“Interestingly, thorium is much more common than uranium, uses a low pressure liquid fuel reactor, and the waste is minimal. It basically solves all the problems that we worry about with present nuclear power.”
Looks like there are few unsolved problems:
LFTRs are quite unlike today’s operating commercial power reactors. These differences create design difficulties and trade-offs:
Mothballed technology - Only a few MSRs have actually been built. Those experimental reactors were constructed more than 40 years ago. This leads some technologists[who?] to say that it is difficult to critically assess the concept.[citation needed]
Startup fuel - Unlike mined uranium, mined thorium does not have a fissile isotope. Thorium reactors breed fissile uranium-233 from thorium, but require a considerable amount of U-233 for initial start up. There is very little of this material available. This raises the problem of how to start the reactors in a reasonable time frame. One option is to produce U-233 in today’s solid fuelled reactors, then reprocess it out of the solid waste. A LFTR can also be started by other fissile isotopes, enriched uranium or plutonium from reactors or decommissioned bombs. For enriched uranium startup, high enrichment is needed. Decommissioned uranium bombs have enough enrichment, but not enough is available to start many LFTRs. It is difficult to separate plutonium fluoride from lanthanide fission products. One option for a two-fluid reactor is to operate with plutonium or enriched uranium in the fuel salt, breed U-233 in the blanket, and store it instead of returning it to the core. Instead, add plutonium or enriched uranium to continue the chain reaction, similar to today’s solid fuel reactors. When enough U-233 is bred, replace the fuel with new fuel, retaining the U-233 for other startups. A similar option exists for a single-fluid reactor operating as a converter. Such a reactor would not reprocess fuel while operating. Instead the reactor would start on plutonium with thorium as the fertile and add plutonium. The plutonium eventually burns out and U-233 is produced in situ. 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.[citation needed]
Salts freezing - Fluoride salt mixtures have melting points ranging from 300 to over 600 degrees Celsius. The salts, especially those with beryllium fluoride, are very viscous near their freezing point. This requires careful design and freeze protection in the containment and heat exchangers. Freezing must be prevented in normal operation, during transients, and during extended downtime. The primary loop salt contains the decay heat-generating fission products, which help to maintain the required temperature. For the MSBR, ORNL planned on keeping the entire reactor room (the hot cell) at high temperature. This avoided the need for individual electric heater lines on all piping and provided more even heating of the primary loop components.17 One “liquid oven” concept developed for molten salt-cooled, solid-fueled reactors employs a separate buffer salt pool containing the entire primary loop.[79] Because of the high heat capacity and considerable density of the buffer salt, the buffer salt prevents fuel salt freezing and 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.[citation needed]
Beryllium toxicity - The proposed salt mixture FLiBe, contains large amounts of beryllium, which is toxic to humans. The salt in the primary cooling loops must be isolated from workers and the environment to prevent beryllium poisoning. This is routinely done in industry.80 Based on this industrial experience, the added cost of beryllium safety is expected to cost only $0.12/MWh.80 After start up, the fission process in the primary fuel salt produces highly radioactive fission products with a high gamma and neutron radiation field. Effective containment is therefore a primary requirement. It is possible to operate instead using lithium fluoride-thorium fluoride eutectic without beryllium, as the French LFTR design, the “TMSR”, has chosen.[81] This comes at the cost of a somewhat higher melting point, but has the additional advantages of simplicity (avoiding BeF
2 in the reprocessing systems), increased solubility for plutonium-trifluoride, reduced tritium production (beryllium produces lithium-6, which in turn produces tritium) and improved heat transfer (BeF
2 increases the viscosity of the salt mixture). Alternative solvents such as the fluorides of sodium, rubidium and zirconium allow lower melting points at a tradeoff in breeding.[2]
