Nuclear reprocessing
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{{portal|Energy}}
'''Nuclear reprocessing''' separates components of [[spent nuclear fuel]] such as:
*[[Reprocessed uranium]]
*[[Plutonium]]
*[[Minor actinides]]
*[[Fission products]]
*[[Activation products]]
*[[Cladding]]
Reprocessing serves multiple purposes, whose relative importance has changed over time:
*Producing plutonium as fuel for [[nuclear weapons]]
*Recycling plutonium, uranium, and other [[actinide]]s as fuel for [[breeder reactor]]s, closing the [[nuclear fuel cycle]] and extending the [[energy]] extractable per original unit of uranium by a factor of more than 60.<ref> {{cite web |url=http://www.uic.com.au/nip75.htm |title=Supply of Uranium: Nuclear issues briefing paper #75 |publisher=[[Uranium Information Centre]], |, Melbourne, Australia |publisherurl=http://www.uic.com.au/nip.html |accessdate=2007-06-17}}</ref>
*After breeder reactors were postponed, recycling plutonium once as [[MOX fuel]] for [[thermal reactor]]s, extending energy extracted by only about 30%
*Management of [[nuclear waste]] [[radioactivity]] by separating:
**actinides for destruction by [[nuclear fission]]
**some [[long-lived fission products]] ([[Tc-99|<sup>99</sup>Tc]], [[Iodine-129|<sup>129</sup>I]]) and activation products for destruction by [[nuclear transmutation]] by [[neutron capture]]
**other [[long-lived fission products]] and activation products for very long term storage in a [[deep geological repository]] which may have limited capacity
**bulky but [[low-level waste]] such as reprocessed uranium for less stringent disposal
**shorter-term secure storage for decay of dangerous medium-lived [[radionuclides]] such as [[cesium-137|<sup>137</sup>Cs]] and [[strontium-90|<sup>90</sup>Sr]]
**radionuclides for industrial uses
==History==
The first large-scale nuclear reactors were built during [[World War II]]. These reactors were designed for the production of plutonium for use in [[nuclear weapon]]s. The only reprocessing required, therefore, was the extraction of the [[plutonium]] (free of [[fission product|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 the [[Bismuth]] [[Phosphate]] process, was developed and tested at the [[Oak Ridge National Laboratory]] (ORNL) in the 1943-1945 period to produce quantities of plutonium for evaluation and use in [[History of nuclear weapons|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 the latter 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 a smaller plant at [[West Valley, New York]] which closed by 1972.<ref> {{cite web |title=Plutonium Recovery from Spent Fuel Reprocessing by Nuclear Fuel Services at West Valley, New York from 1966 to 1972 |url=http://www.osti.gov/opennet/document/purecov/nfsrepo.html | accessdate=2007-06-17 |publisher=U.S. Department of Energy |date=February 1996}}</ref>
Processing of civilian fuel has long been employed in Europe (at the [[COGEMA La Hague site]]) and briefly at the [[West Valley Reprocessing Plant]] in the U.S.
In October 1976, fear 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. This was confirmed by President Jimmy Carter in 1977. After that, only countries that already had large investments in reprocessing infrastructure continued to reprocess spent nuclear fuel. President Reagan lifted 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 own policy and signed a contract with a [[consortium]] comprised of Duke Energy, COGEMA, and Stone & Webster (DCS) to design and operate a [[MOX fuel|Mixed Oxide (MOX) fuel]] fabrication facility. Site preparation at the Savannah River Site (South Carolina) began in October of 2005. <!-- link lacks relevant information [http://www.dcsmox.com/index.html] -->
The [[Global Nuclear Energy Partnership]], announced by the secretary of the Department of Energy, [[Samuel Bodman]], on [[February 6]], [[2006]], is a plan to form an international partnership to reprocess [[spent nuclear fuel]] in a way that renders the plutonium in it usable for [[nuclear fuel]] but not for [[nuclear weapons]].
