attractive to use pyroprocessing, high temperature methods working directly with the hot molten salt. Pyroprocessing does not use radiation sensitive solvents and is not easily disturbed by decay heat. It can be used on 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. One potential advantage of a liquid fuel is that it not only facilitates separating fission-products from the fuel, but also isolating individual fission products from one another, which is lucrative for isotopes that are scarce and in high-demand for various industrial (radiation sources for testing welds via radiography), agricultural (sterilizing produce via irradiation), and medical uses (molybdenum-99 which decays into technetium-99m, which is the most commonly used medical radioisotope).
Details by element group The more noble metals (Pd, Ru, Ag, Mo, Nb, Sb, Tc) do not form fluorides in the normal salt, but instead fine colloidal metallic particles. They can plate out on metal surfaces like the heat exchanger, or preferably on high surface area filters which are easier to replace. Still, there is some uncertainty where they end up, as the MSRE only provided a relatively short operating experience and independent laboratory experiments are difficult. Gases like Xe and Kr come out easily with a sparge of helium. In addition, some of the "noble" metals are removed as an aerosol. The quick removal of Xe-135 is particularly important, as it is a very strong neutron poison and makes reactor control more difficult if unremoved; this also improves neutron economy. The gas (mainly He, Xe and Kr) is held for about 2 days until almost all 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, which needs storage (e.g. in compressed form) for an extended time (several decades) to wait for the decay of Kr-85. 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 some other fission products (e.g. iodine, molybdenum and tellurium). 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.
Optional protactinium-233 separations 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.
Advantages Thorium-fueled molten salt reactors offer many potential advantages compared to conventional solid uranium fueled light water reactors:
Safety 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. 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, 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 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 do not 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. Proliferation resistance. In 2016, Nobel Laureate physicist Dr Carlo Rubbia, former director general of CERN, claimed a primary reason for the United States cutting thorium reactor research in the 1970s is what makes it so attractive today: thorium is difficult to turn into a nuclear weapon.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 does not make much spare fuel. It produces at most 9% more fuel than it burns each year, and it is 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 perfectly working reprocessing LFTR may consume up to 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, this inventory would cost less than $4 million, a modest cost for a multibillion-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 at $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 (83%) have relatively short half-lives 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 about 10,000 times less than that of one through uranium fuel. 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 c