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78-424: The plant is a pool type LMFBR , where the reactor, coolant pumps, intermediate heat exchangers and associated piping are all located in a common liquid sodium pool. The reactor system is housed in a concrete rectilinear building and provided with filtration and gas containment. In the first 24 years of operations, there have been 12 water-into-sodium leaks in the steam generators , routinely addressed by isolating

156-515: 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

234-472: 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

312-838: 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

390-481: A diameter of 2.05 metres (6 ft 9 in). It has 369 fuel assemblies , mounted vertically; each consists of 127 fuel rods enriched to between 17–26% U . In comparison, normal enrichment in other Russian reactors is between 3–4% U. The control and scram system is composed of 27 reactivity control elements including 19 shimming rods, two automatic control rods, and six automatic emergency shut-down rods. On-power refueling equipment allows for charging

468-456: A fuel cycle based on pyrometallurgical reprocessing in facilities integrated with the reactor. The second is a medium to large (500–1,500 MWe) sodium-cooled reactor with mixed uranium-plutonium oxide fuel, supported by a fuel cycle based upon advanced aqueous processing at a central location serving multiple reactors. The outlet temperature is approximately 510–550 degrees C for both. Liquid metallic sodium may be used to carry heat from

546-541: A half-life of only 15 hours. Another problem is leaks. Sodium at high temperatures ignites in contact with oxygen. Such sodium fires can be extinguished by powder, or by replacing the air with nitrogen . A Russian breeder reactor, the BN-600, reported 27 sodium leaks in a 17-year period, 14 of which led to sodium fires. No fission products have a half-life in the range of 100 a–210 ka ... ... nor beyond 15.7 Ma The operating temperature must not exceed

624-655: A large margin to coolant boiling, a primary cooling system that operates near atmospheric pressure, and an intermediate sodium system between the radioactive sodium in the primary system and the water and steam in the power plant. Innovations can reduce capital cost, such as modular designs, removing a primary loop, integrating the pump and intermediate heat exchanger, and better materials. The SFR's fast spectrum makes it possible to use available fissile and fertile materials (including depleted uranium ) considerably more efficiently than thermal spectrum reactors with once-through fuel cycles. In 2020 Natrium received an $ 80M grant from

702-476: 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

780-526: 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

858-522: 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

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936-983: 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

1014-446: 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

1092-482: 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

1170-501: 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,

1248-459: 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

1326-424: 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

1404-461: 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

1482-467: 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

1560-466: 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

1638-423: Is limited by the production of plutonium from uranium. One work-around is to have an inert matrix, using, e.g., magnesium oxide . Magnesium oxide has an order of magnitude lower probability of interacting with neutrons (thermal and fast) than elements such as iron. High-level wastes and, in particular, management of plutonium and other actinides must be handled. Safety features include a long thermal response time,

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1716-456: 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

1794-517: 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

1872-515: Is that metal atoms are weak neutron moderators. Water is a much stronger neutron moderator because the hydrogen atoms found in water are much lighter than metal atoms, and therefore neutrons lose more energy in collisions with hydrogen atoms. This makes it difficult to use water as a coolant for a fast reactor because the water tends to slow (moderate) the fast neutrons into thermal neutrons (although concepts for reduced moderation water reactors exist). Another advantage of liquid sodium coolant

1950-403: Is that sodium melts at 371K (98°C) and boils / vaporizes at 1156K (883°C), a difference of 785K (785°C) between solid / frozen and gas / vapor states. By comparison, the liquid temperature range of water (between ice and gas) is just 100K at normal, sea-level atmospheric pressure conditions. Despite sodium's low specific heat (as compared to water), this enables the absorption of significant heat in

2028-504: Is transferred from the reactor core via three independent circulation loops. Each has a primary sodium pump, two intermediate heat exchangers, a secondary sodium pump with an expansion tank located upstream, and an emergency pressure discharge tank. These feed a steam generator , which in turn supplies a condensing turbine that turns the generator. There is much international interest in the fast-breeder reactor at Beloyarsk. Japan has its own prototype fast-breeder reactors. The operation of

2106-450: 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

2184-614: The Natrium appellation in Kemmerer, Wyoming . Aside from the Russian experience, Japan, India, China, France and the USA are investing in the technology. The nuclear fuel cycle employs a full actinide recycle with two major options: One is an intermediate-size (150–600 MWe) sodium-cooled reactor with uranium - plutonium -minor-actinide- zirconium metal alloy fuel, supported by

2262-704: The US Department of Energy for development of its SFR. The program plans to use High-Assay, Low Enriched Uranium fuel containing 5-20% uranium. The reactor was expected to be sited underground and have gravity-inserted control rods. Because it operates at atmospheric pressure, a large containment shield is not necessary. Because of its large heat storage capacity, it was expected to be able to produce surge power of 500 MWe for 5+ hours, beyond its continuous power of 345 MWe. Sodium-cooled reactors have included: Most of these were experimental plants that are no longer operational. On November 30, 2019, CTV reported that

2340-567: 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

2418-597: 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

BN-600 reactor - Misplaced Pages Continue

2496-457: 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

2574-424: 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

2652-466: The Canadian provinces of New Brunswick , Ontario and Saskatchewan planned an announcement about a joint plan to cooperate on small sodium fast modular nuclear reactors from New Brunswick-based ARC Nuclear Canada. Nuclear reprocessing#Pyroprocessing Nuclear reprocessing is the chemical separation of fission products and actinides from spent nuclear fuel . Originally, reprocessing

