The culmination of the Oak Ridge research over 1970-76 resulted in an MSR design that would use LiF-BeF 2-ThF 4-UF 4 (72:16:12:0.4) as primary coolant with fuel. A total of 10,000 operating hours was logged. This programme prepared the way for building a MSR breeder utilizing thorium, which would operate in the thermal (slow) neutron spectrum. A second campaign (1968-69) used U-233 fuel which was then available, making MSRE the first reactor to use U-233, though it was imported and not bred in the reactor. The original objectives of the MSRE to demonstrate the concept as a simple reliable reactor were achieved by March 1965, and the U-235 campaign concluded. * Fuel salt melting point 434☌, coolant salt melting point 455☌. The coolant salt in a secondary circuit was eutectic lithium + beryllium fluoride (FLiBe).* There was no breeding blanket, this being omitted for simplicity in favour of neutron measurements. The fuel comprised about one percent of the fluid.* In the first campaign (1965-68), uranium-235 tetrafluoride (UF 4) enriched to 33% was dissolved in molten lithium, beryllium and zirconium fluorides at 600-650☌ which flowed through a graphite moderator at ambient pressure. It was the primary back-up option for the fast breeder reactor (cooled by liquid metal) and the small prototype 8 MWt Molten Salt Reactor Experiment (MSRE) operated at Oak Ridge over four years to 1969 (the MSR programme ran 1957-1976). * It had primary coolant of NaF-ZrF 4 with UF 4 and secondary coolant of NaK at 880☌, to produce 60 MWt.ĭuring the 1960s, the USA developed the molten salt breeder reactor concept at the ORNL. The US MSR programme originated in the Aircraft Reactor Experiment* at the Oak Ridge National Laboratory (ORNL), Tennessee (built as part of the wartime Manhattan Project). Graphite as moderator is chemically compatible with the fluoride salts. Intermediate designs and the AHTR have fuel particles in solid graphite and have less potential for thorium use. Thorium, uranium, and plutonium all form suitable fluoride salts that readily dissolve in the LiF-BeF 2 (FLiBe) mixture, and thorium and uranium can be easily separated from one another in fluoride form. Batch reprocessing is likely in the short term, and fuel life is quoted at 4-7 years, with high burn-up. The main MSR concept is to have the fuel dissolved in the coolant as fuel salt, and ultimately to reprocess that online. The salts concerned as primary coolant, mostly lithium-beryllium fluoride and lithium fluoride, remain liquid without pressurization from about 500☌ up to about 1400☌, in marked contrast to a PWR which operates at about 315☌ under 150 atmospheres pressure. There are a number of different MSR design concepts, and a number of interesting challenges in the commercialisation of many, especially with thorium. Much of the interest today in reviving the MSR concept relates to using thorium (to breed fissile uranium-233), where an initial source of fissile material such as plutonium-239 needs to be provided. MSRs may operate with epithermal or fast neutron spectrums, and with a variety of fuels. However, the concept is not new, as outlined below. 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. This itself is not a radical departure when the fuel is solid and fixed. Molten salt reactors (MSRs) use molten fluoride salts as primary coolant, at low pressure. Some have solid fuel similar to HTR fuel, others have fuel dissolved in the molten salt coolant.Global research is currently led by China.A variety of designs is being developed, some as fast neutron types.They are seen as a promising technology today principally as a thorium fuel cycle prospect or for using spent LWR fuel.
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