Areas shown in quotes are taken from Wiki Pedia
Mothballed Technology Only a few MSRs have actually been built…India will have their first reactor online next year
Startup fuel Austrian reactor designer Dr. Walter Seifritz has proposed an 8GWe D-D PACER reactor that would breed 65,000 kg’s of U-233 per year which is enough to start 81 new 1GWe LFTRs per year
Or as the article states you could burn waste products from Nuclear reactors or material from decommissioned bombs. This provides us a way of getting rid of those nasty bombs!
As the wiki article also states
“After many years the plutonium burns out and U-233 is produced in place. At the end of the reactor fuel life, the spent fuel salt can be reprocessed to recover the bred U-233 to start up new LFTRs.”
“Because of the high heat capacity and considerable density of the buffer salt, the buffer salt not only prevents fuel salt freezing, but also participates in the passive decay heat cooling system, provides radiation shielding, and reduces deadweight stresses on primary loop components. This design could also be adopted for LFTRs.”
“It is possible to operate on lithium fluoride-thorium fluoride eutectic without beryllium, as the French LFTR design, the “TMSR”, has chosen”
The wiki article also states that beryllium is handled by other industries and could be applied to LFTR reactors for $0.12/MWh if it was necessary to use for some unknown reason.
Loss of Delayed Neutrons “This makes a fairly compact heat exchanger an important design requirement for an LFTR”
Waste management “For long-term storage, conversion to an insoluble form such as a glass, could be desirable.”
As you can see not a show stopper. The waste is only .1 % about the size of a coke can for 1GW of electricity as I said in a previous article.
Decommissioning costs are uncertain as the article states
“A GWe reactor plant produces over 300 billion kWh of electricity over a 40 year lifetime, so a 0.2 cent per kWh decommissioning fee delivers $600 million at the end of the plant lifetime.”
Noble Metal Buildup in Equipment “such as nickel wool sponge cartridges, will have to be developed to filter and trap the noble metals to prevent them from building up excessively in the piping, reprocessing plant, and heat exchanger over time.”
The solubility of plutonium is limited A thermal spectrum, lower power density core does not have issues with plutonium solubility. This would suggest using more small reactors instead of one larger reactor.
Potential proliferation risks from reprocessing. “If reprocessing of the salt mixture works well, this technology like any advance in reprocessing technology can pose a proliferation-risk. As an alternative method of reprocessing this could be used to separate plutonium from other reactors as well. However, as stated above, plutonium is chemically difficult to separate from thorium, and plutonium cannot be used in bombs if diluted in large amounts of thorium. In addition, the plutonium produced by the thorium fuel cycle is mostly Pu-238 which creates high levels of spontaneous neutrons and decay heat that greatly complicate fission bomb design and reduces the bomb yield drastically by pre-detonation due to Pu-238 starting the chain reaction before the fissile material in the bomb is fully assembled (causing a “fizzle” rather than a fission bomb explosion). Finally, all of the reprocessing steps involve automated, non-human handling in a fully closed and contained hot cell, this is due to the high temperatures and radiation levels, which makes the diversion of bomb material very difficult. Compared to today’s water solvent extraction methods such as PUREX, the pyroprocesses are inaccessible and produce impure fissile materials, often with large amounts of fission products contamination. This is not a problem for an automated reactor-hot cell system, but poses severe difficulties for would-be bomb makers.”
Proliferation risk from protactinium separation for some specific designs “Compact designs can breed only using rapid separation of protactinium, a potential proliferation-risk since this gives possible access to high purity 233-U without the U-232 contamination that is in the core. This is not easy, but still a possible new path to weapons-grade material. Because of this possibility, commercial power reactors may have to be designed without separation of protactinium. In practice, this means either not breeding or operating at a lower power density. For a two fluid design, this could simply operate with a bigger blanket, and keep the high power density core (which has no thorium and, therefore, no protactinium).”
The proliferation of Neptunium-237. “In designs utilizing a fluorinator, Np-237 comes out with uranium as gaseous hexafluoride and can be easily separated using solid fluoride pellet absorption beds. Theoretically, it should be possible to use Np-237 as fission bomb material. No one has ever successfully produced a bomb with this material, but it should be theoretically possible to use it, because of its considerable fast fission cross section and low critical mass. When the Np-237 is kept in the reactor, it will transmute to Pu-238, an extremely high-value fuel for space radioisotope thermal generators. A single gram is worth thousands of dollars. Pu-238 is itself an excellent proliferation deterrent, as explained earlier. Because of this, the Np-237 will likely be sent back to the reactor to be transmuted to Pu-238. In addition, it is possible to use vacuum distillation instead of fluorination, which does not separate neptunium at all. It should be noted, that all reactors, not just thorium reactors, produce considerable amounts of neptunium, which is always present in high (mono)isotopic quality, and it is just as easily extracted chemically. This is therefore not a distinguishing issue for LFTRs in particular. In fact, americium could also be theoretically used for nuclear weapons, and LFTRs do not produce meaningful quantities of americium, indeed they are one of the few reactor types that can burn existing stockpiles of americium and neptunium with high efficiency.”
