A molten salt reactor (MSR) is a type of nuclear fission reactor in which the primary coolant, or even the fuel itself is a molten salt mixture. MSRs run at higher temperatures than water-cooled reactors for higher thermodynamic efficiency, while staying at low vapor pressure.
Operating at near atmospheric pressures reduces the mechanical stress endured by the system, thus simplifying aspects of reactor design and improving safety. Molten Salt Reactors might be possible to construct and operate cheaper than coal power plants.[1]
The nuclear fuel may be solid or dissolved in the coolant itself. In many designs the nuclear fuel is dissolved in the molten fluoride salt coolant as uranium tetrafluoride (UF4). The fluid becomes critical in a graphite core which serves as the moderator. Solid fuel designs rely on ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high temperature cooling. The salts are much more efficient than water at removing heat from the core, reducing the need for pumping, piping, and reducing the size of the core.
The early Aircraft Reactor Experiment (1954) was primarily motivated by the small size that the design could provide, while the Molten-Salt Reactor Experiment (1965–1969) was a prototype for a thorium fuel cycle breeder reactor nuclear power plant. One of the Generation IV reactor designs is a molten salt-cooled, solid-fuel reactor; the initial reference design is 1000 MWe [2] with a deployment target date of 2025.
Another advantage of a small core is that it has fewer materials to absorb neutrons. In a reactor employing thorium fuel, the improved neutron economy makes more neutrons available to breed thorium-232 into uranium-233. Thus, the compact core makes the molten salt design particularly suitable for the thorium fuel cycle.
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Extensive research into molten salt reactors started with the U.S. Aircraft Reactor Experiment (ARE) in support of the U.S. Aircraft Nuclear Propulsion program. The ARE was a 2.5 MWth nuclear reactor experiment designed to attain a high power density for use as an engine in a nuclear powered bomber. The project included several reactor experiments including high temperature reactor and engine tests collectively called the Heat Transfer Reactor Experiments: HTRE-1, HTRE-2, and HTRE-3 at the National Reactor Test Station (now Idaho National Laboratory) as well as an experimental high-temperature molten salt reactor at Oak Ridge National Laboratory - the ARE. The ARE used molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, was moderated by beryllium oxide (BeO), used liquid sodium as a secondary coolant, and had a peak temperature of 860 °C. It operated for 100 MW-hours over nine days in 1954. This experiment used Inconel 600 alloy for the metal structure and piping.[3]
Oak Ridge National Laboratory (ORNL) took the lead in researching the MSR through 1960s, and much of their work culminated with the Molten-Salt Reactor Experiment (MSRE). The MSRE was a 7.4 MWth test reactor simulating the neutronic "kernel" of a type of inherently safe epithermal thorium molten salt breeder reactor called the Liquid fluoride thorium reactor. It tested molten salt fuels of uranium and plutonium. The tested 233UF4 fluid fuel has a unique decay path that minimizes waste. The 650 °C temperature of the reactor could power high-efficiency heat engines such as closed-cycle gas turbines. The large, expensive breeding blanket of thorium salt was omitted in favor of neutron measurements.
The MSRE was located at ORNL. Its piping, core vat and structural components were made from Hastelloy-N and its moderator was pyrolytic graphite. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-30-5-0.1), the graphite core moderated it, and its secondary coolant was FLiBe (2LiF-BeF2). It reached temperatures as high as 650 °C and operated for the equivalent of about 1.5 years of full power operation.
