Liquid fluoride thorium reactor

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, FHRs) 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 is turned into uranium, instead of using uranium directly; 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:
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. 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 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, 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 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:
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, a valuable radiolabel dye for marking cancerous cells in medical scans).
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:
Disadvantages.
LFTRs are quite unlike today's operating commercial power reactors. These differences create design difficulties and trade-offs:
Recent developments.
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. An industry group presented updated plans about FUJI MSR in July 2010. They projected a cost of 2.85 cents per kilowatt hour.
The IThEMS consortium planned to first build a much smaller MiniFUJI 10 MWe reactor of the same design once it had secured an additional $300 million in funding, but IThEMS closed in 2011 after it was unable to secure adequate funding. A new company, Thorium Tech Solution (TTS), was founded in 2011 by Kazuo Furukawa, the chief scientist from IThEMS, and Masaaki Furukawa. TTS acquired the FUJI design and some related patents.
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 of staffing has increased to 700 as of 2015. As of 2016, their plan is for a 10MW pilot LFTR is expected to be made operational in 2025, with a 100MW version set to follow in 2035.
Flibe Energy.
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). An independent technology assessment coordinated with EPRI and Southern Company represents the most detailed information so far publicly available about Flibe Energy's proposed LFTR design.
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 was a British charity founded in 2011, dedicated to raising awareness about the potential of thorium energy and LFTR. It was formally launched at the House of Lords on 8 September 2011. It is named after American nuclear physicist Alvin M. Weinberg, who pioneered the thorium molten salt reactor research.
Thorcon.
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.
Nuclear Research and Consultancy Group.
On 5 September 2017, The Dutch Nuclear Research and Consultancy Group announced that research on the irradiation of molten thorium fluoride salts inside the Petten high-flux reactor was underway.