Editing Nuclear reprocessing (section)

==Separation technologies==
===Water and organic solvents===
====PUREX====
{{Main|PUREX}}

'''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 [[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 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.<ref>[http://belfercenter.hks.harvard.edu/publication/2376/us_program_for_disposition_of_excess_weapons_plutonium.html?breadcrumb=%2Fpublication%2F2014%2Fenabling_a_significant_future_for_nuclear_power U.S. Program for Disposition of Excess Weapons Plutonium] {{Webarchive|url=https://web.archive.org/web/20160408045640/http://belfercenter.hks.harvard.edu/publication/2376/us_program_for_disposition_of_excess_weapons_plutonium.html?breadcrumb=%2Fpublication%2F2014%2Fenabling_a_significant_future_for_nuclear_power |date=8 April 2016 }}, IAEA-SM-346/102, Matthew Bunn, 2002.</ref> Moreover, reactors that are capable of refueling 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.<ref>{{cite book|last=Irvine|first=Maxwell|title=Nuclear power : a very short introduction|year=2011|publisher=Oxford University Press|location=Oxford|isbn=9780199584970|page=55|url=https://books.google.com/books?id=UfSr1oI4lCMC&pg=PA55|access-date=22 February 2016|archive-date=28 March 2020|archive-url=https://web.archive.org/web/20200328003740/https://books.google.com/books?id=UfSr1oI4lCMC&pg=PA55|url-status=live}}</ref>
[[File:Uranium Reprocessing.jpg|thumb|Plutonium Processing]]

====Modifications of PUREX====
Many of these concepts, particularly those that separate the [[minor actinides]] once U and Pu have been extracted, are summarised in the [[advanced reprocessing of spent nuclear fuel]]. 

=====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 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 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 somewhat greater proliferation resistance than with the plutonium extraction stage of the PUREX process.{{citation needed|date=December 2012}}

=====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 US by [[Argonne National Laboratory]] and is designed to remove the transuranic metals (Am/Cm) from waste. The idea is that by lowering the [[Alpha decay#Toxicity|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]] (although the acidic gases could be recovered by a scrubber). The DIAMEX process is being worked on in Europe by the French [[Commissariat à l'énergie atomique|CEA]]. The process is sufficiently mature that an industrial plant could be constructed with the existing knowledge of the process.<ref>{{cite web|title=Nuclear Energy: Fuel of the Future? |url=http://blogs.princeton.edu/chm333/f2006/nuclear/05_reprocessing/possible_methods_of_reprocessing/ |publisher=Princeton University |access-date=6 April 2013 |url-status=dead |archive-url=https://web.archive.org/web/20121001033741/http://blogs.princeton.edu/chm333/f2006/nuclear/05_reprocessing/possible_methods_of_reprocessing/ |archive-date=1 October 2012 }}</ref> 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 [[lanthanide]]s and trivalent minor [[actinide]]s 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 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 [[Commissariat à l'énergie atomique|CEA]] is working on a [[bis-triazinyl pyridine]] (BTP) based process.<ref><!-- These preprints need to be replaced with appropriate peer-reviewed articles. -->C. Hill, D. Guillaneux, X. Hérès, N. Boubals and L. Ramain [http://www-atalante2004.cea.fr/home/liblocal/docs/atalante2000/P3-26.pdf SANEX-BTP PROCESS DEVELOPMENT STUDIES] {{webarchive|url=https://web.archive.org/web/20121115151847/http://www-atalante2004.cea.fr/home/liblocal/docs/atalante2000/P3-26.pdf|date=15 November 2012}}</ref><ref>C. Hill, L. Berthon, P. Bros, J-P. Dancausse and D. Guillaneux [http://www.nea.fr/html/pt/docs/iem/jeju02/session2/Session%20II-19.pdf SANEX-BTP PROCESS DEVELOPMENT STUDIES] {{Webarchive|url=https://web.archive.org/web/20090905052047/http://www.nea.fr/html/pt/docs/iem/jeju02/session2/Session%20II-19.pdf |date=5 September 2009 }}. Commissariat à l'Énergie Atomique</ref><ref>Béatrice Rat, Xavier Hérès [http://www-atalante2004.cea.fr/home/liblocal/docs/atalante2000/P3-24.pdf Modelling and achievement of a SANEX process flowsheet for trivalent actinides/lanthanides separation using BTP extractant (bis-1,2,4-triazinyl-pyridine).] {{webarchive|url=https://web.archive.org/web/20051016074429/http://www-atalante2004.cea.fr/home/liblocal/docs/atalante2000/P3-24.pdf |date=16 October 2005 }}</ref>
Other systems such as the dithiophosphinic acids are being worked on by some other workers.