The culmination of the Oak Ridge National Laboratory research during the 1970–76 timeframe resulted in an Molten Salt Breeder Reactor (MSBR) design which would use LiF-BeF2-ThF4-UF4 (72-16-12-0.4) as fuel, was to be moderated by graphite with a 4 year replacement schedule, use NaF-NaBF4 as the secondary coolant, and have a peak operating temperature of 705 °C.[4]
The molten salt reactor offers many potential advantages:[4]
Research is currently picking up again for reactors that utilize molten salts for coolant. Both the traditional molten salt reactor and the Very High Temperature Reactor (VHTR) have been selected as potential designs to be studied under the Generation Four Initiative (GEN-IV). A version of the VHTR currently being studied is the Liquid Salt Very High Temperature Reactor (LS-VHTR), also commonly called the Advanced High Temperature Reactor (AHTR). It is essentially a standard VHTR design that uses liquid salt as a coolant in the primary loop, rather than a single helium loop. It relies on "TRISO" fuel dispersed in graphite. Early AHTR research focused on graphite would be in the form of graphite rods that would be inserted in hexagonal moderating graphite blocks, but current studies focus primarily on pebble-type fuel. The LS-VHTR has many attractive features, including: the ability to work at very high temperatures (the boiling point of most molten salts being considered are >1400 °C), low pressure cooling that can be used to more easily match hydrogen production facility conditions (most thermo chemical cycles require temperatures in excess of 750 °C), better electric conversion efficiency than a helium cooled VHTR operating at similar conditions, passive safety systems, and better retention of fission products in the event of an accident. This concept is now referred to as Fluoride Salt Cooled High Temperature Reactor (FHR).[5]
Reactors containing molten thorium salt, called Liquid Fluoride Thorium Reactor (LFTR), would tap the abundant energy source of the thorium fuel cycle. Private companies from Japan, Russia, and the United States, and the Chinese government, have expressed interest in developing this technology.[6][7][8]
Advocates estimate that five hundred metric tons of thorium could supply all U.S. energy needs for one year.[9] The U.S. Geological Survey estimates that the largest known U.S. thorium deposit, the Lemhi Pass district on the Montana-Idaho border, contains thorium reserves of 64,000 metric tons of thorium.[10]
The FUJI MSR is a 100 to 200 MWe LFTR, using technology similar to the Oak Ridge National Laboratory Reactor. It is being developed by a consortium including members from Japan, the U.S. and Russia. It would likely take 20 years to develop a full size reactor[11] but the project seems to lack funding.[12]
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.[13][14][15]
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. In 2011, Sorensen founded Flibe Energy, a company aimed at developing 20-50 MW LFTR reactor designs to power military bases. (it is easier to approve novel military designs than civilian power station designs in today's US nuclear regulatory environment).[16][17][18][19]
The Weinberg Foundation is a British non-profit organisation founded in 2011 dedicated to promotion and development of a liquid fluoride thorium reactor.[20][21][22]
Molten-salt-fueled reactors are quite different from molten-salt-cooled solid-fuel reactors, called simply "Molten Salt Reactor System" in the Generation IV proposal, also called MSCR, which is also the acronym for the Molten Salt Converter Reactor design. It cannot reprocess fuel easily and has fuel rods that need to be fabricated and validated, delaying deployment by up to twenty years from project inception. However, since it uses fabricated fuel, reactor manufacturers can still profit by selling fuel assemblies.
The MSCR retains the safety and cost advantages of a low-pressure, high-temperature coolant, also shared by liquid metal cooled reactors. Notably, there is no steam in the core to cause an explosion, and no large, expensive steel pressure vessel. Since it can operate at high temperatures, the conversion of the heat to electricity can also use an efficient, lightweight Brayton cycle gas turbine.
Much of the current research on MSCRs is focused on small compact heat exchangers. By using smaller heat exchangers, less molten salt needs to be used and therefore significant cost savings could be achieved.[24]
Molten salts can be highly corrosive, more so as temperatures rise. For the primary cooling loop of the MSR, a material is needed that can withstand corrosion at high temperatures and intense radiation. Experiments show that Hastelloy-N and similar alloys are quite suited to the tasks at operating temperatures up to about 700 °C. However, long-term experience with a production scale reactor has yet to be gained. Higher operating temperatures would be desirable, but at 850 °C thermo chemical production of hydrogen becomes possible, which creates serious engineering difficulties. Materials for this temperature range have not been validated, though carbon composites, molybdenum alloys (e.g. TZM), carbides, and refractory metal based or ODS alloys might be feasible.
The salt mixtures are chosen to make the reactor safer and more practical. Fluorides are favored because fluorine does not need expensive isotope separation (as chlorine does). It does not easily become radioactive under neutron bombardment. It also absorbs fewer neutrons and slows ("moderates") neutrons better. Low-valence fluorides boil at high temperatures, though many pentafluorides and hexafluorides boil at low temperatures. They also must be very hot before they break down into their simpler components, such molten salts are "chemically stable" when maintained well below their boiling points.