Loss of delayed neutrons - In order to be predictably controlled, nuclear reactors rely on delayed neutrons. They require additional slowly-evolving neutrons from fission product decay to continue the chain reaction. Because the delayed neutrons evolve slowly, this makes the reactor very controllable. In a LFTR, the presence of fission products in the heat exchanger and piping means a portion of these delayed neutrons are also lost.[82] They do not participate in the core’s critical chain reaction, which in turn means the reactor behaves less gently during changes of flow, power, etc. Approximately up to half of the delayed neutrons can be lost. In practice, it means that the heat exchanger must be compact so that the volume outside the core is as small as possible. The more compact (higher power density) the core is, the more important this issue becomes. Having more fuel outside the core in the heat exchangers also means more of the expensive fissile fuel is needed to start the reactor. This makes a fairly compact heat exchanger an important design requirement for a LFTR.[citation needed]
Waste management - About 83% of the radioactive waste has a half-life in hours or days, with the remaining 17% requiring 300 year storage in geologically stable confinement to reach background levels.[71] Because some of the fission products, in their fluoride form, are highly water soluble, fluorides are less suited to long term storage. For example, cesium fluoride has a very high solubility in water. For long term storage, conversion to an insoluble form such as a glass, could be desirable.[citation needed]
Uncertain decommissioning costs - Cleanup of the Molten-Salt Reactor Experiment was about $130 million, for a small 8 MW(th) unit. Much of the high cost was caused by the unexpected evolution of fluorine and uranium hexafluoride from cold fuel salt in storage that ORNL did not defuel and store correctly, but this has now been taken into consideration in MSR design.[83] In addition, decommissioning costs don’t scale strongly with plant size based on previous experience,[84] and costs are incurred at the end of plant life, so a small per kilowatthour fee is sufficient. For example, a GWe reactor plant produces over 300 billion kWh of electricity over a 40-year lifetime, so a $0.001/kWh decommissioning fee delivers $300 million plus interest at the end of the plant lifetime.
Noble metal buildup - Some radioactive fission products, such as noble metals, deposit on pipes. Novel equipment, such as nickel-wool sponge cartridges, must be developed to filter and trap the noble metals to prevent build up.
Limited graphite lifetime - Compact designs have a limited lifetime for the graphite moderator and fuel / breeding loop separator. Under the influence of fast neutrons, the graphite first shrinks, then expands indefinitely until it becomes very weak and can crack, creating mechanical problems and causing the graphite to absorb enough fission products to poison the reaction.[85] The 1960 two-fluid design had an estimated graphite replacement period of four years.2 Eliminating graphite from sealed piping was a major incentive to switch to a single-fluid design.17 Replacing this large central part requires remotely operated equipment. MSR designs have to arrange for this replacement. In a molten salt reactor, virtually all of the fuel and fission products can be piped to a holding tank. Only a fraction of one percent of the fission products end up in the graphite, primarily due to fission products slamming into the graphite. This makes the graphite surface radioactive, and without recycling/removal of at least the surface layer, creates a fairly bulky waste stream. Removing the surface layer and recycling the remainder of the graphite would solve this issue.[original research?] Several techniques exist to recycle or dispose of nuclear moderator graphite.[86] Graphite is inert and immobile at low temperatures, so it can be readily stored or buried if required.[86] At least one design used graphite balls (pebbles) floating in salt, which could be removed and inspected continuously without shutting down the reactor.[87] Reducing power density increases graphite lifetime.88 By comparison, solid-fueled reactors typically replace 1/3 of the fuel elements, including all of the highly radioactive fission products therein, every 12 to 24 months. This is routinely done under a protecting and cooling column layer of water.