2730-560: 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

2808-609: 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

2886-476: 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

2964-470: 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

3042-507: 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

3120-596: 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

3198-510: 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

BN-600 reactor - Misplaced Pages Continue

3276-458: The coolant (the Phénix reactor outlet temperature was 833K (560°C)) permit a higher thermodynamic efficiency than in water cooled reactors. The electrically conductive molten sodium can be moved by electromagnetic pumps . The fact that the sodium is not pressurized implies that a much thinner reactor vessel can be used (e.g. 2 cm thick). Combined with the much higher temperatures achieved in

3354-446: The core with fresh fuel assemblies, repositioning and turning the fuel assemblies within the reactor, and changing control and scram system elements remotely. The unit employs a three-circuit coolant arrangement; sodium coolant circulates in both the primary and secondary circuits. Water and steam flow in the third circuit. The sodium is heated to a maximum of 550 °C (1,022 °F) in the reactor during normal operations. This heat

3432-423: The core. Sodium has only one stable isotope, sodium-23 , which is a weak neutron absorber. When it does absorb a neutron it produces sodium-24 , which has a half-life of 15 hours and decays to stable isotope magnesium-24 . The two main design approaches to sodium-cooled reactors are pool type and loop type. In the pool type, the primary coolant is contained in the main reactor vessel, which therefore includes

3510-404: 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

3588-681: 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

3666-566: 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

3744-424: 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

3822-473: The faulty module with gate valves. These incidents did not have off-site impact, did not generate radioactive material (sodium in the secondary circuit is not neutron-activated) and were not reported to IAEA, since they were deemed to have no impact on safety. As of 2022, the cumulative "energy Availability factor " recorded by the IAEA was 76.3%. The reactor core is 1.03 metres (3 ft 5 in) tall with

3900-445: 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

3978-414: 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

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4056-431: The fuel's boiling temperature. Fuel-to-cladding chemical interaction (FCCI) has to be accommodated. FCCI is eutectic melting between the fuel and the cladding; uranium, plutonium, and lanthanum (a fission product ) inter-diffuse with the iron of the cladding. The alloy that forms has a low eutectic melting temperature. FCCI causes the cladding to reduce in strength and even rupture. The amount of transuranic transmutation

4134-590: 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

4212-432: 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

4290-517: 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

4368-457: The liquid phase, while maintaining large safety margins. Moreover, the high thermal conductivity of sodium effectively creates a reservoir of heat capacity that provides thermal inertia against overheating. Sodium need not be pressurized since its boiling point is much higher than the reactor's operating temperature , and sodium does not corrode steel reactor parts, and in fact, protects metals from corrosion. The high temperatures reached by

4446-403: 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

4524-485: The metal-fueled integral fast reactor . Several sodium-cooled fast reactors have been built and some are in current operation, particularly in Russia. Others are in planning or under construction. For example, in 2022, in the US, TerraPower (using its Traveling Wave technology ) is planning to build its own reactors along with molten salt energy storage in partnership with GEHitachi's PRISM integral fast reactor design, under

4602-487: 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

4680-413: 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

4758-431: 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

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4836-481: 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

4914-448: 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

4992-413: 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

5070-495: The reaction. A disadvantage of sodium is its chemical reactivity, which requires special precautions to prevent and suppress fires. If sodium comes into contact with water it reacts to produce sodium hydroxide and hydrogen, and the hydrogen burns in contact with air. This was the case at the Monju Nuclear Power Plant in a 1995 accident. In addition, neutron capture causes it to become radioactive; albeit with

5148-442: The reactor core and a heat exchanger . The US EBR-2 , French Phénix and others used this approach, and it is used by India's Prototype Fast Breeder Reactor and China's CFR-600 . In the loop type, the heat exchangers are outside the reactor tank. The French Rapsodie , British Prototype Fast Reactor and others used this approach. All fast reactors have several advantages over the current fleet of water based reactors in that

5226-540: 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

5304-596: The reactor is an international study in progress; Russia, France, Japan, and the United Kingdom currently participate. The reactor is expected to operate up until 2040. 56°50′30″N 61°19′21″E  /  56.8416°N 61.3224°E  / 56.8416; 61.3224 Pool type LMFBR A sodium-cooled fast reactor is a fast neutron reactor cooled by liquid sodium . The initials SFR in particular refer to two Generation IV reactor proposals, one based on existing liquid metal cooled reactor (LMFR) technology using mixed oxide fuel (MOX), and one based on

5382-419: The reactor, this means that the reactor in shutdown mode can be passively cooled. For example, air ducts can be engineered so that all the decay heat after shutdown is removed by natural convection, and no pumping action is required. Reactors of this type are self-controlling. If the temperature of the core increases, the core will expand slightly, which means that more neutrons will escape the core, slowing down

5460-415: 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

5538-577: 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

5616-459: 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

5694-506: 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

5772-501: The waste streams are significantly reduced. Crucially, when a reactor runs on fast neutrons, the plutonium isotopes are far more likely to fission upon absorbing a neutron. Thus, fast neutrons have a smaller chance of being captured by the uranium and plutonium, but when they are captured, have a much bigger chance of causing a fission. This means that the inventory of transuranic waste is non existent from fast reactors. The primary advantage of liquid metal coolants, such as liquid sodium,

5850-541: 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

5928-405: 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

6006-400: 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

6084-445: 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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