104-642: The Thorcon nuclear reactor is a design of a molten salt reactor with a graphite moderator, proposed by the US-based Thorcon company. These nuclear reactors are designed as part of a floating power plant, to be manufactured on an assembly line in a shipyard, and to be delivered via barge to any ocean or major waterway shoreline, similar to the US's MH-1A from 1968 and the Russian Akademik Lomonosov operating since 2020. The reactors are to be delivered as
208-508: A UREX ( UR anium EX traction) process which could be used to save space inside high level nuclear waste disposal sites, such as the Yucca Mountain nuclear waste repository , by removing the uranium which makes up the vast majority of the mass and volume of used fuel and recycling it as reprocessed uranium . The UREX process is a PUREX process which has been modified to prevent the plutonium from being extracted. This can be done by adding
312-423: A breeder reactor . Increased research into Generation IV reactor designs renewed interest in the 21st century with multiple nations starting projects. As of June 2023, China has been operating its TMSR-LF1 thorium unit. MSRs eliminate the nuclear meltdown scenario present in water-cooled reactors because the fuel mixture is kept in a molten state. The fuel mixture is designed to drain without pumping from
416-468: A byproduct. Because this could allow for weapons grade nuclear material , nuclear reprocessing is a concern for nuclear proliferation and is thus tightly regulated. Relatively high cost is associated with spent fuel reprocessing compared to the once-through fuel cycle, but fuel use can be increased and waste volumes decreased. Nuclear fuel reprocessing is performed routinely in Europe, Russia, and Japan. In
520-838: A chamber full of fluorine. This is known as flame fluorination; the heat produced helps the reaction proceed. Most of the uranium , which makes up the bulk of the fuel, is converted to uranium hexafluoride , the form of uranium used in uranium enrichment , which has a very low boiling point. Technetium , the main long-lived fission product , is also efficiently converted to its volatile hexafluoride. A few other elements also form similarly volatile hexafluorides, pentafluorides, or heptafluorides. The volatile fluorides can be separated from excess fluorine by condensation, then separated from each other by fractional distillation or selective reduction . Uranium hexafluoride and technetium hexafluoride have very similar boiling points and vapor pressures, which makes complete separation more difficult. Many of
624-403: A closed fuel cycle—as opposed to the once-through fuel currently used in conventional nuclear power generators. MSRs exploit a negative temperature coefficient of reactivity and a large allowable temperature rise to prevent criticality accidents . For designs with the fuel in the salt, the salt thermally expands immediately with power excursions. In conventional reactors the negative reactivity
728-691: A helium-cooled VHTR operating in similar conditions; passive safety systems and better retention of fission products in the event of an accident. Reactors containing molten thorium salt, called liquid fluoride thorium reactors (LFTR), would tap the thorium fuel cycle . Private companies from Japan, Russia, Australia and the United States, and the Chinese government, have expressed interest in developing this technology. Advocates estimate that five hundred metric tons of thorium could supply U.S. energy needs for one year. The U.S. Geological Survey estimates that
832-409: A large variety of fuel compositions due to its on-line fuel control and flexible fuel processing. The standard MSFR would be a 3000 MWth reactor that has a total fuel salt volume of 18 m with a mean fuel temperature of 750 °C. The core's shape is a compact cylinder with a height to diameter ratio of 1 where liquid fluoride fuel salt flows from the bottom to the top. The return circulation of
936-476: A method for removing zirconium fuel cladding, instead of mechanical decladding. Chlorides are likely to be easier than fluorides to later convert back to other compounds, such as oxides. Chlorides remaining after volatilization may also be separated by solubility in water. Chlorides of alkaline elements like americium , curium , lanthanides , strontium , caesium are more soluble than those of uranium , neptunium , plutonium , and zirconium . To determine
1040-474: A moderator (usually graphite) to slow the neutrons down and moderate temperature. They can accept a variety of fuels (low-enriched uranium, thorium, depleted uranium , waste products) and coolants (fluoride, chloride, lithium, beryllium, mixed). Fuel cycle can be either closed or once-through. They can be monolithic or modular, large or small. The reactor can adopt a loop, modular or integral configuration. Variations include: The molten-salt fast reactor (MSFR)
1144-516: A neutron driven nuclear reaction. To date the extraction system for the SANEX process has not been defined, but currently several different research groups are working towards a process. For instance the French CEA is working on a bis-triazinyl pyridine (BTP) based process. Other systems such as the dithiophosphinic acids are being worked on by some other workers. The UN iversal EX traction process
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#17330863300391248-470: A neutron to become Cl , then degrades by beta decay to S . Lithium must be in the form of purified Li , because Li effectively captures neutrons and produces tritium . Even if pure Li is used, salts containing lithium cause significant tritium production, comparable with heavy water reactors. Reactor salts are usually close to eutectic mixtures to reduce their melting point. A low melting point simplifies melting