Graphite-caused positive reactivity feedback - When graphite heats up, it increases U-233 fission, causing an undesirable positive feedback.[44] The LFTR design must avoid certain combinations of graphite and salt and certain core geometries. If this problem is addressed by employing adequate graphite and thus a well-thermalized spectrum, it is difficult to reach break-even breeding.[44] The alternative of using little or no graphite results in a faster neutron spectrum. This requires a large fissile inventory and radiation damage increases.[44]
Limited plutonium solubility - Fluorides of plutonium, americium and curium occur as trifluorides, which means they have three fluorine atoms attached (PuF
3, AmF
3, CmF
3). Such trifluorides have a limited solubility in the FLiBe carrier salt. This complicates startup, especially for a compact design that uses a smaller primary salt inventory. Solubility can be increased by operating with less or no beryllium fluoride (which has no solubility for trifluorides) or by operating at a higher temperature[citation needed](as with most other liquids, solubility rises with temperature). A thermal spectrum, lower power density core does not have issues with plutonium solubility.
Proliferation risk from reprocessing - Effective reprocessing implies a proliferation-risk. LFTRs could be used to handle 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 produces high levels of spontaneous neutrons and decay heat that make it impossible to construct a fission bomb with this isotope alone, and extremely difficult to construct one containing even very small percentages of it. The heat production rate of 567 W/kg[89] means that a bomb core of this material would continuously produce several kilowatts of heat. The only cooling route is by conduction through the surrounding high explosive layers, which are poor conductors. This creates unmanageably high temperatures that would destroy the assembly. The spontaneous fission rate of 1204 kBq/g[89] is over twice that of Pu-240. Even very small percentages of this isotope would reduce bomb yield drastically by “predetonation” due to neutrons from spontaneous fission starting the chain reaction causing a “fizzle” rather than an explosion. Reprocessing itself involves automated handling in a fully closed and contained hot cell, which complicates diversion. Compared to today’s extraction methods such as PUREX, the pyroprocesses are inaccessible and produce impure fissile materials, often with large amounts of fission product contamination. While not a problem for an automated system, it poses severe difficulties for would-be proliferators.[citation needed]
Proliferation risk from protactinium separation - Compact designs can breed only using rapid separation of protactinium, a proliferation-risk, since this potentially gives access to high purity 233-U. This is difficult as the 233-U from these reactors will be contaminated with 232-U, a high gamma radiation emitter, requiring a protective hot enrichment facility[71] as a possible path to weapons-grade material. Because of this, commercial power reactors may have to be designed without separation. In practice, this means either not breeding, or operating at a lower power density. A two-fluid design might operate with a bigger blanket and keep the high power density core (which has no thorium and therefore no protactinium).[citation needed]
Proliferation of neptunium-237 - In designs utilizing a fluorinator, Np-237 appears with uranium as gaseous hexafluoride and can be easily separated using solid fluoride pellet absorption beds. No one has produced such a bomb, but Np-237’s considerable fast fission cross section and low critical mass imply the possibility.[90] When the Np-237 is kept in the reactor, it transmutes to Pu-238, a high value fuel for space radioisotope thermal generators.[91] A single gram is worth thousands of dollars. Pu-238 is itself an excellent proliferation deterrent. Because of this, Np-237 could be returned to the reactor and transmuted. Vacuum distillation does not separate neptunium. All reactors produce considerable neptunium, which is always present in high (mono)isotopic quality, and is easily extracted chemically.[90] However, this is a byproduct of traditional uranium fission, along with elements like plutonium, americium, and curium.[63] Np-239 or isotopes thereof are not necessarily produced in a LFTR.[92]