Neutron poisoning and tritium production from lithium-6. “Lithium-6 is a strong neutron poison; using LiF with natural lithium, with its 7.5% Lithium-6 content, will not even allow the reactor to start up. The high neutron density in the core rapidly transmutes lithium-6 to tritium, losing precious neutrons that are required to sustain at least break-even breeding, and also producing tritium in the process. Tritium is a radioactive isotope of hydrogen, which is nearly identical, chemically, to ordinary hydrogen. In hot fluoride salts, hydrogen will be present as elemental tritium. In the MSR, the tritium is quite mobile because, in its elemental form, it rapidly diffuses through metals at high temperature. So some effort is needed to keep the emissions low. If the lithium is isotopically enriched in lithium-7, and the isotopic separation level is high enough (99.995% lithium-7), the amount of tritium produced is only a few hundred grams per year for a 1 GWe reactor. This much smaller amount of tritium comes mostly from the lithium-7 – tritium reaction and from beryllium, which can produce tritium indirectly by first transmuting to lithium-6, thus regenerating the tritium-producing, lithium-6. Because of these reasons, if an LFTR design uses a lithium salt, it uses the lithium-7 isotope, via enrichment of natural lithium, to reduce tritium formation. In the MSRE, tritium formation was also reduced by this removal of lithium-6 from the fuel salt via isotopic enrichment. Since lithium-7 is at least 16% heavier than lithium-6 and is the most common isotope of lithium, the lithium-6 is comparatively easy and inexpensive to extract from naturally occurring lithium. Vacuum distillation of lithium achieves efficiencies of up to 8% per stage and only requires heating of raw lithium in a vacuum chamber. The aforementioned methods of chemistry control and lithium isotopic separation worked in preventing hydrogen corrosion and excessive tritium generation in the MSRE. Practical MSRs also operate the salt under a blanket of dry inert gas, usually helium. Fortunately, in an LFTR, there is also a good chance to recover the tritium, since it is not diluted in a large amount of water as it is in CANDU reactors. Various methods exist to trap tritium, such as hydriding it to a stable metal hydride using titanium, oxidizing it to less mobile (but still volatile) forms such as by using sodium fluoroborate or molten nitrate salt as coolant loops, or trapping it in the turbine power cycle gas and off-gas using copper oxide pellets (p41). For future systems, ORNL developed a secondary loop coolant system that would chemically trap the few hundred grams of residual tritium to a less mobile form, so that it could be trapped and removed from the secondary coolant rather than diffusing into the turbine power cycle. This technique by itself would reduce tritium emissions to the actual environment to acceptable levels, ORNL calculated.”
Radiation damage to nickel alloys. “ORNL addressed this problem by adding 1-2% titanium or niobium to the Hastelloy N. This small titanium addition changed the internal structure of the alloy so that the helium produced inside it would not concentrate in specific areas but would be finely divided in it. This relieved the stress and allowed the new modified Hastelloy N to withstand considerable neutron flux. However, the maximum temperature is limited to about 650°C. A number of other alloys also showed promise. “
This seems to be the only open-ended problem remaining. Limiting temperature means a faster flow of coolant necessary with the current design. It would require some advancement in materials to fix this problem but is not a show stopper. This would be addressed in a couple of years during the Manhattan style project.
Long term fuel salt storage issues “For longer term storage, fluoride containing wastes could go through a vitrification process to be encased in insoluble borosilicate glass suitable for long-term disposal.“
Business Model Today’s solid fuelled reactor vendors make long-term revenues by making the profit on the fuel fabrication. Yeah, so what we are trying to save money NOT spend it! The long-term models of all other major energy production methods cost a lot more money to maintain. Spend the money on research and development instead on something like fusion or a million other projects.
Development of the power cycle. “Currently, molten nitrate salt steam generators are being used in concentrated solar thermal power plants such as Andasol in Spain. Such a nitrate salt loop and steam generator could also be used for a molten salt reactor, as an additional, third circulating loop, where it would allow the use of existing molten nitrate salt steam generator equipment, and as an added bonus effectively trap any tritium that diffuses through the primary and secondary heat exchange.”
There is also a different type of engine that could be used that increases the efficiency greatly. I forget what the engine is called but will add it here when I remember.