1352-636: A peak temperature of 860 °C. It produced 100 MWh over nine days in 1954. This experiment used Inconel 600 alloy for the metal structure and piping. An MSR was operated at the Critical Experiments Facility of the Oak Ridge National Laboratory in 1957. It was part of the circulating-fuel reactor program of the Pratt & Whitney Aircraft Company (PWAC). This was called Pratt and Whitney Aircraft Reactor-1 (PWAR-1). The experiment
1456-518: A plutonium reductant before the first metal extraction step. In the UREX process, ~99.9% of the uranium and >95% of technetium are separated from each other and the other fission products and actinides . The key is the addition of acetohydroxamic acid (AHA) to the extraction and scrub sections of the process. The addition of AHA greatly diminishes the extractability of plutonium and neptunium , providing somewhat greater proliferation resistance than with
1560-730: A sealed unit and never opened on site. All reactor maintenance and fuel processing is done at an off-site location. As of 2022, no reactor of this type has been built. A prototype of 500 MW (TMSR-500) output should be activated in Indonesia by 2029. ThorCon has proposed a power station closely based on the Molten-Salt Reactor Experiment in the 1960s, claiming that its design requires no new technology. The power station would contain two 250 MWe small modular reactors . The replaceable reactors are to be removed and replaced every four years. As molten salt reactors , they are designed for
1664-413: A slow-decaying isotope between them which facilitates neutron absorption by Cl . Chlorides permit fast breeder reactors to be constructed. Much less research has been done on reactor designs using chloride salts. Chlorine, unlike fluorine, must be purified to isolate the heavier stable isotope, Cl , thus reducing production of sulfur tetrachloride that occurs when Cl absorbs
1768-928: A smaller plant at West Valley Reprocessing Plant which closed by 1972 because of its inability to meet new regulatory requirements. Reprocessing of civilian fuel has long been employed at the COGEMA La Hague site in France, the Sellafield site in the United Kingdom, the Mayak Chemical Combine in Russia, and at sites such as the Tokai plant in Japan, the Tarapur plant in India, and briefly at
1872-446: A solid aluminium cathode. As an alternative to electrowinning, the wanted metal can be isolated by using a molten alloy of an electropositive metal and a less reactive metal. Since the majority of the long term radioactivity , and volume, of spent fuel comes from actinides, removing the actinides produces waste that is more compact, and not nearly as dangerous over the long term. The radioactivity of this waste will then drop to
1976-482: A vacuum. If a carrier salt like lithium fluoride or sodium fluoride is being used as a solvent, high-temperature distillation is a way to separate the carrier salt for reuse. Molten salt reactor designs carry out fluoride volatility reprocessing continuously or at frequent intervals. The goal is to return actinides to the molten fuel mixture for eventual fission, while removing fission products that are neutron poisons , or that can be more securely stored outside
2080-501: Is a generic term for high-temperature methods. Solvents are molten salts (e.g. LiCl + KCl or LiF + CaF 2 ) and molten metals (e.g. cadmium, bismuth, magnesium) rather than water and organic compounds. Electrorefining , distillation , and solvent-solvent extraction are common steps. These processes are not currently in significant use worldwide, but they have been pioneered at Argonne National Laboratory with current research also taking place at CRIEPI in Japan,
2184-607: Is a proposed design with the fuel dissolved in a fluoride salt coolant. The MSFR is one of the two variants of MSRs selected by the Generation IV International Forum (GIF) for further development, the other being the FHR or AHTR. The MSFR is based on a fast neutron spectrum and is believed to be a long-term substitute to solid-fueled fast reactors. They have been studied for almost a decade, mainly by calculations and determination of basic physical and chemical properties in
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#17330863300392288-576: Is also a proposed Generation IV molten-salt reactor variant regarded promising for the long-term future. The FHR/AHTR reactor uses a solid-fuel system along with a molten fluoride salt as coolant. One version of the Very-high-temperature reactor (VHTR) under study was the liquid-salt very-high-temperature reactor (LS-VHTR). It uses liquid salt as a coolant in the primary loop, rather than a single helium loop. It relies on " TRISO " fuel dispersed in graphite. Early AHTR research focused on graphite in
2392-459: Is an obsolete process that adds significant unnecessary material to the final radioactive waste. The bismuth phosphate process has been replaced by solvent extraction processes. The bismuth phosphate process was designed to extract plutonium from aluminium-clad nuclear fuel rods , containing uranium. The fuel was decladded by boiling it in caustic soda . After decladding, the uranium metal was dissolved in nitric acid . The plutonium at this point
2496-424: Is applied, causing the uranium metal (or sometimes oxide, depending on the spent fuel) to plate out on a solid metal cathode while the other actinides (and the rare earths) can be absorbed into a liquid cadmium cathode. Many of the fission products (such as caesium , zirconium and strontium ) remain in the salt. As alternatives to the molten cadmium electrode it is possible to use a molten bismuth cathode, or
2600-461: Is decreased. Most of the plutonium and some of the uranium will initially remain in ash which drops to the bottom of the flame fluorinator. The plutonium-uranium ratio in the ash may even approximate the composition needed for fast neutron reactor fuel. Further fluorination of the ash can remove all the uranium, neptunium , and plutonium as volatile fluorides; however, some other minor actinides may not form volatile fluorides and instead remain with
