In previous posts we have discussed thorium reactors based on solid fuels. There is another approach to creating a reactor that will burn thorium. Thorium and uranium can be dissolved in liquid composed of molten salts. The liquid is then pumped between a core and a heat exchanger where the heat is transferred generate steam.
In the single liquid design, the thorium and uranium dissolved in the molten salts sit in a big reaction vessel with graphite rods for moderation. If breeding U-233 is needed, the core must be large. Considerable reprocessing is required to recover U-233. If not breeding U-233, then uranium refueling is required.
In a two liquid design, there is a high density core that burns the U-233. There is a separate shell of thorium salts which absorbs neutrons and produces protactinium-233 which eventually decays to U-233. It is relatively easy to remove the U-233 from the outer shell and put it in the core for fuel. Benefits are simplified fuel processing with the thorium kept separate from the core and the smaller amount of fissile material required to start the reactor. There are concerns with the complexity of the graphite plumbing and the effect of neutrons on the graphite.
There is a third design that is a hybrid of the first two. Thorium is included in the fuel in the core. This results in a mixture of the advantages and disadvantages of the first two designs.
Between 1965 and 1969, the Molten-Salt Reactor Experiment was conducted at Oak Ridge National Laboratory (ORNL) to model the concept of a liquid fluoride thorium reactor. The functional core was built but the expensive thorium breeding blanket in the design was not included. Pyrolytic graphite was used as a moderator with a fluorine – lithium – beryllium compound as a secondary coolant. The fuel was a combination of molten lithium, beryllium, zirconium, fluorine salts with U-233 dissolved in the salts. It was turned on in 1965 and was operated successfully for the equivalent of 1.5 years reaching temperatures of 650 °C.
Research continued at ORNL between 1970 and 1976 resulting the Molten Salt Breeder Reactor design for a thorium reactor. Thorium would be added to the molten salt mix, the secondary coolant would be a sodium – fluorine – beryllium compound. The reactor would have a maximum operating temperature of 650 °C. The funding was cancelled because the technology was not well understood and the AEC was allocating available funds to a different breeder reactor programmed based on uranium.
There are many potential advantages to a thorium molten salt reactor. There is a lot of thorium available for fuel. The waste given off has less of the actinide transuranic wastes of uranium reactors and more short lived radioactive materials. They can be used to burn some types of reactor waste products reducing need for waste disposal. They can react to changes in electrical load demand in less than one minute, far faster than conventional commercial reactors. They operate at lower temperatures and pressures than conventional reactors. They make economical use of neutrons when compared to uranium reactors. These last two factors make it possible to build small thorium reactors which could be used in ships or even airplanes.
Molten Salt Reactor graphite core:
