View source for Physics:Nuclear reprocessing

{{Short description|Chemical operations that separate fissile material from spent fuel to be recycled as new fuel}}
[[File:Aerial_view_Sellafield,_Cumbria_-_geograph.org.uk_-_50827.jpg|thumb|Sellafield nuclear reprocessing site, UK]]
'''Nuclear reprocessing''' is the chemical separation of fission products and [[Chemistry:Actinide|actinide]]s from [[Physics:Spent nuclear fuel|spent nuclear fuel]].<ref>Andrews, A. (27 March 2008). [https://fas.org/sgp/crs/nuke/RS22542.pdf Nuclear Fuel Reprocessing: U.S. Policy] {{Webarchive|url=https://web.archive.org/web/20160303171717/http://www.fas.org/sgp/crs/nuke/RS22542.pdf |date=3 March 2016 }}. CRS Report For Congress. Retrieved 25 March 2011</ref> Originally, reprocessing was used solely to extract plutonium for producing [[Engineering:Nuclear weapon|nuclear weapon]]s. With commercialization of [[Physics:Nuclear power|nuclear power]], the reprocessed plutonium was recycled back into MOX nuclear fuel for thermal reactors.<ref>MOX fuel can extend the energy extracted by about 12% but slightly reduces plutonium stocks. [http://www.world-nuclear.org/info/inf29.html Information from the World Nuclear Association about MOX] {{Webarchive|url=https://web.archive.org/web/20130301173649/http://www.world-nuclear.org/info/inf29.html |date=1 March 2013 }}</ref> The [[Physics:Reprocessed uranium|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 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 [[Physics:PUREX|PUREX]] process dominating. Alternatives include heating to drive off volatile elements, burning via oxidation, and [[Physics:Fluoride volatility|fluoride volatility]] (which uses extremely reactive [[Chemistry:Fluorine|Fluorine]]). Each process results in some form of refined nuclear product, with [[Physics:Radioactive waste|radioactive waste]] as a byproduct. Because this could allow for [[Physics:Weapons-grade nuclear material|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 [[Physics:Radioactive waste|waste]] volumes decreased.<ref name=bas2011>{{cite web |url=http://www.thebulletin.org/web-edition/features/managing-nuclear-spent-fuel-policy-lessons-10-country-study |title=Managing nuclear spent fuel: Policy lessons from a 10-country study |author=Harold Feiveson |year=2011 |work=Bulletin of the Atomic Scientists |display-authors=0 |access-date=8 July 2011 |archive-date=26 April 2012 |archive-url=https://web.archive.org/web/20120426011518/http://www.thebulletin.org/web-edition/features/managing-nuclear-spent-fuel-policy-lessons-10-country-study |url-status=live }}</ref> Nuclear fuel reprocessing is performed routinely in Europe, Russia, and Japan. In 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.<ref>{{cite journal |title=Adieu to nuclear recycling |doi=10.1038/460152b |year=2009|journal=Nature |volume=460 |issue=7252 |page=152 |bibcode = 2009Natur.460R.152. |pmid=19587715|doi-access=free }}</ref>  Not all nuclear fuel requires reprocessing; a [[Engineering:Breeder reactor|breeder reactor]] is not restricted to using recycled plutonium and uranium. It can employ all the [[Chemistry:Actinide|actinide]]s, closing the [[Physics:Nuclear fuel cycle|nuclear fuel cycle]] and potentially multiplying the energy extracted from [[Physics:Natural uranium|natural uranium]] by about 60 times.<ref>{{cite web |title=Supply of Uranium |url=http://world-nuclear.org/info/inf75.html |url-status=live |archive-url=https://web.archive.org/web/20130212223705/http://www.world-nuclear.org/info/inf75.html |archive-date=12 February 2013 |access-date=29 January 2010 |publisher=[[Organization:World Nuclear Association|World Nuclear Association]]}}</ref><ref>{{cite web |title=Fast Neutron Reactors |url=http://www.world-nuclear.org/info/inf98.html |url-status=live |archive-url=https://web.archive.org/web/20130224035726/http://www.world-nuclear.org/info/inf98.html |archive-date=24 February 2013 |access-date=11 March 2012 |publisher=World Nuclear Association}}</ref>

== Separated components and disposition ==
The potentially useful components dealt with in nuclear reprocessing comprise specific [[Chemistry:Actinide|actinide]]s (plutonium, uranium, and some minor actinides). The lighter [[Chemistry:Chemical element|elements]] components include fission products, activation products, and cladding.

{| class="wikitable"
|-
!material!!disposition
|-
|plutonium, [[Chemistry:Minor actinide|minor actinide]]s, [[Physics:Reprocessed uranium|reprocessed uranium]]||[[Physics:Nuclear fission|fission]] in [[Physics:Fast reactor|fast]], [[Physics:Nuclear fusion–fission hybrid|fusion fission hybrid]], or [[Physics:Subcritical reactor|subcritical reactor]] or use as [[Physics:MOX fuel|MOX fuel]]
|-
|reprocessed uranium, filters || less stringent storage as intermediate-level waste
|-
|long-lived fission and [[Physics:Activation product|activation product]]s ||[[Physics:Nuclear transmutation|nuclear transmutation]] or geological repository
|-
|medium-lived fission products <sup>137</sup>Cs and [[Physics:Strontium-90|<sup>90</sup>Sr]]||medium-term storage as high-level waste; [[Physics:Decay heat|decay heat]] could be used to drive a [[Physics:Stirling engine|Stirling engine]]
|-
|useful radionuclides, rare earths (lanthanides), and [[Chemistry:Noble metal|noble metal]]s|| industrial and medical uses
|-
| cladding, fission product [[Chemistry:Zirconium|zirconium]]||re-use for zircalloy cladding or storage as intermediate level waste
|}