2704-670: Is delayed since the heat from the fuel must be transferred to the moderator. An additional method is to place a separate, passively cooled container below the reactor. Fuel drains into the container during malfunctions or maintenance, which stops the reaction. The temperatures of some designs are high enough to produce process heat, which led them to be included on the GEN-IV roadmap. MSRs offer many potential advantages over light water reactors: MSRs can be cooled in various ways, including using molten salts. Molten-salt-cooled solid-fuel reactors are variously called "molten-salt reactor system" in
2808-463: Is free from uranium and contains recovered transuranics in an inert matrix such as metallic zirconium . In the PYRO-B processing of such fuel, an electrorefining step is used to separate the residual transuranic elements from the fission products and recycle the transuranics to the reactor for fissioning. Newly generated technetium and iodine are extracted for incorporation into transmutation targets, and
2912-466: Is in the +4 oxidation state. It was then precipitated out of the solution by the addition of bismuth nitrate and phosphoric acid to form the bismuth phosphate. The plutonium was coprecipitated with this. The supernatant liquid (containing many of the fission products ) was separated from the solid. The precipitate was then dissolved in nitric acid before the addition of an oxidant (such as potassium permanganate ) to produce PuO 2 . The plutonium
3016-452: Is no longer a problem. This results in a breeder reactor with a fast neutron spectrum that operates in the Thorium fuel cycle. MSFRs contain relatively small initial inventories of U . MSFRs run on liquid fuel with no solid matter inside the core. This leads to the possibility of reaching specific power that is much higher than reactors using solid fuel. The heat produced goes directly into
3120-543: Is not absorbed but instead knocks a neutron out of the nucleus), generate U . Because U has a short half-life and its decay chain contains hard gamma emitters, it makes the isotopic mix of uranium less attractive for bomb-making. This benefit would come with the added expense of a larger fissile inventory or a 2-fluid design with a large quantity of blanket salt. The necessary fuel salt reprocessing technology has been demonstrated, but only at laboratory scale. A prerequisite to full-scale commercial reactor design
3224-541: Is not created in the core (as is present in boiling water reactors ), and no large, expensive steel pressure vessel (as required for pressurized water reactors ). Since it can operate at high temperatures, the conversion of the heat to electricity can use an efficient, lightweight Brayton cycle gas turbine. Much of the current research on FHRs is focused on small, compact heat exchangers that reduce molten salt volumes and associated costs. Molten salts can be highly corrosive and corrosivity increases with temperature. For
ThorCon nuclear reactor - Misplaced Pages Continue
3328-452: Is substantially different from the usual uranium or mixed uranium-plutonium oxides (MOX) that most current reactors were designed to use. Another pyrochemical process, the PYRO-B process, has been developed for the processing and recycling of fuel from a transmuter reactor ( a fast breeder reactor designed to convert transuranic nuclear waste into fission products ). A typical transmuter fuel
3432-517: Is that by lowering the alpha activity of the waste, the majority of the waste can then be disposed of with greater ease. In common with PUREX this process operates by a solvation mechanism. As an alternative to TRUEX, an extraction process using a malondiamide has been devised. The DIAMEX ( DIAM ide EX traction) process has the advantage of avoiding the formation of organic waste which contains elements other than carbon , hydrogen , nitrogen , and oxygen . Such an organic waste can be burned without
3536-455: Is the R&;D to engineer an economically competitive fuel salt cleaning system. Reprocessing refers to the chemical separation of fissionable uranium and plutonium from spent fuel. Such recovery could increase the risk of nuclear proliferation . In the United States the regulatory regime has varied dramatically across administrations. A systematic literature review from 2020 concludes that there
3640-418: Is the lead-cooled, salt-fueled reactor. MSR research started with the U.S. Aircraft Reactor Experiment (ARE) in support of the U.S. Aircraft Nuclear Propulsion program. ARE was a 2.5 MW th nuclear reactor experiment designed to attain a high energy density for use as an engine in a nuclear-powered bomber. The project included experiments, including high temperature and engine tests collectively called
3744-434: Is triggered and achieved by redundant and reliable devices such as detection and opening technology. During operation, the fuel salt circulation speed can be adjusted by controlling the power of the pumps in each sector. The intermediate fluid circulation speed can be adjusted by controlling the power of the intermediate circuit pumps. The temperature of the intermediate fluid in the intermediate exchangers can be managed through
3848-449: Is very limited information on economics and finance of MSRs, with low quality of the information and that cost estimations are uncertain. In the specific case of the stable salt reactor (SSR) where the radioactive fuel is contained as a molten salt within fuel pins and the primary circuit is not radioactive, operating costs are likely to be lower. While many design variants have been proposed, there are three main categories regarding
3952-450: Is very wide, but all agreed that under then-current economic conditions the reprocessing-recycle option is the more costly one. While the uranium market - particularly its short term fluctuations - has only a minor impact on the cost of electricity from nuclear power, long-term trends in the uranium market do significantly affect the economics of nuclear reprocessing. If uranium prices were to rise and remain consistently high, "stretching
