Molten salt reactor
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A molten salt reactor is a type of nuclear reactor where the primary coolant is a molten salt.
There have been many designs put forward for use of this type of reactor as a nuclear power plant and a few prototypes built. The concept is one of those proposed for development as a generation IV reactor.
The early concepts and many current ones had the nuclear fuel dissolved in the molten fluoride salt coolant as uranium tetrafluoride (UF4), the fluid would reach criticality by flowing into a graphite core which also served as the moderator. Many current concepts rely on ceramic fuel that is dispersed in a graphite matrix with the molten salt providing low pressure, high temperature cooling.
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[edit] History
[edit] The aircraft reactor experiment
Extensive research into molten salt reactors started with the US Aircraft Reactor Experiment (ARE). The US Aircraft Reactor Experiment 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 resulted in several experiments. Three of which resulted in engine tests collectively called the Heat Transfer Reactor Experiments, of which there were three iterations: HTRE-l, HTRE-2, and HTRE-3. One experiment used the molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel and was moderated by beryllium oxide (BeO), used liquid sodium as a secondary coolant, and had a peak temperature of 860°C, it operated for a 1000 hour cycle in 1954. This experiment used Inconel 600 alloy for the metal structure and piping.
[edit] The Molten-Salt Reactor Experiment
Oak Ridge National Laboratory took the lead in researching the MSR through 1970s and much of their work culminated with the Molten-Salt Reactor Experiment (MSRE). The molten-salt reactor experiment was a 7.4 MWth test reactor. It was located at ORNL, its piping and structural components were made from Hastelloy-N, and its core was 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 operated at a peak temperature of 650°C and operated for the equivalent of about 1.5 years of full power operation.
[edit] Oak Ridge National Laboratory reactor
The culmination of the Oak Ridge National Laboratory research during the 1970-76 timeframe resulted in a MSR 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. However, to date the molten salt reactor remains a "paper design", that is, no commercial molten salt reactors have been built besides the MSRE experiment, which did not operate for very long.
[edit] Liquid salt very high temperature reactor
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 picked 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). 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. The fuel graphite would be in the form of graphite rods that would be inserted in hexagonal moderating graphite blocks. The molten salt would pass through holes drilled in the graphite blocks. 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 case an accident occurred.
[edit] Technological issues
Much of the current research on this class of reactor designs is focused on small compact heat exchangers. Since the coolant in the designs which rely on dispersed fuel in the molten salt will be undergoing fission, it would be extremely radioactive. Because of the high amounts of radioactivity, all tubes containing the primary, fuel bearing, molten salt must be shielded, adding to the expense of construction. By using smaller heat exchangers, less molten salt needs to be used and therefore significant cost savings could be achieved.
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, especially since at 850°C thermo chemical production of hydrogen becomes possible. Materials for this temperature range have not yet been found, though carbon composites, carbides, and refractory metal based or ODS alloys might be feasible.
[edit] Fused salt selection
The types of fused salts that are chosen come from an optimization of salt characteristics. Fused fluorides are generally chosen over other salts because of the usefulness of the elements without isotope separation, better neutron economy and moderating efficiency, lower vapor pressure and better chemical stability. Chlorides have also been considered for molten salt reactors, but not nearly as much work has been done on reactor designs that utilize them. Additionally, whenever lithium fluoride is used as part of the salt composition, the lithium must be enriched to a very high purity (99.999%?) in lithium-7 to get tritium production under control.
Due to the high "redox window" available for fused fluoride salts, allowing for the chemical potential of the fused salt system to be manipulated, the following types of salts are the most promising. FLiBe can be used in conjunction with beryllium additions to drive down the electrochemical potential and virtually eliminate corrosion issues. However, beryllium is extremely toxic to humans. Many other salts have potential corrosion issues, especially at the elevated temperatures being talked about for future hydrogen production facilities.
To date, most research has focused on FLiBe for the nuclear heat transport system. For the fuel carrying salts, generally 1% or 2% by mole fraction of UF4 is added, however thorium and plutonium fluorides could be used.
| 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 |
Above is a table comparing the neutron capture and moderating efficiency of several materials. Red are Be bearing salts, blue are ZrF4 bearing salts, and green are LiF bearing salts. (Source: ORNL/TM-2005/218, Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR), December 2005, D. T. Ingersoll)
[edit] Fused salt purification and reprocessing
Salts must be extremely pure initially, and would most likely be continuously cleaned in a large-scale molten salt reactor. Any water vapor in the salt will form hydrofluoric acid (HF) which is extremely corrosive. Other impurities can cause non-beneficial chemical reactions and would most likely have to be cleansed from the system. It should be noted that most power plants have to ensure that the primary coolant they are using is extremely pure; otherwise, they would encounter corrosion issues as well.
