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A molten salt reactor (MSR) 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. HISTORY The aircraft reactor experiment See Also: 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. The Molten-Salt Reactor Experiment See Also: Molten-Salt Reactor Experiment Oak Ridge National Laboratory 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 an inherently safe epithermal Thorium breeder reactor. It tested molten salt fuels of Uranium and Plutonium. The tested 233UF4 fluid fuel has a Unique Decay Path that minimizes waste, with waste isotopes having half-lives under 50 years. The red-hot 650°C temperature of the reactor could power high-efficiency heat engines such as 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 operated as hot as 650°C and operated for the equivalent of about 1.5 years of full power operation. (For more information, see the main article) 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 molten salt reactors have been built other than the experimental MSRE. Liquid salt very high temperature reactor See Also: 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. TECHNOLOGICAL ISSUES Molten-salt Fueled Reactors The classic MSFR has been very exciting to many nuclear engineers. Its most prominent champion was Alvin Weinberg , who patented the light-water reactor, and was a director of the U.S.'s Oak Ridge National Laboratory, a prominent nuclear research center. Two concepts were investigated. The "two fluid" reactor had a high-neutron-density core that burned U233 from the Thorium fuel cycle. A blanket of thorium salts absorbed the neutrons and was eventually transmuted to U233 fuel. The engineers discovered that by carefully sculpting the moderator rods (to get neutron densities similar to a core and blanket), and modifying the fuel reprocessing chemistry, both Thorium and Uranium salts could coexist in a simpler, less expensive yet efficient "one fluid" reactor. The power reactor design produced by Weinberg's research group was similar to the MSRE above, which was designed to validate the risky hot, high-neutron-density "kernel" part of the "kernel and blanket" thorium breeder. The advantages cited by Weinberg and his associates at Oak Ridge National Laboratory include:
Combining the above, some form of molten-salt thorium breeder could be the most efficient well-developed energy source known, whether measured by cost per kW, capital cost or social costs. There are some design and social advantages:
The Th/U233 molten salt reactor is not flawless. Known problems include:
] Viewed 6/2007 Most MSR designs do not use graphite as a structural material, and arrange for it to be easy to replace. At least one design used graphite balls floating in salt, which could be removed and inspected continuously without shutting down the reactor. ORNL-4548: Molten-Salt Reactor Program: Semiannual Progress Report for Period Ending February 28, 1970, pg. 57
An MSR based on chloride salts has many of the same advantages. However, the larger, less-dense atoms of Chlorine causes the reactor to be a fast breeder. Theoretically, it wastes even fewer neutrons and breeds more efficiently, though it may be less safe. It would require a salt with an isotopically-separated Chlorine, Cl37, to prevent neutronic activation of the Chlorine into sulfur which would form corrosive sulfur chloride. Molten-salt cooled reactors Molten-salt-fueled reactors (MSFR) are quite different from molten-salt-cooled reactors (MSCR), a Gen IV proposal. The MSCR can't 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 proft by selling fuel assemblies. Also, the reactor's core retains many safety and cost advantages. Notably, there's 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, light weight Brayton Cycle gas turbine. Much of the current research on MSCRs is focused on small compact Heat Exchanger s. 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, Carbide s, and refractory metal based or ODS alloys might be feasible. 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. Beryllium is popular because when a neutron hits a Beryllium necleus it "doubles" the neutron, improving neutron economy. For the fuel carrying salts, generally 1% or 2% by mole fraction of UF4 is added, however 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. 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'') 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. POLITICAL ISSUES To exploit the molten salt reactor's breeding potential to the fullest, the reactor must be co-located with a reprocessing facility. 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 that they don't fit in the current nuclear economy. Today the nuclear industry profits not from building reactors, but by selling fuel bundles to the reactor operator. This business model is inapplicable to a molten-salt fueled reactors because the fuel does not need to be fabricated. MSRs are also inefficient breeders of fuel for light-water reactors because natural Uranium is too cheap and the MSR's breeding capabilities are too slow. For these reasons, nuclear companies do not have an interest in commercializing the MSR. COMPARISON TO ORDINARY LIGHT WATER REACTOR S Molten salt reactors are an immature technology. No large-scale reactor has been built and operated for a long period, and unexpected problems are likely. Whether an MSR will be economically and technologically viable is unknown. Small, experimental MSRs have operated as long as several years, and the problems were fixed. MSRs may be safer. 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 atmospheric pressure, and has no steam, so a pressure explosion is impossible. Even in the unlikely case of an accident, most radioactive fission products would stay in the salt instead dispersing into the atmosphere. A molten core is meltdown-proof, so the worst possible accident would be a leak. 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. Some types of molten salt reactor are very efficient. 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, with exremely high efficiencies when producing electricity. The tempearture 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. Molten-salt-fueled thorium breeders close the Nuclear Fuel Cycle and potentially eliminate the need for both fuel enrichment and fuel fabrication, both major expenses. The MSR also has far better neutron economy and, depending on the design, a harder neutron spectrum. 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 Reactor s. 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. Molten salt-fueled thorium breeders can operate for extended periods, possibly decades, without refueling, by chemically precipitating neutronic poisons. 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 proposedWeinberg et al. WASH 1080, ORNL. Because of their low 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. REFERENCES
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