4056-492: The West Valley Reprocessing Plant in the United States. In October 1976, concern of nuclear weapons proliferation (especially after India demonstrated nuclear weapons capabilities using reprocessing technology) led President Gerald Ford to issue a Presidential directive to indefinitely suspend the commercial reprocessing and recycling of plutonium in the U.S. On 7 April 1977, President Jimmy Carter banned
4160-576: The bismuth phosphate process , was developed and tested at the Oak Ridge National Laboratory (ORNL) between 1943 and 1945 to produce quantities of plutonium for evaluation and use in the US weapons programs . ORNL produced the first macroscopic quantities (grams) of separated plutonium with these processes. The bismuth phosphate process was first operated on a large scale at the Hanford Site , in
4264-408: The corrosivity of hot salts and the changing chemical composition of the salt as it is transmuted by the neutron flux . MSRs, especially those with fuel in the molten salt, offer lower operating pressures, and higher temperatures. In this respect an MSR is more similar to a liquid metal cooled reactor than to a conventional light water cooled reactor. MSR designs are often breeding reactors with
ThorCon nuclear reactor - Misplaced Pages Continue
4368-457: The diluent is a polar aromatic such as nitrobenzene . Other diluents such as meta -nitrobenzotri fluoride and phenyl trifluoromethyl sulfone have been suggested as well. An exotic method using electrochemistry and ion exchange in ammonium carbonate has been reported. Other methods for the extraction of uranium using ion exchange in alkaline carbonate and "fumed" lead oxide have also been reported. The bismuth phosphate process
4472-424: The fission products volatilized are the same ones volatilized in non-fluorinated, higher-temperature volatilization, such as iodine , tellurium and molybdenum ; notable differences are that technetium is volatilized, but caesium is not. Some transuranium elements such as plutonium , neptunium and americium can form volatile fluorides, but these compounds are not stable when the fluorine partial pressure
4576-530: The Energy Innovation Reform Project looked at the ThorCon and concluded that "if power plants featuring these technologies are able to produce electricity at the average LCOE price projected here (much less the low-end estimate), it would have a significant impact on electricity markets." Molten salt reactor A molten-salt reactor ( MSR ) is a class of nuclear fission reactor in which
4680-497: The European Union and Russian Federation. A MSFR is regarded sustainable because there are no fuel shortages. Operation of a MSFR does in theory not generate or require large amounts of transuranic (TRU) elements . When steady state is achieved in a MSFR, there is no longer a need for uranium enrichment facilities. MSFRs may be breeder reactors . They operate without a moderator in the core such as graphite, so graphite life-span
4784-542: The Generation IV proposal, molten-salt converter reactors (MSCR), advanced high-temperature reactors (AHTRs), or fluoride high-temperature reactors (FHR, preferred DOE designation). FHRs cannot reprocess fuel easily and have fuel rods that need to be fabricated and validated, requiring up to twenty years from project inception. FHR retains the safety and cost advantages of a low-pressure, high-temperature coolant, also shared by liquid metal cooled reactors . Notably, steam
4888-631: The Heat Transfer Reactor Experiments: HTRE-1, HTRE-2 and HTRE-3 at the National Reactor Test Station (now Idaho National Laboratory ) as well as an experimental high-temperature molten-salt reactor at Oak Ridge National Laboratory – the ARE. ARE used molten fluoride salt NaF/ZrF 4 /UF 4 (53-41-6 mol% ) as fuel, moderated by beryllium oxide (BeO). Liquid sodium was a secondary coolant. The experiment had
4992-573: The Molten-Salt Reactor Experiment (MSRE). MSRE was a 7.4 MW th test reactor simulating the neutronic "kernel" of a type of epithermal thorium molten salt breeder reactor called the liquid fluoride thorium reactor (LFTR). The large (expensive) breeding blanket of thorium salt was omitted in favor of neutron measurements. Nuclear reprocessing Nuclear reprocessing is the chemical separation of fission products and actinides from spent nuclear fuel . Originally, reprocessing
5096-549: The New York Times reported "...11 years after the government awarded a construction contract, the cost of the project has soared to nearly $ 5 billion. The vast concrete and steel structure is a half-finished hulk, and the government has yet to find a single customer, despite offers of lucrative subsidies." TVA (currently the most likely customer) said in April 2011 that it would delay a decision until it could see how MOX fuel performed in
5200-585: The Nuclear Research Institute of Řež in Czech Republic, Indira Gandhi Centre for Atomic Research in India and KAERI in South Korea. The electrolysis methods are based on the difference in the standard potentials of uranium, plutonium and minor actinides in a molten salt. The standard potential of uranium is the lowest, therefore when a potential is applied, the uranium will be reduced at
5304-472: The PUREX process, there have been efforts to develop alternatives to the process, some of them compatible with PUREX (i.e. the residue from one process could be used as feedstock for the other) and others wholly incompatible. None of these have (as of the 2020s) reached widespread commercial use, but some have seen large scale tests or firm commitments towards their future larger scale implementation. Pyroprocessing
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#17330863300395408-466: The United States, the Obama administration stepped back from President Bush's plans for commercial-scale reprocessing and reverted to a program focused on reprocessing-related scientific research. Not all nuclear fuel requires reprocessing; a breeder reactor is not restricted to using recycled plutonium and uranium. It can employ all the actinides , closing the nuclear fuel cycle and potentially multiplying
