Molten Salt Reactors
(Updated January 2015)
- Molten Salt Reactors (MSR) operated in the 1960s.
- They are seen as a promising technology today principally as a thorium fuel cycle prospect.
- A variety of designs is being developed, many as fast neutron types.
- Some have solid fuel similar to HTR fuel.
MSRs use molten fluoride salts as primary coolant, at low pressure. This itself is not a radical departure when the fuel is solid and fixed. But extending the concept to dissolving the fissile and fertile fuel in the salt certainly represents a leap in lateral thinking relative to nearly every reactor operated so far. However, the concept is not new, as outlined below. MSRs may operate with epithermal or fast neutron spectrums, and with a variety of fuels. Much of the interest today in reviving the MSR concept relates to using thorium (to breed fissile uranium-233). There are a number of different MSR design concepts, and a number of interesting challenges in the commercialisation of many, especially with thorium.
The salts concerned as primary coolant, mostly lithium-beryllium fluoride and lithium fluoride, remain liquid without pressurization from about 500°C up to about 1400°C, in marked contrast to a PWR which operates at about 315°C under 150 atmospheres pressure. The ultimate MSR concept is to have the fuel dissolved in the coolant as fuel salt. Intermediate designs and the AHTR have fuel particles in solid graphite and have less potential for thorium use.
During the 1960s, the USA developed the molten salt breeder reactor concept as the primary back-up option for the fast breeder reactor (cooled by liquid metal) and a small prototype 8 MWt Molten Salt Reactor Experiment (MSRE) operated at Oak Ridge over four years to 1969 (the MSR program ran 1957-1976). In the first campaign (1965-68), uranium-235 tetrafluoride (UF4) enriched to 33% was dissolved in molten lithium, beryllium and zirconium fluorides at 600-700°C which flowed through a graphite moderator at ambient pressure. The fuel comprised about one percent of the fluid.
The coolant salt in a secondary circuit was lithium + beryllium fluoride (FLiBe).* There was no breeding blanket, this being omitted for simplicity in favour of neutron measurements.
The original objectives of the MSRE were achieved by March 1965, and the U-235 campaign concluded. A second campaign (1968-69) used U-233 fuel which was then available, making MSRE the first reactor to use U-233. This program prepared the way for building a MSR breeder utilising thorium.
The R&D program demonstrated the feasibility of this system, including online reprocessing, and highlighted some unique corrosion and safety issues that would need to be addressed if constructing a larger pilot MSR. It also showed that breeding required a different design, with a larger blanket loop and two fluids (heterogeneous). Tritium production was a problem (see below re lithium enrichment).
There is now renewed interest in the MSR concept in Japan, Russia, China, France and the USA, and one of the six Generation IV designs selected for further development is the MSR in two distinct variants, the molten salt fast reactor (MSFR) and the Advanced High Temperature Reactor (AHTR) – also known as the fluoride salt-cooled high-temperature reactor (FHR) with solid fuel, or PB-FHR specifically with pebble fuel.
In the normal or basic MSR concept, the fuel is a molten mixture of lithium and beryllium fluoride (FLiBe) salts with dissolved enriched uranium (U-235 or U-233) fluorides (UF4). The core consists of unclad graphite moderator arranged to allow the flow of salt at about 700°C and at low pressure. Much higher temperatures are possible but not yet tested. Heat is transferred to a secondary salt circuit and thence to steam or process heat. The basic design is not a fast neutron reactor, but with some moderation by the graphite is epithermal (intermediate neutron speed) and breeding ratio is less than 1.
However, this concept, with fuel dissolved in the salt, is further from commercialisation than solid fuel designs, where the ceramic fuel may be set in prisms, plates, or pebbles.