==Aqueous / organic solvent methods==
===Obsolete methods===
==== Bismuth phosphate ====
The '''bismuth phosphate''' process is a very old process which adds lots of material to the final highly active [[waste]]. It was replaced by solvent extraction processes. The process was designed to extract [[plutonium]] from [[aluminium]]-clad [[uranium]] metal fuel. The fuel was declad by boiling it in [[caustic soda]]. After decladding, the uranium metal was dissolved in [[nitric acid]]. The plutonium at this point is in the +4 oxidation state. It was then precipitated by the addition of [[bismuth]] nitrate and [[phosphoric acid]] to form the bismuth phosphate. The plutonium was [[Coprecipitation|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]] which converted the plutonium to PuO<sub>2</sub><sup>2+</sup> (Pu VI), then a [[dichromate]] salt was added to maintain the plutonium in the +6 oxidation state. The bismuth phosphate was then re-precipitated leaving the plutonium in solution. Then an [[iron]] (II) salt such as ''ferrous sulfate'' was added and the plutonium re-precipitated again using a bismuth phosphate carrier precipitate. Then [[lanthanum]] salts and [[fluoride]] were added to create solid lanthanum fluoride which acted as a carrier for the Pu. This was converted to the oxide by the action of a base. The lanthanum plutonium oxide was then collected and extracted with nitric acid to form plutonium nitrate.<ref> {{cite web |url=http://www.bonestamp.com/sgt/process.htm |title=The plutonium production story at the Hanford Site: processes and facilities history (WHC-MR-0521) (excerpts) |publisher=Department of Energy |author=Gerber, Michelle}}</ref>
==== Hexone or Redox ====
This is a liquid-liquid extraction process which uses methyl isobutyl ketone 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 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 has been replaced by PUREX.<ref>{{cite web |url=http://www.osti.gov/energycitations/product.biblio.jsp?osti_id=4134289 |title=Method for separation of plutonium from uranium and fission products by solvent extraction |author=Seaborg, Glenn T. et al|date=1960-08-23|publisher=U.S. Patent and Trademark Office}}</ref><ref>{{cite web |url=http://www.llnl.gov/tid/lof/documents/pdf/235702.pdf |format=[[PDF]] |title=From separations to reconstitution--a short history of plutonium in the U.S. and Russia (UCRL-JC-133802) |author=L.W. Gray |date=1999-04-15 |publisher=Lawrence Livermore National Laboratory preprint}}</ref>
Pu<sup>4+</sup> + 4NO<sub>3</sub><sup>-</sup> + 2S --> [Pu(NO<sub>3</sub>)<sub>4</sub>S<sub>2</sub>]
==== Butex, β,β'-dibutyoxydiethyl ether ====
A process based on a solvation extraction process using the triether extractant named above. This process has 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]] many years ago. This process has been replaced by PUREX.
===PUREX, the current method ===
{{Main|PUREX}}
'''PUREX''' 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]], in order to extract [[uranium]] and [[plutonium]], independent of each other, from the [[Nuclear fission|fission]] products. This is the most developed and widely used process in the industry at present.
When used on fuel from commercial power reactors the plutonium extracted typically contains too much Pu-240 to be useful in a nuclear weapon. However, reactors that are capable of refuelling frequently can be used to produce [[nuclear weapon|weapon-grade]] plutonium, which can later be recovered using PUREX. Because of this, PUREX chemicals are monitored.{{Fact|date=August 2007}}
===UREX===
The PUREX process can be modified to make a '''UREX''' ('''UR'''anium '''EX'''traction) process which could be used to save space inside high level [[nuclear waste]] disposal sites, such as [[Yucca Mountain]], 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 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 [[actinide]]s. 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 greater proliferation resistance than with the plutonium extraction stage of the PUREX process.
===TRUEX===
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 USA by Argonne National Laboratory and is designed to remove the transuranic metals (Am/Cm) from waste. The idea 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.
===DIAMEX===
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 the formation of acidic gases which could contribute to [[acid rain]]. 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.
===SANEX===
'''S'''elective '''A'''cti'''N'''ide '''EX'''traction. As part of 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. In order 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 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-triaiznyl pyridine (BTP) based process.<ref><!-- These preprints need to be replaced with appropriate peer-reviewed articles. --> [http://www-atalante2004.cea.fr/home/liblocal/docs/atalante2000/P3-26.pdf] [http://www.nea.fr/html/pt/docs/iem/jeju02/session2/Session%20II-19.pdf] [http://www-atalante2004.cea.fr/home/liblocal/docs/atalante2000/P3-24.pdf]</ref>
Other systems such as the dithiophosphinic acids are being worked on by some other workers.