5512-507: The alkaline fission products. Some noble metals may not form fluorides at all, but remain in metallic form; however ruthenium hexafluoride is relatively stable and volatile. Distillation of the residue at higher temperatures can separate lower-boiling transition metal fluorides and alkali metal (Cs, Rb) fluorides from higher-boiling lanthanide and alkaline earth metal (Sr, Ba) and yttrium fluorides. The temperatures involved are much higher, but can be lowered somewhat by distilling in
5616-582: The ban in 1981, but did not provide the substantial subsidy that would have been necessary to start up commercial reprocessing. In March 1999, the U.S. Department of Energy (DOE) reversed its policy and signed a contract with a consortium of Duke Energy , COGEMA , and Stone & Webster (DCS) to design and operate a mixed oxide (MOX) fuel fabrication facility. Site preparation at the Savannah River Site (South Carolina) began in October 2005. In 2011
5720-501: The cathode out of the molten salt solution before the other elements. These processes were developed by Argonne National Laboratory and used in the Integral Fast Reactor project. PYRO-A is a means of separating actinides (elements within the actinide family, generally heavier than U-235) from non-actinides. The spent fuel is placed in an anode basket which is immersed in a molten salt electrolyte. An electric current
5824-454: The core to a containment vessel in emergency scenarios, where the fuel solidifies, quenching the reaction. In addition, hydrogen evolution does not occur. This eliminates the risk of hydrogen explosions (as in the Fukushima nuclear disaster ). They operate at or close to atmospheric pressure , rather than the 75–150 times atmospheric pressure of a typical light-water reactor (LWR). This reduces
5928-401: The deposition of solid particles in reactor operation. Sulfur must be removed because of its corrosive attack on nickel-based alloys at operational temperature. Structural metals such as chromium, nickel, and iron must be removed for corrosion control. A water content reduction purification stage using HF and helium sweep gas was specified to run at 400 °C. Oxide and sulfur contamination in
6032-437: The design to prevent its release into the environment. Many other salts can cause plumbing corrosion, especially if the reactor is hot enough to make highly reactive hydrogen. To date, most research has focused on FLiBe, because lithium and beryllium are reasonably effective moderators and form a eutectic salt mixture with a lower melting point than each of the constituent salts. Beryllium also performs neutron doubling, improving
6136-404: The disadvantage of requiring the use of a salting-out reagent (aluminium nitrate ) to increase the nitrate concentration in the aqueous phase to obtain a reasonable distribution ratio. This process was used at Windscale in 1951-1964. This process has been replaced by PUREX, which was shown to be a superior technology for larger scale reprocessing. The sodium uranyl acetate process was used by
6240-681: The distribution of radioactive metals for analytical purposes, Solvent Impregnated Resins (SIRs) can be used. SIRs are porous particles, which contain an extractant inside their pores. This approach avoids the liquid-liquid separation step required in conventional liquid-liquid extraction . For the preparation of SIRs for radioanalytical separations, organic Amberlite XAD-4 or XAD-7 can be used. Possible extractants are e.g. trihexyltetradecylphosphonium chloride(CYPHOS IL-101) or N,N0-dialkyl-N,N0-diphenylpyridine-2,6-dicarboxyamides (R-PDA; R = butyl, octy I, decyl, dodecyl). The relative economics of reprocessing-waste disposal and interim storage-direct disposal
6344-557: The early Soviet nuclear industry to recover plutonium from irradiated fuel. It was never used in the West; the idea is to dissolve the fuel in nitric acid , alter the oxidation state of the plutonium, and then add acetic acid and base. This would convert the uranium and plutonium into a solid acetate salt. Explosion of the crystallized acetates-nitrates in a non-cooled waste tank caused the Kyshtym disaster in 1957. As there are some downsides to
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#17330863300396448-424: The energy extracted from natural uranium by about 60 times. The potentially useful components dealt with in nuclear reprocessing comprise specific actinides (plutonium, uranium, and some minor actinides ). The lighter elements components include fission products , activation products , and cladding . The first large-scale nuclear reactors were built during World War II . These reactors were designed for
6552-476: The form of graphite rods that would be inserted in hexagonal moderating graphite blocks, but current studies focus primarily on pebble-type fuel. The LS-VHTR can work at very high temperatures (the boiling point of most molten salt candidates is >1400 °C); low-pressure cooling that can be used to match hydrogen production facility conditions (most thermochemical cycles require temperatures in excess of 750 °C); better electric conversion efficiency than
6656-445: The formation of acidic gases which could contribute to acid rain (although the acidic gases could be recovered by a scrubber). The DIAMEX process is being worked on in Europe by the French CEA . The process is sufficiently mature that an industrial plant could be constructed with the existing knowledge of the process. In common with PUREX this process operates by a solvation mechanism. S elective A cti N ide EX traction. As part of
6760-414: The fuel or increases its surface area to enhance penetration of reagents in following reprocessing steps. Simply heating spent oxide fuel in an inert atmosphere or vacuum at a temperature between 700 °C (1,292 °F) and 1,000 °C (1,830 °F) as a first reprocessing step can remove several volatile elements, including caesium whose isotope caesium-137 emits about half of the heat produced by
