In chemistry a salt is an ionic compound that is produced when acid undergoes a neutralizing reaction with a base. Some salts are compounds of metallic and non-metallic elements. Salts become molten (or liquid) when heated. Different salts have very different melting temperatures. Some salts may melt at relatively low temperatures, while other salts require several hundred degrees centigrade of heat before they melt. Mixed combinations of salts may melt at a lower temperature than individual salts will melt.
Some molten salts are very good conductors of heat, and some molten salts may not boil until they reach 1400 degrees centigrade, or even higher. These qualities make molten salts potentially excellent coolants for high temperature reactors. Research on high temperature fluoride salt cooled reactors continues at Oak Ridge National Laboratory (ORNL).
Two families of salts have been identified as potentially excellent reactor coolants. They are Fluoride and Chloride salts. Of these two salt families, ORNL scientists quickly chose the former, as presenting fewer developmental challenges while offering greater opportunities for commercial reactor use. Liquid chloride salts were identified as presenting more problems for reactor developers. Chlorine was, in particular, less useful as a neutron moderator than fluorine. However, Chloride salts were suitable for fast reactors, and a Molten Chloride Fast Breeder Reactor was both possible, and probably technologically less challenging than the Liquid Metal Fast Breeder Reactor, as a uranium fuel cycle breeder. In addition, the MCFBR would have offered far fewer safety problems than the LMFBR, while offering a route to technologically superior and lower cost fuel reprocessing.
ORNL Scientists however, also noted that liquid fluoride salt offered a superior performance precisely because of their neutron moderating performance. Moderation removes energy from neutrons. Heavily moderated neutrons are called thermal neutrons, and reactors that are built with large amounts of neutrons moderating materials such as graphite or heavy water are called thermal reactors. Neutron moderation decreases the amount of fissionable material required to sustain a chain reaction, and graphite and heavy water moderated reactors can produce chain reactions with natural uranium. While graphite and heavy water moderated reactors are useful tools for the production of plutonium, they are not particularly efficient tools for burning Pu-239 or even U-235 as nuclear fuels. However, it is possible to breed thorium in thermal reactors. Thermal reactor breeding of thorium offers significant advantages over fast reactor breeding of U-238. Thermal Reactors can operate with 10% perhaps as little as 3% of the fissionable material required to operate fast reactors. Thus thermal reactors can potentially be started at a far more rapid rate than fast reactors.
Fast reactor advocates claim that it is possible for fast reactors to breed at a much more rapid rate, but there appear to be safety problems involved in more rapid fast reactor breeding, and current documented advanced fast reactor designs appear to breed at a similar rate to thermal thorium breeding molten salt reactors.
The term breeding refers to the production more nuclear fuel than is used in a nuclear reactor. If a fertile material, either uranium 238 or thorium 232, is included with the fissionable material placed inside a reactor core along, the result will almost inevitably be the conversion of some of that fertile material into a fissionable material. In the case of thorium, a thorium 232 atom inside the reactor can absorb a neutron, and then becomes a thorium 233 atom. The nucleus, that is the center, of a thorium 233 atom is unstable. A neutron in the unstable Th 233 atom will eventually emit an electron, changing the neutron to a proton, the new proton in turn converts the atom to protactinium 233, and Pa 233 is also unstable. After a few days, a Pa neutron emits an electron, and as a consequence converts to a proton. The added proton makes the atom uranium 233. U 233 is fissionable, and can be used as reactor fuel.
The uranium fuel cycle is similar. If an U-238 atom absorbs a neutron a process that is similar to the process we find with thorium 233 occurs, and the U-239 atom is converted into plutonium 239.
Most reactors produce added nuclear fuel by converting U-238 into Pu-239. The amount of plutonium produced usually equals somewhere in the neighborhood of 60% of the amount of U-235 burned in a conventional reactor. Neither U-235 nor Pu-239 are ideal nuclear fuels at conventional neutron speed. WASH-1097 states
From a nuclear standpoint, the use of U-233 in a thermal reactor makes it possible to achieve higher fuel conversion ratios and longer fuel burnups than is practical with either U-235 or Pu-239. . .WASH-1097 defines the fuel conversion ratio,
The higher conversion ratios which can be obtained in thermal-spectrum reactors when using U- 233 instead of Pu-239 can result in a significantly better utilization of natural uranium fuel resources with thorium-fueled reactors than with the low-enrichment, light-water cooled uranium-fueled reactors . . .
The fuel conversion ratio (CR) is the ratio of the amount of fissile fuel produced per unit of fissile fuel destroyed. . .Breeding takes place when the conversion ration is greater than 1 to 1. While plutonium theoretically produces more neutrons and therefore faster breeding in fast reactors,
A higher breeding ratio can be obtained with Pu-239 than with U-233 in a very high-energy, fast- neutron spectrum reactor. On the other hand, in a degraded (10 to 100 keV) fast spectrum, U-233 would probably be as good as, or better than, Pu-239. Also, the variation of U-233 and Pu-239 cross sections with energy are such that improved reactivity coefficients would be obtained with the use of U-233 in a large sodium-cooled FBR. This leads to improved nuclear safety characteristics. . . .Thus not only is thorium breeding attractive in molten salt reactors, thorium breeding in fast reactors enhances their safety, and increases their conversion ratios. This point has not been lost on Indian reactor scientists who plan large scale production of thorium-uranium fast reactor breeding hybrids. Breeding thorium as a nuclear fuel in fast reactors would produce a large amount of fissionable U-233 that can be used as nuclear fuel in conventional reactors. This point has not been lost on Indian nuclear scientists, who plan to use the extra U-233 in Advanced Heavy Water Reactors. U-233 fueled non-breeder or converter Molten Salt Reactors offer attractive, safer and lower cost alternatives to conventional water cooled reactors, while reducing but not eliminating the nuclear waste problem. Thus, a modified Indian system could be developed, that would feature thorium breeding in IFRs, with U-233 burning MSRs. The rub for such a system would be that LFTRs would be cheaper to develop, cheaper to build, safer and cheaper to operate, and would virtually eliminate the nuclear waste problem.
The energy dependence of the fast-fission cross sections of Th-232 and U-238 is such that the use of Th-232 would produce an improved reactivity coefficient in a liquid-metal-cooled FBR. The fast fission cross-section of Th-232 is much lower than that of U-238 so that use of the latter leads to much larger conversion ratios in fast-spectrum reactors.
The LFTR then is a thorium breeding MSR, that offers what may well be the simplest and best solution to the problem of producing sustainable nuclear power.