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, prevents reactors from starting. The high neutron density in the core rapidly transmutes lithium-6 to tritium, losing neutrons that are required to sustain break-even breeding. Tritium is a radioactive isotope of hydrogen, which is nearly identical, chemically, to ordinary hydrogen.[93] In the MSR the tritium is quite mobile because, in its elemental form, it rapidly diffuses through metals at high temperature. 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 tritium-producing lithium-6. LFTR designs that use a lithium salt, choose the lithium-7 isotope. In the MSRE, lithium-6 was successfully removed from the fuel salt via isotopic enrichment. Since lithium-7 is at least 16% heavier than lithium-6, and is the most common isotope, lithium-6 is comparatively easy and inexpensive to extract. Vacuum distillation of lithium achieves efficiencies of up to 8% per stage and requires only heating in a vacuum chamber.[94] However, about one fission in 90,000 produces helium-6, which quickly decays to lithium-6 and one fission in 12,500 produces an atom of tritium directly (in all reactor types). Practical MSRs operate under a blanket of dry inert gas, usually helium. LFTRs offer a good chance to recover the tritium, since it is not highly diluted in water as in CANDU reactors. Various methods exist to trap tritium, such as hydriding it to titanium,[95] oxidizing it to less mobile (but still volatile) forms such as sodium fluoroborate or molten nitrate salt, or trapping it in the turbine power cycle gas and offgasing it using copper oxide pellets.96 ORNL developed a secondary loop coolant system that would chemically trap residual tritium so that it could be removed from the secondary coolant rather than diffusing into the turbine power cycle. ORNL calculated that this would reduce Tritium emissions to acceptable levels.[93]
Corrosion from tellurium - The reactor makes small amounts of tellurium as a fission product. In the MSRE, this caused small amounts of corrosion at the grain boundaries of the special nickel alloy, Hastelloy-N. Metallurgical studies showed that adding 1 to 2% niobium to the Hastelloy-N alloy improves resistance to corrosion by tellurium.25 Maintaining the ratio of UF
4/UF
3 to less than 60 reduced corrosion by keeping the fuel salt slightly reducing. The MSRE continually contacted the flowing fuel salt with a beryllium metal rod submerged in a cage inside the pump bowl. This caused a fluorine shortage in the salt, reducing tellurium to a less aggressive (elemental) form. This method is also effective in reducing corrosion in general, because the fission process produces more fluorine atoms that would otherwise attack the structural metals.97
Radiation damage to nickel alloys - The standard Hastelloy N alloy was found to be embrittled by neutron radiation. Neutrons reacted with nickel to form helium. This helium gas concentrated at specific points inside the alloy, where it increased stresses. ORNL addressed this problem by adding 1-2% titanium or niobium to the Hastelloy N. This changed the alloy’s internal structure so that the helium would be finely distributed. This relieved the stress and allowed the alloy to withstand considerable neutron flux. However the maximum temperature is limited to about 650 °C.[98] Other alloys also showed promise.[99] The outer vessel wall that contains the salt can have neutronic shielding, such as boron carbide, to effectively protect it from neutron damage.[100]
Long term fuel salt storage - If the fluoride fuel salts are stored in solid form over many decades, radiation can cause the release of corrosive fluorine gas and uranium hexafluoride.[101] The salts must be defueled and wastes removed before extended shutdowns and stored above 100 degrees Celsius.[83] Fluorides are less suitable for long term storage because some have high water solubility unless vitrified in insoluble borosilicate glass.[102]
Business model - Today’s solid fuelled reactor vendors make long term revenues by fuel fabrication.[dubious – discuss] Without any fuel to fabricate and sell, a LFTR would adopt a different business model.
Development of the power cycle - Developing a large helium or supercritical carbon dioxide turbine is needed for highest efficiency. These gas cycles offer numerous potential advantages for use with molten salt-fueled or molten salt-cooled reactors.[103] These closed gas cycles face design challenges and engineering upscaling work for a commercial turbine-generator set.[104] A standard supercritical steam turbine could be used at a small penalty in efficiency (the net efficiency of the MSBR was designed to be approximately 44%, using an old 1970s steam turbine).[105] A molten salt to steam generator would still have to be developed. Currently, molten nitrate salt steam generators are used in concentrated solar thermal power plants such as Andasol in Spain. Such a generator could be used for an MSR as a third circulating loop, where it would also trap any tritium that diffuses through the primary and secondary heat exchanger.
And the Sierra Club would probably not be a supporter…
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