6864-440: The fuel salt. Usually the fuel salt temperature can be brought down by 100 °C using a 3% proportion of bubbles. MSFRs have two draining modes, controlled routine draining and emergency draining. During controlled routine draining, fuel salt is transferred to actively cooled storage tanks. The fuel temperature can be lowered before draining, this may slow down the process. This type of draining could be done every 1 to 5 years when
6968-472: The heat transfer fluid. In the MSFR, a small amount of molten salt is set aside to be processed for fission product removal and then returned to the reactor. This gives MSFRs the capability of reprocessing the fuel without stopping the reactor. This is very different compared to solid-fueled reactors because they have separate facilities to produce the solid fuel and process spent nuclear fuel. The MSFR can operate using
7072-490: The heavy water CANDU or the Atucha-class PHWRs, light water cooled graphite moderated RBMK , and British-built gas-cooled reactors such as Magnox , AGR ). MSR operating temperatures are around 700 °C (1,292 °F), significantly higher than traditional LWRs at around 300 °C (572 °F). This increases electricity-generation efficiency and process-heat opportunities. Relevant design challenges include
7176-590: The industry at present. When used on fuel from commercial power reactors the plutonium extracted typically contains too much Pu-240 to be considered "weapons-grade" plutonium, ideal for use in a nuclear weapon. Nevertheless, highly reliable nuclear weapons can be built at all levels of technical sophistication using reactor-grade plutonium. Moreover, reactors that are capable of refueling frequently can be used to produce weapon-grade plutonium, which can later be recovered using PUREX. Because of this, PUREX chemicals are monitored. The PUREX process can be modified to make
7280-428: The intermediate product protactinium Pa would be removed from the reactor and allowed to decay into highly pure U , an attractive bomb-making material. More modern designs propose to use a lower specific power or a separate thorium breeding blanket. This dilutes the protactinium to such an extent that few protactinium atoms absorb a second neutron or, via a (n, 2n) reaction (in which an incident neutron
7384-460: The inventory of fission products, control corrosion and improve neutron economy by removing fission products with high neutron absorption cross-section, especially xenon . This makes the MSR particularly suited to the neutron-poor thorium fuel cycle . Online fuel processing can introduce risks of fuel processing accidents, which can trigger release of radio isotopes . In some thorium breeding scenarios,
7488-523: The largest-known U.S. thorium deposit, the Lemhi Pass district on the Montana - Idaho border, contains thorium reserves of 64,000 metric tons. Traditionally, these reactors were known as molten salt breeder reactors (MSBRs) or thorium molten-salt reactors (TMSRs), but the name LFTR was promoted as a rebrand in the early 2000s by Kirk Sorensen. The stable salt reactor is a relatively recent concept which holds
7592-432: The later part of 1944. It was successful for plutonium separation in the emergency situation existing then, but it had a significant weakness: the inability to recover uranium. The first successful solvent extraction process for the recovery of pure uranium and plutonium was developed at ORNL in 1949. The PUREX process is the current method of extraction. Separation plants were also constructed at Savannah River Site and
7696-517: The level of various naturally occurring minerals and ores within a few hundred, rather than thousands of, years. The mixed actinides produced by pyrometallic processing can be used again as nuclear fuel, as they are virtually all either fissile , or fertile , though many of these materials would require a fast breeder reactor to be burned efficiently. In a thermal neutron spectrum, the concentrations of several heavy actinides ( curium -242 and plutonium-240 ) can become quite high, creating fuel that
7800-403: The management of minor actinides it has been proposed that the lanthanides and trivalent minor actinides should be removed from the PUREX raffinate by a process such as DIAMEX or TRUEX. To allow the actinides such as americium to be either reused in industrial sources or used as fuel, the lanthanides must be removed. The lanthanides have large neutron cross sections and hence they would poison
7904-409: The molten salt fuel statically in traditional LWR fuel pins. Pumping of the fuel salt, and all the corrosion/deposition/maintenance/containment issues arising from circulating a highly radioactive, hot and chemically complex fluid, are no longer required. The fuel pins are immersed in a separate, non-fissionable fluoride salt which acts as primary coolant. A prototypical example of a dual fluid reactor
8008-484: The need and cost for reactor pressure vessels . The gaseous fission products ( Xe and Kr ) have little solubility in the fuel salt, and can be safely captured as they bubble out of the fuel, rather than increasing the pressure inside the fuel tubes , as happens in conventional reactors. MSRs can be refueled while operating (essentially online- nuclear reprocessing ) while conventional reactors shut down for refueling (notable exceptions include pressure tube reactors like
8112-451: The neutron economy. This process occurs when the beryllium nucleus emits two neutrons after absorbing a single neutron. For the fuel carrying salts, generally 1% or 2% (by mole ) of UF 4 is added. Thorium and plutonium fluorides have also been used. Techniques for preparing and handling molten salt were first developed at ORNL. The purpose of salt purification is to eliminate oxides, sulfur and metal impurities. Oxides could result in