=====UNEX=====
The '''''UN'''iversal'' '''EX'''traction process 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]].<ref>{{cite web |url=http://www.usembassy.it/file2001_12/alia/a1121910.htm |archive-url=https://web.archive.org/web/20140728001906/http://www.usembassy.it/file2001_12/alia/a1121910.htm |archive-date=28 July 2014 | title=U.S.-Russia Team Makes Treating Nuclear Waste Easier |publisher=U.S. embassy press release(?) | date=19 December 2001 |access-date=14 June 2007}}</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=1 September 2001 |author=J. Banaee |display-authors=etal |access-date=28 January 2006 |archive-date=13 May 2013 |archive-url=https://web.archive.org/web/20130513122509/http://www.osti.gov/bridge/product.biblio.jsp?osti_id=765723 |url-status=live }}</ref> The chemistry is based upon the interaction of [[caesium]] and [[strontium]] with [[polyethylene glycol]]<ref>{{cite journal|doi=10.1081/SEI-100001371|title=The Universal Solvent Extraction (Unex) Process. Ii. Flowsheet Development and Demonstration of the Unex Process for the Separation of Cesium, Strontium, and Actinides from Actual Acidic Radioactive Waste|year=2001|last1=Law|first1=Jack D.|last2=Herbst|first2=R. Scott|last3=Todd|first3=Terry A.|last4=Romanovskiy|first4=Valeriy N.|last5=Babain|first5=Vasily A.|last6=Esimantovskiy|first6=Vyatcheslav M.|last7=Smirnov|first7=Igor V.|last8=Zaitsev|first8=Boris N.|journal=Solvent Extraction and Ion Exchange|volume=19|page=23|s2cid=98103735}}</ref><ref>{{cite journal|doi=10.1081/SEI-100001370|title=The Universal Solvent Extraction (Unex) Process. I. Development of the Unex Process Solvent for the Separation of Cesium, Strontium, and the Actinides from Acidic Radioactive Waste|year=2001|last1=Romanovskiy|first1=Valeriy N.|last2=Smirnov|first2=Igor V.|last3=Babain|first3=Vasily A.|last4=Todd|first4=Terry A.|last5=Herbst|first5=R. Scott|last6=Law|first6=Jack D.|last7=Brewer|first7=Ken N.|journal=Solvent Extraction and Ion Exchange|volume=19|page=1|s2cid=98166395}}</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 diluents such as ''meta''-nitrobenzotri[[fluoride]]<ref>{{Cite conference |last=Smirnov |first=Igor |date=2–6 March 2014 |title=UNEX-T Solvent for Cs, Sr and Actinides Separation from PUREX Raffinate – 14154 |url=https://archivedproceedings.econference.io/wmsym/2014/papers/14154.pdf |conference=Waste Management Symposium |via=econference.io}}</ref> and phenyl trifluoromethyl [[sulfone]]<ref>{{cite web| url=http://www.wmsym.org/Abstracts/2001/62/62-7.pdf | archive-url=https://web.archive.org/web/20070928141036/http://www.wmsym.org/Abstracts/2001/62/62-7.pdf | archive-date=28 September 2007 | 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| publisher=WM 2001 conference proceedings| date=1 March 2001|access-date=17 June 2006|display-authors=etal}}</ref> have been suggested as well.