===UNEX===
This is the '''''UN'''iversal'' '''EX'''traction process which was developed in [[Russia]] and the [[Czech Republic]], it is a process designed to remove all of the most troublesome (Sr, Cs and [[minor actinides]]) [[radioisotopes]] from the raffinates left after the extraction of uranium and plutonium from used [[nuclear fuel]].<ref> {{cite web |url=http://www.usembassy.it/file2001_12/alia/a1121910.htm | title=U.S.-Russia Team Makes Treating Nuclear Waste Easier |publisher=U.S. embassy press release(?) | date=2001-12-19 |accessdate=2007-06-14}}</ref><ref>{{cite web |url=http://www.osti.gov/bridge/product.biblio.jsp?osti_id=765723 | title=INTEC High-Level Waste Studies Universal Solvent Extraction Feasibility Study |publisher=INEEL Technical report |date=2001-09-01 |author=J. Banaee et al.}}</ref> The chemistry is based upon the interaction of [[caesium]] and [[strontium]] with poly [[ethylene oxide]] (poly [[ethylene glycol]]) <!-- this is inappropriate as a reference.. looks like a progress report to a sponsor <ref>{{cite web url=http://www.osti.gov/em52/2003projsum/81895.pdf |format=PDF |title=project summary}}</ref> --> and a [[cobalt]] [[carborane]] [[anion]] (known as chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and the [[diluent]] is a polar [[aromatic]] such as [[nitrobenzene]]. Other dilents such as ''meta''-nitrobenzotri[[fluoride]] and phenyl trifluoromethyl [[sulfone]]<ref> {{cite web| url=http://www.wmsym.org/Abstracts/2001/62/62-7.pdf | title=Flowsheet testing of the universal solvent extraction process for the simultaneous separation of caesium, strontium, and the actinides from dissolved INEEL calcine |author=J.D. Law et al. | publisher=WM 2001 conference proceedings| date=2001-03-01|accessdate=2006-06-17}}</ref> have been suggested as well.
===Electrochemical method in aqueous alkali===
An exotic method using [[electrochemistry]] and [[ion exchange]] in [[ammonium]] [[carbonate]] has been reported.<ref>{{cite journal |url=http://www.jstage.jst.go.jp/article/jnst/43/3/255/_pdf |title=Andodic dissociation of UO<sub>2</sub> pellet containing simulated fission products in ammonium carbonate solution| journal=Journal of Nuclear Science and Technology |volume=43 |pages=255–262 |author=Asanuma, Noriko, et al |doi=10.3327/jnst.43.255 |year=2006}}</ref>
== Pyroprocessing ==
'''[[Pyroprocessing]]''' is a generic term for several kinds of [[Pyrometallurgical]] Reprocessing. These processes are not currently in significant use worldwide, but they have been researched and developed at Argonne National Laboratory and elsewhere. The principles behind them are well understood, and no significant technical barriers exist to their adoption {{Fact|date=May 2008}}. The primary economic hurdle to widespread adoption is that reprocessing as a whole is not currently ([[2005]]) in favor, and places that do reprocess already have PUREX plants constructed. Consequently, there is little demand for new pyrometalurgical systems, although there could be if the [[Generation IV reactor]] programs become reality.
Pyrometallurgical processing techniques involve several stages: volatilisation, liquid-liquid extraction using immiscible metal-metal phases or metal-salt phases, [[electrorefining]] in molten salt, fractional crystallisation, etc. They are generally based on the use of either fused (low-melting point) salts such as chlorides or fluorides (eg LiCl+KCl or LiF+CaF2) or fused metals such as cadmium, bismuth or aluminium. They are most readily applied to metal rather than oxide fuels.
===Advantages and disadvantages===
'''Advantages'''
* Pyroprocessing can readily be applied to high burn-up fuel and fuel which has had little cooling time, since the operating temperatures are high already.