Considering liquid-fuel MSR designs, thorium can be dissolved with the uranium in a single fluid MSR, known as a homogeneous design. Two-fluid, or heterogeneous MSRs, would have fertile salt containing thorium in a second loop separate from the fissile salt containing uranium and could operate as a breeder reactor (MSBR). Here, the U-233 is progressively removed and transferred to the primary circuit. However, graphite degradation from neutron flux limits the useful life of the reactor core with the fuel and breeding fluids in close juxtaposition, and in the 1960s it was assumed that the entire reactor vessel in the two-fluid design would be replaced after about eight years.*
The fission products dissolve in the fuel salt and are removed continuously in an adjacent on-line reprocessing loop and replaced with fissile uranium, plutonium and other actinides or, potentially, fertile Th-232 or U-238. Xenon is removed rapidly by outgassing, but protactinium is a problem with thorium as a fuel source. (It is an intermediate product in producing U-233 and is a major neutron absorber.) Constant removal of fission products means that a much higher fuel burn-up can be achieved (> 50%) and the removal of fission products means less decay heat to contend with after reactor shut down. Actinides are fully recycled and remain in the reactor until they fission or are converted to higher actinides which do so. The high-level waste comprises fission products only, hence shorter-lived radioactivity.
Compared with solid-fuelled reactors, MSR systems with fuel salt are claimed to have lower fissile inventories*, no radiation damage constraint on fuel burn-up, no requirement to fabricate and handle solid fuel or solid used fuel, and a homogeneous isotopic composition of fuel in the reactor. These and other characteristics may enable MSRs to have unique capabilities and competitive economics for actinide burning and extending fuel resources. Safety is high due to passive cooling up to any size.
MSRs have large negative temperature and void coefficients of reactivity, and are designed to shut down due to expansion of the fuel salt as temperature increases beyond design limits. The negative temperature and void reactivity coefficients passively reduces the rate of power increase in the case of an inadvertent control rod withdrawal (technically known as a ‘reactivity insertion’). When tests were made on the MSRE, a control rod was intentionally withdrawn during normal reactor operations at full power (8 MWt) to observe the dynamic response of core power. Such was the rate of fuel salt thermal expansion that reactor power levelled off at 9 MWt without any operator intervention.
In the MSBR, the reactor-grade U-233 bred in the secondary circuit needs to be removed, or it will fission there and contaminate that circuit with ‘hot’ fission products. Therefore in practice the protactinium (Pa-233) formed from the thorium needs to be removed before it decays to U-233*, but this process is unproven at any scale. It is relatively easy to remove the U-233 from the Pa-233 by fluorination to UF6 before reducing it to UF4 for adding to the primary fuel salt circuit. However, the U-233 is contaminated with up to 400 ppm U-232 which complicates processing, due to its highly gamma-active decay progeny.
* Th-232 gains a neutron to form Th-233, which soon beta decays (half-life 22 minutes) to protactinium-233. The Pa-233 (half-life of 27 days) decays into U-233. Some U-232 is also formed via Pa-232 along with Th-233, and a decay product of this is very gamma active.
Primary and secondary cooling, the fluoride salts
Fluoride salts have very low vapour pressure even at red heat, carry more heat than the same volume of water, have reasonably good heat transfer properties, are not damaged by radiation, do not react violently with air or water, and are inert to some common structural metals. However having the fuel in solution also means that the primary coolant salt becomes radioactive, complicating maintenance procedures, and the chemistry of the salt must be monitored closely to maintain a chemically reduced state to minimise corrosion. Also the beryllium in the salt is toxic, which leads to at least one design avoiding it, though this requires higher temperatures to keep LiF liquid. LiF however can carry a higher concentration of uranium than FLiBe, allowing less enrichment.
Lithium used in the salt must be fairly pure Li-7, since Li-6 produces tritium when fissioned by neutrons. Li-7 has a very small neutron cross-section (0.045 barns). This means that lithium must be enriched beyond its natural 92.5% Li-7 level. Lithium-7 is being produced at least in Russia and possibly China today as a by-product of enriching lithium-6 to produce tritium for thermonuclear weapons. See also Lithium paper.