==== Electrochemical and ion exchange methods ====
An exotic method using [[electrochemistry]] and [[ion exchange]] in [[ammonium]] [[carbonate]] has been reported.<ref name=":0">{{cite journal|author=Asanuma, Noriko|last2=Harada|first2=Masayuki|last3=Nogami|first3=Masanobu|last4=Suzuki|first4=Kazunori|last5=Kikuchi|first5=Toshiaki|last6=Tomiyasu|first6=Hiroshi|last7=Ikeda|first7=Yasuhisa|display-authors=1|year=2006|title=Andodic dissociation of UO<sub>2</sub> pellet containing simulated fission products in ammonium carbonate solution|url=http://www.jstage.jst.go.jp/article/jnst/43/3/255/_pdf|journal=Journal of Nuclear Science and Technology|volume=43|issue=3|pages=255–262|doi=10.3327/jnst.43.255|doi-access=free}}{{dead link|date=October 2017|bot=medic}}{{cbignore|bot=medic}}</ref> Other methods for the extraction of uranium using ion exchange in alkaline carbonate and "fumed" lead oxide have also been reported.<ref>{{Cite patent|country=US|number=4366126|title=Recovery of uranium from uranium bearing solutions containing molybdenum|pubdate=1982-12-28|inventor1-last=Gardner|inventor1-first=Harry E.|assign1=[[Union Carbide|Union Carbide Corp.]]}}</ref>

====Obsolete methods====
===== Bismuth phosphate =====
The [[bismuth phosphate process]] 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 rod]]s, 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 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 [[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]]) to produce PuO<sub>2</sub><sup>2+</sup>. The plutonium 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 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.<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 |access-date=7 January 2006 |archive-date=11 May 2006 |archive-url=https://web.archive.org/web/20060511030328/http://www.bonestamp.com/sgt/process.htm |url-status=live }}</ref>

===== Hexone or REDOX =====
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 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 Site#Separation facilities|Hanford S plant (REDOX plant)]] and has been replaced by the PUREX process.<ref>{{Cite patent|country=US|number=2950166|pubdate=1960-08-23|title=Method for separation of plutonium from uranium and fission products by solvent extraction|assign1=[[United States Atomic Energy Commission]]|inventor1-last=Seaborg|inventor1-first=Glenn T.|inventorlink1=Glenn_T._Seaborg|inventor2-last=Blaedel, Jr.|inventor2-first=Walter J.|inventor3-last=Walling|inventor3-first=Matthew T.}}</ref><ref>{{cite web |url=http://www.llnl.gov/tid/lof/documents/pdf/235702.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=15 April 1999 |publisher=Lawrence Livermore National Laboratory preprint |access-date=7 January 2006 |archive-date=29 November 2007 |archive-url=https://web.archive.org/web/20071129121200/http://www.llnl.gov/tid/lof/documents/pdf/235702.pdf |url-status=live }}</ref>

{{chem2|Pu(4+) + 4NO3(−) + 2S → [Pu(NO3)4S2]}}

===== 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]] in 1951-1964. This process has been replaced by PUREX, which was shown to be a superior technology for larger scale reprocessing.<ref>{{Cite book|last=Taylor|first=Robin|title=Reprocessing and Recycling of Spent Nuclear Fuel|publisher=Woodhead Publishing|year=2015}}</ref>

===== Sodium acetate =====
The sodium uranyl [[acetate]] process was used by the early Soviet nuclear industry to recover plutonium from irradiated fuel.<ref>{{cite journal |last=Foreman |first=Mark R. St J. |title=Reactor accident chemistry an update |journal=Cogent Chemistry |date=2018 |volume=4 |issue=1 |doi=10.1080/23312009.2018.1450944 |doi-access=free}}</ref> 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.