Reactor salts are also eutectic mixtures to reduce their melting point. This makes a heat engine more efficient, because more heat can be removed from the salt before reheating it in the reactor.
Some salts are so useful that isotope separation is worthwhile. Chlorides permit fast breeder reactors to be constructed using molten salts. Not nearly as much work has been done on reactor designs using them. Chlorine must be purified to chlorine-37 to reduce production of sulfur tetrafluoride when the radioactive chlorine decays to sulfur. Also, any lithium in a salt mixture must be purified lithium-7 to reduce tritium production (the tritium forms hydrogen fluoride).
Due to the high "redox window" of fused fluoride salts, the chemical potential of the fused salt system can be changed. Fluorine-Lithium-Beryllium ("FLiBe") can be used with beryllium additions to lower the electrochemical potential and almost eliminate corrosion. However, since beryllium is extremely toxic, special precautions must be engineered into the design to prevent its release into the environment. Many other salts can cause plumbing corrosion, especially if the reactor is hot enough to make highly reactive hydrogen.
To date, most research has focused on FLiBe, because Lithium and Beryllium are reasonably effective moderators, and form a eutectic salt mixture with a lower melting point than each of the constituent salts. Beryllium also performs neutron doubling, improving the neutron economy. This process occurs when the Beryllium nucleus re-emits two neutrons after absorbing a single neutron. For the fuel carrying salts, generally 1% or 2% (by mole) of UF4 is added. thorium and plutonium fluorides have also been used. The MSFR is the only system that has run a single reactor, the MSRE, from all three known nuclear fuels.
| Material | Total neutron capture relative to graphite (per unit volume) |
Moderating ratio (Avg. 0.1 to 10 eV) |
|---|---|---|
| Heavy water | 0.2 | 11449 |
| Light water | 75 | 246 |
| Graphite | 1 | 863 |
| Sodium | 47 | 2 |
| UCO | 285 | 2 |
| UO2 | 3583 | 0.1 |
| 2LiF–BeF2 | 8 | 60 |
| LiF–BeF2–ZrF4 (64.5–30.5–5) | 8 | 54 |
| NaF–BeF2 (57–43) | 28 | 15 |
| LiF–NaF–BeF2 (31–31–38) | 20 | 22 |
| LiF–ZrF4 (51–49) | 9 | 29 |
| NaF–ZrF4 (59.5–40.5) | 24 | 10 |
| LiF-NaF–ZrF4 (26–37–37) | 20 | 13 |
| KF–ZrF4 (58–42) | 67 | 3 |
| RbF–ZrF4 (58–42) | 14 | 13 |
| LiF–KF (50–50) | 97 | 2 |
| LiF–RbF (44–56) | 19 | 9 |
| LiF–NaF–KF (46.5–11.5–42) | 90 | 2 |
| LiF–NaF–RbF (42–6–52) | 20 | 8 |
Techniques for preparing and handling molten salt had been first developed at Oak Ridge National Lab.[26] The purpose of salt purification was to eliminate oxides, Sulfur, and metal impurities. Oxides could result in the deposition of solid particles in reactor operation. Sulfur had to be removed because of their corrosive attack on nickel-base alloys at operational temperature. Structural metal such as Chromium, Nickel, and Iron had to be removed for corrosion control.
A water content reduction purification stage using HF and Helium sweep gas was specified to run at 400 °C. Oxide and Sulfur contamination in the salt mixtures were removed using gas sparging of HF - H2 mixture, with the salt heated to 600 °C.[26](p8) Structural metal contamination in the salt mixtures were removed using Hydrogen gas sparging, at 700 °C.[26](p26) Solid ammonium hydrofluoride at 125 °C was proposed as a safer alternative for oxide removal.[27]
The possibility of online processing can be an advantage of the MSR design. Continuous processing would reduce the inventory of fission products, control corrosion and improve neutron economy by removing fission products with high neutron absorption cross-section, especially Xenon. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. Online fuel processing can introduce risks of fuel processing accidents[28](p15), which can trigger release of radio isotopes.