LiF is exceptionally stable chemically, and the LiF-BeF2 mix ('FLiBe')* is eutectic (at 459°C it has a lower melting point than either ingredient – LiF is about 500°C). It boils at 1430°C. It is favoured in MSR and AHTR primary cooling and when uncontaminated has a low corrosion effect. The three nuclides (Li-7, Be, F) are among the few to have low enough thermal neutron capture cross-sections not to interfere with fission reactions. LiF without the toxic beryllium solidifies at about 500°C and boils at about 1200°C. FLiNaK (LiF-NaF-KF) is also eutectic and solidifies at 454°C and boils at 1570°C. It has a higher neutron cross-section than FLiBe or LiF but can be used intermediate cooling loops.
The hot molten salt in the primary circuit can be used with secondary salt circuit or secondary helium coolant generating power via the Brayton cycle as with HTR designs, with potential thermal efficiencies of 48% at 750°C to 59% at 1000°C, or simply with steam generators. In industrial applications molten fluoride salts (possibly simply cryolite – Na-Al fluoride) are a preferred interface fluid in a secondary circuit between the nuclear heat source and any chemical plant. The aluminium smelting industry provides substantial experience in managing them safely.
In the secondary cooling circuit of the AHTR concept, air is compressed, heated, flows through gas turbines producing electricity, enters a steam recovery boiler producing steam that produces additional electricity, and exits to the atmosphere. Added peak power can be produced by injecting natural gas (or hydrogen in the future) after nuclear heating of the compressed air to raise gas temperatures and plant output, giving it rapidly variable output (of great value in grid stability and for peak load demand where renewables have significant input). This is described as an air Brayton combined-cycle (ABCC) system in secondary circuit.
In the 1960s MSRE, an alternative secondary coolant salt considered was 8% NaF + 92% NaBF4 with melting point 385°C, though this would be more corrosive.
MSR research emphasis
American researchers and the China Academy of Sciences/ SINAP are working primarily on solid fuel MSR technology. The main reason is that this is a realistic first step. In China this is focused on thorium-fueled versions (see LFTR/ TMSR below). The technical difficulty of using molten salts is significantly lower when they do not have the very high activity levels associated with them bearing the dissolved fuels and wastes. The experience gained with component design, operation, and maintenance with clean salts makes it much easier then to move on and consider the use of liquid fuels, while gaining several key advantages from the ability to operate reactors at low pressure and deliver higher temperatures.
Russia's Molten Salt Actinide Recycler and Transmuter (MOSART) is a fast reactor fuelled only by transuranic fluorides from uranium and MOX LWR used fuel. It is part of the MARS project (minor actinide recycling in molten salt) involving RIAR, Kurchatov and other research organisations. The 2400 MWt design has a homogeneous core of Li-Na-Be or Li-Be fluorides without graphite moderator and has reduced reprocessing compared with original US design. Thorium may also be used, though it is described as a burner-converter rather than a breeder.
In the Generation IV program for the MSR, collaborative R&D is pursued by interested members under the auspices of a provisional steering committee. There will be a long lead time to prototypes, and the R&D orientation has changed since the project was set up, due to increased interest. It now has two baseline concepts:
- The Molten Salt Fast Neutron Reactor (MSFR), which will take in thorium fuel cycle, recycling of actinides, closed Th/U fuel cycle with no U enrichment, with enhanced safety and minimal wastes. it is a liquid-fuel design.
- The Advanced High-Temperature Reactor (AHTR) – also known as the fluoride salt-cooled high-temperature reactor (FHR) – with the same graphite and solid fuel core structures as the VHTR and molten salt as coolant instead of helium, enabling power densities 4 to 6 times greater than HTRs and power levels up to 4000 MWt with passive safety systems. A 5 MWt prototype is under construction at Shanghai Institute of Nuclear Applied Physics (SINAP, under the China Academy of Sciences) with 2015 target for operation.
The GIF 2014 Roadmap said that a lot of work needed to be done on salts before demonstration reactors were operational, and suggested 2025 as the end of the viability R&D phase.
China's dual program
The China Academy of Sciences in January 2011 launched an R&D program on LFTR, known there as the thorium-breeding molten-salt reactor (Th-MSR or TMSR), and claimed to have the world's largest national effort on it, hoping to obtain full intellectual property rights on the technology. The TMSR Research Centre has a 5 MWe MSR prototype under construction at Shanghai Institute of Applied Physics (SINAP, under the Academy) with 2015 target for operation.