* It does not use water. Water is problematic in nuclear chemistry for many reasons. First of all, it tends to serve as a moderator, and accelerate nuclear reactions. Secondly, it is easily contaminated, and not easily cleaned up, and it tends to evaporate, potentially taking [[tritium]] with it. This is not as large a disadvantage as it might first appear as it is possible to treat normal oxide fuel using a process called Voloxidation<ref name=advancedheadend /> which removes 99% of the tritium from used fuel. The tritium can be recovered in the form of a strong solution which might be suitable for use as a supply of tritium for industrial applications.
* It separates out all [[actinide]]s, and therefore produces fuel that is heavily spiked with heavy actinides, such as Plutonium (240+), and Curium 242. This does not prevent the fuel from being suitable for reactors, but it makes it hard to manipulate, steal, or make nuclear weapons from. (However, the difficulty has been questioned.<ref>{{cite web|url=http://www.princeton.edu/~globsec/publications/pdf/13_3%20Kang%20vonhippel.pdf|title=Limited Proliferation-Resistance Benefits from Recycling Unseparated Transuranics and Lanthanides from Light-Water Reactor Spent Fuel|page=4}}</ref>) In contrast, the PUREX process can easily produce separated Uranium and Plutonium, and also tends to leave the remaining actinides (like Curium) behind, producing more dangerous nuclear waste.
* It is somewhat more efficient and considerably more compact than aqueous processing methods, allowing the possibility of on-site reprocessing of reactor wastes. This circumvents various transportation and security issues, allowing the reactor to simply store a small volume (perhaps a few percent of the original volume of the spent fuel) of fission product laced salt on site until decommissioning, when everything could be dealt with at once.
* Since pyrometalurgy recovers all the actinides, the remaining waste is not nearly as long lived as it would otherwise be. Most of the long term (past a couple hundred years) radioactivity produced by nuclear waste is produced by the actinides. These actinides can (mostly) be consumed by reactors as fuel, so extracting them from the waste and reinserting them into the reactor reduces the long term threat from the waste, and reduces the fuel needs of the reactor.
'''Disadvantages'''
* The used salt from pyro processing is not suitable for conversion into a glass in the same way as the [[raffinate]] from PUREX processing.
===PYRO-A and -B for IFR===
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 electrical current 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.<ref> {{cite web |url=http://criepi.denken.or.jp/en/e_publication/pdf/den363.pdf|publisher=CRIEPI News | title=Development of pyro-process fuel cell technology|date=July 2002}} </ref><ref> {{cite web| url=http://www.nea.fr/html/pt/docs/iem/madrid00/Proceedings/Paper56.pdf|title=Development of plutonium recovery process by molten salt electrorefining with liquid cadmium cathode|author=Masatoshi Iizuka|publisher=Proceedings of the 6th information exchange meeting on actinide and fission product partitioning and transmutation (Madrid, Spain) |date=2001-12-12}}</ref><ref>http://www.nea.fr/html/pt/iempt8/abstracts/Abstracts/Session_II/zvejskova.ppt</ref> As alternatives to the moltern cadmium electrode it is possible to use a molten [[bismuth]] cathode, or a solid aluminium cathode.<ref> [http://www.nea.fr/html/pt/docs/iem/jeju02/session2/SessionII-06.pdf Elecrochemical Behaviours of Lanthanide Fluorides in the Electrolysis System with LiF-NaF-KF Salt<!-- Bot generated title -->]</ref><!-- this reference is not good enough.. notes from a conference.. <ref>[http://www.nea.fr/html/pt/docs/iem/jeju02/session2/Summary_sessionII.pdf]</ref> -->
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.<ref>http://www.merck.de/servlet/PB/show/1332930/10.Molten%20Salts%20Lanthanides.pdf</ref>
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 the level of various naturally occurring minerals and ores within a few hundred, rather than thousands, years.<ref>{{cite web
| url= http://www.ne.doe.gov/AFCI/neAFCI.html
| title= Advanced Fuel Cycle Initiative
| publisher= [[United States Department of Energy|U.S. Department of Energy]]
| accessdate= 2008-05-03 }}</ref>
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]] in order to be burned efficiently. In a [[thermal neutron]] spectrum, the concentrations of several heavy actinides ([[Curium|Curium-242]] and [[Plutonium|Plutonium-240]]) can become quite high, creating fuel that is substantially different from the usual [[Uranium]] or mixed 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 is free of uranium and contains recovered [[transuranic]]s 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 the other fission products are sent to waste.