In some thorium breeding scenarios, the intermediate product protactinium-233 would be removed from the reactor and allowed to decay into highly pure uranium-233, an attractive bomb-making material. If left in the fuel, protactinium would absorb too many neutrons to make breeding with a graphite moderator and thermal spectrum possible. More modern design propose to use a larger quantity of thorium. This dilutes the protactinium to such an extent that few protactinium atoms absorb a second neutron or, via a (n, 2n) reaction (in which an incident neutron is not absorbed but instead knocks a neutron out of the nucleus), generate uranium-232. Because U-232 has a short half-life and its decay chain contains hard gamma emitters, it makes the isotopic mix of uranium less attractive for bomb-making. This benefit would come with the added expense of processing a larger quantity of blanket salt. Other designs propose to use heavy water as a super efficient moderator to improve neutron economy allowing more loss to protactinium absorption. However these designs would operate at lower temperatures and thus lower thermal efficiency. The necessary fuel salt reprocessing technology has been demonstrated, but only at laboratory scale. A prerequisite to full-scale commercial reactor design is the R&D to engineer an economically competitive fuel salt cleaning system.
Reprocessing refers to the chemical separation of fissionable uranium and plutonium from spent nuclear fuel. [29] The recovery of uranium or plutonium could be subject to the risk of nuclear proliferation. In the United States the regulatory regime has varied dramatically in different administrations. [29]
In the original 1971 Molten Salt Breeder Reactor proposal, uranium reprocessing was scheduled every ten days as part of reactor operation. [30](p181) Subsequently a once-through fueling design was proposed that limited uranium reprocessing to every thirty years at the end of useful salt life. [31](p98) A mixture of uranium 238 was called for to make sure recovered uranium would not be weapons grade. If reprocessing were to be prohibited then the uranium would be disposed with other fission products.
MSRs can be safer than ordinary light water reactors. Molten salts trap fission products chemically, and react slowly or not at all in air. Also, the fuel salt does not burn in air or water. The core and primary cooling loop is operated at near atmospheric pressure, and has no steam, so a pressure explosion is impossible. Even in the case of an accident, most radioactive fission products would stay in the salt instead of dispersing into the atmosphere. A molten core is meltdown-proof, so the worst possible accident would be a leak.[4] In this case, the fuel salt can be drained into passively cooled storage, managing the accident. Neutron-producing accelerators have even been proposed for some super-safe subcritical experimental designs, and the initiation of thorium transmutation to 233U can be directly accomplished with what is essentially a medical proton-beam source.[32]
Some types of molten salt reactors are very inexpensive. Since the core and primary coolant loop are low pressure, it can be constructed of thin, relatively inexpensive weldments. So, it can be far less expensive than the massive pressure vessel required by the core of a light water reactor. Also, some form of fluid-fueled thorium breeder could use less fissile material per megawatt than any other reactor. Molten salt reactors can run at extremely high temperatures, yielding high efficiencies to produce electricity. The temperatures of some proposed designs are high enough to produce process heat for hydrogen production or other chemical reactions. Because of this, they have been included in the GEN-IV roadmap for further study.[33]
The MSR also has far better neutron economy and, depending on the design, a harder neutron spectrum than conventional light water reactors. So, it can operate with less reactive fuels. Some designs (such as the MSRE) can operate a single design from all three common nuclear fuels. For example, it can breed from uranium-238, thorium or even burn the transuranic spent nuclear fuel from light water reactors. In contrast, a water-cooled reactor cannot completely consume the plutonium it produces, because the increasing impurities from the fission wastes capture too many neutrons, "poisoning" the reaction.[34]
MSRs scale over a wide range of powers. Reactors as small as several megawatts have been constructed and operated. Theoretical designs up to several gigawatts have been proposed.[35]
Because of their lightweight structures and compact cores, MSRs weigh less per watt (that is, they have a greater "specific power") than other proven reactor designs. So, in small sizes, with long refueling intervals, they are an excellent choice to power vehicles, including ships, aircraft and spacecraft. This was proved by their initial prototype, the aircraft reactor experiment.[3]
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