SINAP has two streams of TMSR development – solid fuel (TRISO in pebbles or prisms/ blocks) with once-through fuel cycle, and liquid fuel (dissolved in FLiBe coolant) with reprocessing and recycle.
- The TMSR-SF stream has only partial utilization of thorium, relying on some breeding as with U-238, and needing fissile uranium input as well. SINAP aims at a 2 MW pilot plant by about 2015, and a 100 MWt demonstration pebble bed plant with open fuel cycle by about 2025. TRISO particles will be with both low-enriched uranium and thorium, separately.
- The TMSR-LF stream claims full closed Th-U fuel cycle with breeding of U-233 and much better sustainability with thorium but greater technical difficulty. SINAP aims for a 10 MWt pilot plant by 2025 and a 100 MWt demonstration plant by 2035.
- A TMSFR-LF fast reactor optimized for burning minor actinides is to follow.
SINAP sees molten salt fuel being superior to the TRISO fuel in effectively unlimited burn-up, less waste, and lower fabricating cost, but achieving lower temperatures (600°C+) than the TRISO fuel reactors (1200°C+). Near-term goals include preparing nuclear-grade ThF4 and ThO2 and testing them in a MSR.
The US Department of Energy is collaborating with the China Academy of Sciences on the program, which had a start-up budget of $350 million. The target date for TMSR commercial deployment is 2032.
Other solid-fuel types
AHTR / FHR
Research on molten salt coolant has been revived at Oak Ridge National Laboratory (ORNL) in the USA with the Advanced High Temperature Reactor (AHTR). This is a larger reactor using a coated-particle graphite-matrix fuel like that in the GT-MHR (see Small Reactors paper) and with molten fluoride salt as primary coolant. It is also known as the Fluoride High-temperature reactor (FHR). While similar to the gas-cooled HTR it operates at low pressure (less than 1 atmosphere) and higher temperature, and gives better heat transfer than helium. The FLiBe salt is used solely as coolant, and achieves temperatures of 750-1000°C or more while at low pressure.
This could be used in thermochemical hydrogen manufacture. As noted above, a 5 MW thorium-fueled prototype is under construction at Shanghai Institute of Nuclear Applied Physics (SINAP, under the China Academy of Sciences) with 2015 target for operation. A 100 MWt demonstration pebble bed plant with open fuel cycle is planned by about 2025. SINAP sees this design having potential for higher temperatures than MSRs with fuel salt.
In the USA a consortium including UC Berkeley, ORNL and Westinghouse is designing a 100 MWe pebble bed FHR, with modular construction and able to deliver 240 MWe with gas co-firing. A 410 MWe/900 MWt pebble bed version was also being designed with UC-Berkeley.
AHTR reactor sizes of 1500 MWe/3600 MWt are envisaged, with capital costs estimated at less than $1000/kW.
Other liquid-fuel types
Liquid Fluoride Thorium Reactor (LFTR/TMSR)
The Liquid Fluoride Thorium Reactor (LFTR) is a heterogeneous MSR design which breeds its U-233 fuel from a fertile blanket of lithium-beryllium fluoride (FLiBe) salts with thorium fluoride. Some of the neutrons released during fission of the U-233 salt in the reactor core are absorbed by the thorium in the blanket salt. The resulting U-233 is separated from the blanket salt and in FLiBe becomes the liquid core fuel. LFTRs are designed to be able to rapidly change their power output, and hence be used for load-following.
Actinides are less-readily formed than in fuel with atomic mass greater than 235. The blanket circuit contains a significant amount of thorium tetrafluoride in the molten FLiBe salt. Newly-formed U-233 forms soluble uranium tetrafluoride (UF4), which is converted to gaseous uranium hexafluoride (UF6) by bubbling fluorine gas through the salt (which does not chemically affect the less-reactive thorium tetrafluoride). The volatile uranium hexafluoride is captured, reduced back to soluble UF4 by hydrogen gas, and finally is directed to the core to serve as fissile fuel. Protactinium – a neutron absorber – is not a major problem in the blanket salt.