8216-482: The nitrate concentration in the aqueous phase to obtain a reasonable distribution ratio (D value). Also, hexone is degraded by concentrated nitric acid. This process was used in 1952-1956 on the Hanford plant T and has been replaced by the PUREX process. Pu + 4NO − 3 + 2S → [Pu(NO 3 ) 4 S 2 ] A process based on a solvation extraction process using the triether extractant named above. This process has
8320-413: The nuclear accident at Fukushima Daiichi . PUREX , the current standard method, is an acronym standing for P lutonium and U ranium R ecovery by EX traction . The PUREX process is a liquid-liquid extraction method used to reprocess spent nuclear fuel , to extract uranium and plutonium , independent of each other, from the fission products. This is the most developed and widely used process in
8424-431: The other fission products are sent to waste. Voloxidation (for volumetric oxidation ) involves heating oxide fuel with oxygen, sometimes with alternating oxidation and reduction, or alternating oxidation by ozone to uranium trioxide with decomposition by heating back to triuranium octoxide . A major purpose is to capture tritium as tritiated water vapor before further processing where it would be difficult to retain
8528-477: The plutonium extraction stage of the PUREX process. Adding a second extraction agent, octyl(phenyl)-N, N-dibutyl carbamoylmethyl phosphine oxide (CMPO) in combination with tributylphosphate, (TBP), the PUREX process can be turned into the TRUEX ( TR ans U ranic EX traction) process. TRUEX was invented in the US by Argonne National Laboratory and is designed to remove the transuranic metals (Am/Cm) from waste. The idea
8632-448: The plutonium. Addition of an alkali produced an oxide. The combined lanthanum plutonium oxide was collected and extracted with nitric acid to form plutonium nitrate. This is a liquid-liquid extraction process which uses methyl isobutyl ketone codenamed hexone as the extractant. The extraction is by a solvation mechanism. This process has the disadvantage of requiring the use of a salting-out reagent ( aluminium nitrate ) to increase
8736-477: The primary nuclear reactor coolant and/or the fuel is a mixture of molten salt with a fissile material. Two research MSRs operated in the United States in the mid-20th century. The 1950s Aircraft Reactor Experiment (ARE) was primarily motivated by the technology's compact size, while the 1960s Molten-Salt Reactor Experiment (MSRE) aimed to demonstrate a nuclear power plant using a thorium fuel cycle in
8840-671: The primary cooling loop, a material is needed that can withstand corrosion at high temperatures and intense radiation . Experiments show that Hastelloy-N and similar alloys are suited to these tasks at operating temperatures up to about 700 °C. However, operating experience is limited. Still higher operating temperatures are desirable—at 850 °C (1,560 °F) thermochemical production of hydrogen becomes possible. Materials for this temperature range have not been validated, though carbon composites, molybdenum alloys (e.g. TZM), carbides , and refractory metal based or ODS alloys might be feasible. The salt mixtures are chosen to make
8944-413: The production of plutonium for use in nuclear weapons . The only reprocessing required, therefore, was the extraction of the plutonium (free of fission-product contamination) from the spent natural uranium fuel. In 1943, several methods were proposed for separating the relatively small quantity of plutonium from the uranium and fission products. The first method selected, a precipitation process called
9048-540: The reactor core while awaiting eventual transfer to permanent storage. Many of the elements that form volatile high- valence fluorides will also form volatile high-valence chlorides. Chlorination and distillation is another possible method for separation. The sequence of separation may differ usefully from the sequence for fluorides; for example, zirconium tetrachloride and tin tetrachloride have relatively low boiling points of 331 °C (628 °F) and 114.1 °C (237.4 °F). Chlorination has even been proposed as
9152-720: The reactor safer and more practical. Fluorine has only one stable isotope ( F ), and does not easily become radioactive under neutron bombardment. Compared to chlorine and other halides, fluorine also absorbs fewer neutrons and slows (" moderates ") neutrons better. Low- valence fluorides boil at high temperatures, though many pentafluorides and hexafluorides boil at low temperatures. They must be very hot before they break down into their constituent elements. Such molten salts are "chemically stable" when maintained well below their boiling points. Fluoride salts dissolve poorly in water, and do not form burnable hydrogen. Chlorine has two stable isotopes ( Cl and Cl ), as well as
9256-415: The reprocessing of commercial reactor spent nuclear fuel . The key issue driving this policy was the risk of nuclear weapons proliferation by diversion of plutonium from the civilian fuel cycle, and to encourage other nations to follow the US lead. After that, only countries that already had large investments in reprocessing infrastructure continued to reprocess spent nuclear fuel. President Reagan lifted
9360-577: The reprocessing of other nuclear reactor material, such as Zircaloy cladding. The high radioactivity of spent nuclear material means that reprocessing must be highly controlled and carefully executed in advanced facilities by specialized personnel. Numerous processes exist, with the chemical based PUREX process dominating. Alternatives include heating to drive off volatile elements, burning via oxidation, and fluoride volatility (which uses extremely reactive Fluorine ). Each process results in some form of refined nuclear product, with radioactive waste as
9464-536: The role of molten salt: The use of molten salt as fuel and as coolant are independent design choices – the original circulating-fuel-salt MSRE and the more recent static-fuel-salt SSR use salt as fuel and salt as coolant; the DFR uses salt as fuel but metal as coolant; and the FHR has solid fuel but salt as coolant. MSRs can be burners or breeders. They can be fast or thermal or epithermal . Thermal reactors typically employ