===Voloxidation===
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]].<ref name="advancedheadend"/> A major purpose is to capture [[tritium]] as tritiated water vapor before further processing where it would be difficult to retain the tritium. Other volatile elements leave the fuel and must be recovered, especially [[iodine]], [[technetium]], and [[carbon-14]]. Voloxidation also breaks up the fuel or increases its surface area to enhance penetration of reagents in following reprocessing steps.
===Volatilization in isolation===
Simply heating spent oxide fuel in an inert atmosphere or vacuum at a temperature between 700°C and 1000°C as a first reprocessing step can remove several volatile elements, including caesium whose isotope [[Cs-137]] emits about half of the heat produced by the spent fuel over the following 100 years of cooling (however, most of the other half is from [[Sr-90]] which remains).
The estimated overall mass balance for 20,000 grams of processed fuel with 2,000 grams of cladding is:<ref> {{cite web| url=http://web.mac.com/mosb1000/iWeb/Bob's%20Site/Examples_files/Sr_Design_Rpt.pdf |format=PDF | title=Removal of caesium from spent nuclear fuel destined for the electrorefiner fuel treatment process| author=Wolverton, Daren et al. | publisher=University of Idaho (dissertation?) | date=2005-05-11}}</ref>
{| class="wikitable"
! !!Input !!Residue !![[Zeolite]]<br>filter!!Carbon<br>filter!!Particle<br>filters
|-
|[[Palladium]]||28||14||14
|-
|[[Tellurium]]||10||5||5
|-
|[[Molybdenum]]||70|| ||70
|-
|[[Caesium]]||46|| ||46
|-
|[[Rubidium]]||8|| ||8
|-
|[[Silver]]||2|| ||2
|-
|[[Iodine]]||4|| || ||4
|-
|Cladding||2000||2000||
|-
|[[Uranium]]||19218||19218|| || ||?
|-
|Others||614||614|| || ||?
|-
|Total||22000||21851||145||4||0
|}
Tritium is not mentioned in this paper.
=== Fluoride volatility ===
{{Main|Fluoride volatility}}
[[Image:fission yield volatile.png|thumb|450px|Blue elements have volatile fluorides or are already volatile; green elements do not but have volatile chlorides; red elements have neither, but the elements themselves are volatile at very high temperatures. Yields at 10<sup>0,1,2,3</sup> years after [[fission]], not considering later [[neutron capture]], fraction of 100% not 200%. [[Beta decay]] [[Kr-85]]→[[Rubidium|Rb]], [[Sr-90]]→[[Zirconium|Zr]], [[Ru-106]]→[[Palladium|Pd]], [[Sb-125]]→[[Tellurium|Te]], [[Cs-137]]→[[Barium|Ba]], [[Ce-144]]→[[Neodymium|Nd]], [[Sm-151]]→[[Europium|Eu]], [[Eu-155]]→[[Gadolinium|Gd]] visible.]]
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 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 the [[fission product]]s 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 is decreased.<ref>http://books.google.com/books?id=SJOE00whg44C&pg=PA66&lpg=PA66&dq=fission+product+cross+sections&source=web&ots=G5cQlwmIEq&sig=2K5eRWNKbAJ0T_jDpckg1tCAMSQ&hl=en#PPA66,M1</ref> 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 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 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 [[actinide]]s to the molten fuel mixture for eventual fission, while removing [[fission product]]s that are [[neutron poison]]s, or that can be more securely stored outside the reactor core while awaiting eventual transfer to permanent storage.
===Chloride volatility and solubility===
Many of the elements that form volatile high-[[valence (chemistry)|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 and 114.1°C. Chlorination has even been proposed as a method for removing zirconium fuel cladding,<ref name=advancedheadend> {{cite web |url=http://www.ornl.gov/~webworks/cppr/y2001/pres/123514.pdf |format=PDF |title=Advanced Head-End Processing of Spent Fuel: A Progress Report |author=Guillermo D. Del Cul, et al |work=2005 ANS annual meeting |publisher= [[Oak Ridge National Laboratory]], U.S. DOE |accessdate=2008-05-03}}</ref> 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]].