Safety is achieved with a freeze plug which if power is cut allows the fuel to drain into subcritical geometry in a catch basin.
Because they are expected to be inexpensive to build and operate, 100 MWe LFTRs could be used as peak and back-up reserve power units. They would normally operate at around 700°C, and hence have potential for process heat.
Flibe Energy has a 40 MW thermal LFTR design that uses FLiBe salt as its coolant. This is based on earlier US work on the MSR Program.
Canada-based Terrestrial Energy Inc (TEI) has designed the Integral MSR. This simplified MSR integrates the primary reactor components, including primary heat exchangers to secondary clean salt circuit, in a sealed and replaceable core vessel that has a projected life of seven years. The IMSR will operate at approximately 700°C, which can support many industrial process heat applications. It operates in the thermal neutron spectrum with a hexagonal arrangement of graphite elements forming the moderator. The fuel-salt is a eutectic of low-enriched uranium-235 fuel (as UF4) and a fluoride carrier salt at atmospheric pressure. Emergency cooling and residual heat removal are passive. The IMSR is designed in three sizes: 80 MWth, 300 MWth, and 600 MWth. The total levelized cost of electricity from the largest is projected to be competitive with natural gas. The smallest is designed for off-grid, remote power applications. The company hopes to commission its first commercial reactor by the early 2020s. In January 2015 the company announced a collaborative agreement with US Oak Ridge National Laboratory (ORNL) to advance the design.
Transatomic Power Corp is a new US company partly funded by Founders Fund and aiming to develop a MSR using very low-enriched uranium fuel (1.8%) or the entire actinide component of used LWR fuel. The TAP reactor has an efficient zirconium hydride* moderator and a LiF-based fuel salt bearing the UF4, hence a very compact core. Owing to the ZrH moderator, there are significantly more neutrons in the thermal region (less than 1 eV) compared with a graphite moderator, thereby enabling the reactor to generate power from very low-enriched uranium or used LWR fuel. The epithermal (1 eV - 1 MeV) spectrum is lower than that with graphite, but in the fast spectrum (over 1 MeV) the neutron flux is greater than with graphite moderator, and therefore contributes strongly to actinide burning.
* as used in TRIGA research reactors and TOPAZ and SNAP reactors for space program.
The envisaged first commercial plant will be 1250 MWt/ 550 MWe running at 44% thermal efficiency with 650°C in primary loop, using steam cycle via an intermediate loop with FLiNaK salt (LiF-KF-NaF). It would give up to 96% actinide burn-up. It has negative void and thermal coefficients, and the moderator starts to fail at higher temperatures due to hydrogen loss. Decay heat removal can be by convection. The overnight cost for an nth-of-a-kind 550 MWe plant, including lithium-7 inventory and on-line fission product removal and storage, is estimated at $2 billion with a three-year construction schedule. A version of the reactor may utilize thorium fuel.
The Fuji MSR is a 100-200 MWe design to operate as a near-breeder and was being developed internationally by a Japanese, Russian and US consortium: the International Thorium Molten Salt Institute (ITHMSI). A 10 MWe mini Fuji is also on the drawing board.
Martingale in USA is designing the ThorCon MSR, which is a 250 MWe scaled-up Oak Ridge MSRE, and aims for an operating prototype by 2020, with modular construction. Several such units would comprise a power station, and a 1000 MWe Thorcon plant would comprise about 200 factory- or shipyard-build modules installed below grade (30 m down). All components are deigned to be easily and frequently replaced. For instance, every four years the entire primary loop would be changed out, returned to a centralized recycling facility, decontaminated, disassembled, inspected, and refurbished. Incipient problems would be rectified and major upgrades could be introduced without significantly disrupting power generation. The company claims generation costs of 3 to 5 c/kWh depending on scale, and is "targeting its first installations in forward-looking countries that support technology-neutral nuclear regulations and see the benefits of the license-by-test process."
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