9568-451: The salt at startup and reduces the risk of the salt freezing as it is cooled in the heat exchanger. Due to the high " redox window" of fused fluoride salts, the redox potential of the fused salt system can be changed. Fluorine-lithium-beryllium (" FLiBe ") can be used with beryllium additions to lower the redox potential and nearly eliminate corrosion. However, since beryllium is extremely toxic, special precautions must be engineered into
9672-418: The salt mixtures were removed using gas sparging of HF / H 2 mixture, with the salt heated to 600 °C. Structural metal contamination in the salt mixtures were removed using hydrogen gas sparging, at 700 °C. Solid ammonium hydrofluoride was proposed as a safer alternative for oxide removal. The possibility of online processing can be an MSR advantage. Continuous processing would reduce
9776-464: The salt, from top to bottom, is broken up into 16 groups of pumps and heat exchangers located around the core. The fuel salt takes approximately 3 to 4 seconds to complete a full cycle. At any given time during operation, half of the total fuel salt volume is in the core and the rest is in the external fuel circuit (salt collectors, salt-bubble separators, fuel heat exchangers, pumps, salt injectors and pipes). MSFRs contain an emergency draining system that
9880-420: The sectors are replaced. Emergency draining is done when an irregularity occurs during operation. The fuel salt can be drained directly into the emergency draining tank either by active devices or by passive means. The draining must be fast to limit the fuel salt heating in a loss of heat removal event. The fluoride salt-cooled high-temperature reactor (FHR), also called advanced high temperature reactor (AHTR),
9984-459: The spent fuel over the following 100 years of cooling (however, most of the other half is from strontium-90 , which has a similar half-life). The estimated overall mass balance for 20,000 g of processed fuel with 2,000 g of cladding is: In the fluoride volatility process, fluorine is reacted with the fuel. Fluorine is so much more reactive than even oxygen that small particles of ground oxide fuel will burst into flame when dropped into
10088-506: The tritium. Tritium is a difficult contaminant to remove from aqueous solution, as it cannot be separated from water except by isotope separation. However, tritium is also a valuable product used in industry science and nuclear weapons , so recovery of a stream of hydrogen or water with a high tritium content can make targeted recovery economically worthwhile. Other volatile elements leave the fuel and must be recovered, especially iodine , technetium , and carbon-14 . Voloxidation also breaks up
10192-420: The use of a double bypass. This allows the temperature of the intermediate fluid at the conversion exchanger inlet to be held constant while its temperature is increased in a controlled way at the inlet of the intermediate exchangers. The temperature of the core can be adjusted by varying the proportion of bubbles injected in the core since it reduces the salt density. As a result, it reduces the mean temperature of
10296-534: The use of fuel in liquid form, which also serves as primary coolant. The fuel would be about 20% enriched uranium tetrafluoride and thorium tetrafluoride . The ThorCon design is a floating power station to be built in a shipyard and then towed to the location of operation. Thorcon claims that this reactor design will be safer than traditional nuclear reactors. The design includes several features intended to prevent meltdowns , contain radioactive materials , and protect from terrorism and sabotage . A 2017 study by
10400-534: Was developed in Russia and the Czech Republic ; it is designed to completely remove the most troublesome radioisotopes (Sr, Cs and minor actinides ) from the raffinate remaining after the extraction of uranium and plutonium from used nuclear fuel . The chemistry is based upon the interaction of caesium and strontium with polyethylene glycol and a cobalt carborane anion (known as chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and
10504-405: Was maintained in the +6 oxidation state by addition of a dichromate salt. The bismuth phosphate was next re-precipitated, leaving the plutonium in solution, and an iron(II) salt (such as ferrous sulfate ) was added. The plutonium was again re-precipitated using a bismuth phosphate carrier and a combination of lanthanum salts and fluoride added, forming a solid lanthanum fluoride carrier for
10608-427: Was run for a few weeks and at essentially zero power, although it reached criticality. The operating temperature was held constant at approximately 675 °C (1,250 °F). The PWAR-1 used NaF/ZrF 4 /UF 4 as the primary fuel and coolant. It was one of three critical MSRs ever built. Oak Ridge National Laboratory (ORNL) took the lead in researching MSRs through the 1960s. Much of their work culminated with
10712-400: Was the focus of much debate over the first decade of the 2000s. Studies have modeled the total fuel cycle costs of a reprocessing-recycling system based on one-time recycling of plutonium in existing thermal reactors (as opposed to the proposed breeder reactor cycle) and compare this to the total costs of an open fuel cycle with direct disposal. The range of results produced by these studies
10816-445: Was used solely to extract plutonium for producing nuclear weapons . With commercialization of nuclear power , the reprocessed plutonium was recycled back into MOX nuclear fuel for thermal reactors . The reprocessed uranium , also known as the spent fuel material, can in principle also be re-used as fuel, but that is only economical when uranium supply is low and prices are high. Nuclear reprocessing may extend beyond fuel and include
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