== Economics of reprocessing nuclear fuel ==
The relative [[economics]] of reprocessing-waste disposal and interim storage-direct disposal has been the focus of much debate over the past ten years. Studies have modelled the total fuel cycle costs of a reprocessing-recycling system based on one-time recycling of plutonium in existing [[thermal reactor]]s (as opposed to the proposed [[fast 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 is very wide, but all are agreed that under current (2005) economic conditions the reprocessing-recycle option is the more costly.
If reprocessing is undertaken only to reduce the radioactivity level of spent fuel it should be taken into account that spent nuclear fuel becomes less radioactive over time. After 40 years its radioactivity drops by 99.9%,<ref>{{cite web
| url= http://www.world-nuclear.org/how/wastemanag.html
| title= Waste Management and Disposal
| publisher= [[World Nuclear Association]]
| accessdate= 2008-05-03 }}</ref> though it still takes over a thousand years for the level of radioactivity to approach that of natural uranium.<ref>{{cite web| url=http://www.world-nuclear.org/info/inf103.html |publisher= World Nuclear Association |title=Radioactive Wastes: Myths and Realities |date=2006-06 |accessdate=2008-05-03}}</ref> However the level of [[transuranic element]]s,
including [[plutonium-239]], remains high for over 100,000 years, so if not reused as nuclear fuel, then those elements need secure disposal because of [[nuclear proliferation]] reasons as well as radiation hazard.
*[http://www.wise-uranium.org/nfccr.html Recycled Nuclear Fuel Cost Calculator designed by the WISE Uranium Project]
==List of nuclear reprocessing sites==
{| class="wikitable"
! Fuel type
! Reprocessing site
! Reprocessing<br>capacity
|-
| Light Water Reactor Fuel
| [[COGEMA La Hague site]], France
| 1700 tonnes/year
|-
|
| [[Thorp nuclear fuel reprocessing plant]] at [[Sellafield]], United Kingdom
| 900 tonnes/year
|-
|
| Rokkasho nuclear fuel reprocessing plant, Japan
| 800 tonnes/year
|-
|
| [[Mayak]], Russia
| 400 tonnes/year
|-
| Other Nuclear Fuels
| [[B205]] at [[Sellafield]], United Kingdom
| 1500 tonnes/year
|-
|
| Kalpakkam Atomic reprocessing plant, India
| 275 tonnes/year
|}
==See also==
* [[Nuclear fuel cycle]]
* [[Nuclear breeder reactor]]
* [[Spent nuclear fuel shipping cask]]
* [[Global Nuclear Energy Partnership]] announced February, 2006
* [[Megatons to Megawatts Program]]
== References ==
{{reflist}}
*OECD Nuclear Energy Agency, [http://www.nea.fr/html/ndd/reports/efc/ The Economics of the Nuclear Fuel Cycle], Paris, 1994
*I. Hensing and W Schultz, Economic Comparison of Nuclear Fuel Cycle Options, Energiewirtschaftlichen Instituts, Cologne, 1995.
*Cogema, Reprocessing-Recycling: the Industrial Stakes, presentation to the Konrad-Adenauer-Stiftung, Bonn, 9 May 1995.
*OECD Nuclear Energy Agency, Plutonium Fuel: An Assessment, Paris, 1989.
*National Research Council, "Nuclear Wastes: Technologies for Separation and Transmutation", National Academy Press, Washington D.C. 1996.
==External links==
* [http://www.world-nuclear.org/info/inf69.htm Processing of Used Nuclear Fuel], World Nuclear Association
*[http://www.euronuclear.org/info/encyclopedia/p/purex-process.htm PUREX Process, European Nuclear Society]
*[http://www.world-nuclear.org/info/inf29.htm Mixed Oxide Fuel (MOX)] - World Nuclear Association
*[http://www.globalsecurity.org/wmd/library/report/crs/97-564.htm Disposal Options for Surplus Weapons-Usable Plutonium] - Congressional Research Service Report for Congress
*[http://mfnl.xjtu.edu.cn/gov-doe-ornl/RDF/history/03_FuelReprocessinHistory.html Brief History of Fuel Reprocessing]
[[Category:Nuclear reprocessing| ]]
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