The possibility of online reprocessing can be an advantage of the MSR design. Continuous reprocessing ensures a low inventory of fission products at all times, which improves neutron economy. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. To allow breeding from thorium, the intermediate product protactinium has to be removed from the reactor and stored for some months while it decays into uranium 233. Left in the fuel it would absorb too many neutrons to make breeding with a graphite moderator and thermal spectrum possible (though some heavy water moderated reactor designs could overcome this, albeit at a lower thermal efficiency ). The necessary reprocessing technology, which has to process the complete fuel every 10 days, has only been demonstrated at laboratory scale. For a power reactor such a large reprocessing facility is currently deemed uneconomic.
[edit] Political issues
To exploit the molten salt reactors potential to the fullest, the reactor must be co-located with a reprocessing facility. While this simple electrochemical reprocessing would close the fuel cycle and make transports of spent fuel almost unnecessary while producing no toxic byproducts, any kind of nuclear reprocessing is still illegal in many countries. Some people fear that operating an MSR could pave the way to the plutonium economy with its associated proliferation dangers. (A similar argument lead to the shutdown of the Integral Fast Reactor project in 1994.)
Molten salt reactors are sufficiently different from solid core reactors so that they don't fit in the current nuclear economy. Today the nuclear industry makes no profit building reactors, but by selling fuel bundles to the reactor operator.[verification needed] This business model is inapplicable to an MSR because fuel does not need to be manufactured any more. MSRs do not fit as producers of fissile fuel for water-cooled reactors either because natural uranium is too cheap and the MSR's breeding capabilities are too weak. For these reasons, nuclear companies do not have an interest in commercializing the MSR.[verification needed]
[edit] Comparison to water cooled reactors
A primary benefit of this technology is that it closes the nuclear fuel cycle and potentially eliminates the need for fuel fabrication, however, it may be necessary for "drivers" to still be fabricated (depends on design). The molten salt reactor has the potential to use a minimal amount of fissile material per megawatt generated compared to any other current reactor. Molten salt reactors can also potentially be run at extremely high temperatures, such as those proposed for hydrogen production facilities. Because of this, they have been included in the GEN-IV roadmap for further study. Molten salts are also extremely good at trapping fission products, and are mostly non-reactive in air.
Generally speaking, molten salt reactors are an immature technology. No large-scale reactor has been built and operated today, and unexpected difficulties will surely be found. Whether an MSR will be economically and technologically viable is yet to be determined.
The accident potential of the MSR is far lower than that of a water reactor. The primary cooling loop is operated at atmospheric pressure and the fuel salt does not react violently with air or water. Even in the unlikely case of an accident, most radioactive fission products would stay in the salt instead of being dispersed into the atmosphere. An already molten core is of course meltdown-proof, the worst possible accident would be a leak. In this case, the fuel salt can easily be drained into passively cooled storage, making the accident manageable.
The MSR has far better neutron economy and, depending on the design, a harder neutron spectrum. This makes it far less demanding on the quality of the fuel, allowing all three major nuclear fuels to be used, breeding from uranium-238 and thorium and even burning of transuranic wastes. In contrast, a water-cooled reactor cannot completely consume the plutonium it produces, because the increasing impurities would capture too many neutrons.
Compared to water, molten salts allow higher operating temperatures, which in turn give better efficiency in electricity generation and makes cooling easier. High operating temperatures are also of potential use in the chemical industry.
[edit] References
- J.H. Devan et al. (unknown date). Material Considerations for Molten Salt Accelerator-based Plutonium Conversion Systems, pg. 475-486
- W.D. Manely et al. (1960). Metalurgical Problems in Molten Fluoride Systems. Progress in Nuclear Energy, VOl. 2, pg. 164-179
- Bruce Hoglund's Eclectic Interests Home Page Nuclear Power, Thorium, Molten Salt reactors, etc.
- "The First Nuclear Era : The Life and Times of a Technological Fixer", by Alvin Martin Weinberg (1994). Book by a former director of the Oak Ridge National Laboratory, and a promoter of nuclear power and molten salt reactors.