==History==
The first large-scale nuclear reactors were built during [[Social:World War II|World War II]]. These reactors were designed for the production of plutonium for use in [[Engineering:Nuclear weapon|nuclear weapon]]s. The only reprocessing required, therefore, was the extraction of the [[Chemistry:Plutonium|plutonium]] (free of fission-product contamination) from the spent [[Physics:Natural uranium|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 [[Physics:Bismuth phosphate process|bismuth phosphate process]], was developed and tested at the [[Organization:Oak Ridge National Laboratory|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 [[Astronomy:Hanford Site|Hanford Site]], in 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.<ref>{{Cite journal|url=https://digital.library.unt.edu/ark:/67531/metadc871031/|title=PUREX PROCESS FOR PLUTONIUM AND URANIUM RECOVERY|first1=W. B.|last1=Lanham|first2=T. C.|last2=Runion|date=1 October 1949|website=Other Information: Decl. with deletions Apr. 18, 1960. Orig. Receipt Date: 31-DEC-60|doi=10.2172/4165457 |doi-access=free}}</ref> The [[Physics:PUREX|PUREX]] process is the current method of extraction. Separation plants were also constructed at Savannah River Site and a smaller plant at West Valley Reprocessing Plant which closed by 1972 because of its inability to meet new regulatory requirements.<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 |access-date=17 June 2007 |publisher=U.S. Department of Energy |date=February 1996 |archive-date=14 March 2021 |archive-url=https://web.archive.org/web/20210314191712/http://www.osti.gov/opennet/document/purecov/nfsrepo.html |url-status=live }}</ref>

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 the West Valley Reprocessing Plant in the United States.

In October 1976,<ref>[http://www.presidency.ucsb.edu/ws/?pid=6561#axzz1zILTm1BT Gerald Ford 28 October 1976 Statement on Nuclear Policy] {{Webarchive|url=https://web.archive.org/web/20180926112728/http://www.presidency.ucsb.edu/ws/?pid=6561#axzz1zILTm1BT |date=26 September 2018 }}. Retrieved 30 June 2012.</ref> 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 [[Biography:Jimmy Carter|Jimmy Carter]] banned the reprocessing of commercial reactor [[Physics:Spent nuclear fuel|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.<ref>{{cite journal|title=The Politics, Science, Environment, and common sense of Spent Nuclear Fuel Reprocessing 3 decades Later |author=Dr. Ned Xoubi |journal=Symposium on the Technology of Peaceful Nuclear Energy, Irbid, Jordan |year=2008 |url=http://www.ctaps.yu.edu.jo/stpne/Symposium-Presentationns/Xounbi-The%20politics%20science%20Environment%20and%20common%20sense%20of%20SNF%20Reprocessing.pdf |url-status=dead |archive-url=https://web.archive.org/web/20110516130917/http://www.ctaps.yu.edu.jo/stpne/Symposium-Presentationns/Xounbi-The%20politics%20science%20Environment%20and%20common%20sense%20of%20SNF%20Reprocessing.pdf |archive-date=16 May 2011 }}</ref><ref>{{Cite web | url=http://depletedcranium.com/why-you-cant-build-a-bomb-from-spent-fuel/ |title = Depleted Cranium » Blog Archive » Why You Can't Build a Bomb from Spent Fuel|archive-url = https://web.archive.org/web/20120204194157/http://depletedcranium.com/why-you-cant-build-a-bomb-from-spent-fuel/|archive-date = 4 February 2012}}</ref><ref>{{cite news|url=https://atomicinsights.com/proving-a-negative-why-modern-used-nuclear-fuel-cannot-be-used-to-make-a-weapon/|title=Proving a Negative – Why Modern Used Nuclear Fuel Cannot Be Used to Make a Weapon – Atomic Insights|date=17 February 2015|newspaper=Atomic Insights|access-date=4 April 2018|archive-date=7 January 2018|archive-url=https://web.archive.org/web/20180107031323/https://atomicinsights.com/proving-a-negative-why-modern-used-nuclear-fuel-cannot-be-used-to-make-a-weapon/|url-status=live}}</ref> 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.<ref>[https://fas.org/sgp/crs/nuke/RS22542.pdf Nuclear Fuel Reprocessing: U.S. Policy Development] {{Webarchive|url=https://web.archive.org/web/20160303171717/http://www.fas.org/sgp/crs/nuke/RS22542.pdf |date=3 March 2016 }}. (PDF). Retrieved 10 December 2011.</ref>

In March 1999, the U.S. Department of Energy (DOE) reversed its policy and signed a contract with a [[Organization:Consortium|consortium]] of Duke Energy, COGEMA, and Stone & Webster (DCS) to design and operate a [[Physics:MOX fuel|mixed oxide (MOX) fuel]] fabrication facility. Site preparation at the Savannah River Site (South Carolina) began in October 2005.<ref>[https://www.nrc.gov/materials/fuel-cycle-fac/mox/reactors.html Duke, Cogema, Stone & Webster (DCS) Reports sent to NRC] {{Webarchive|url=https://web.archive.org/web/20170623145215/https://www.nrc.gov/materials/fuel-cycle-fac/mox/reactors.html |date=23 June 2017 }}. Nrc.gov. Retrieved 10 December 2011.</ref> In 2011 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 the nuclear accident at Fukushima Daiichi.<ref>[https://www.nytimes.com/2011/04/11/us/11mox.html?_r=1 New Doubts About Turning Plutonium Into a Fuel] {{Webarchive|url=https://web.archive.org/web/20170911160019/http://www.nytimes.com/2011/04/11/us/11mox.html?_r=1 |date=11 September 2017 }}, 10 April 2011</ref>

==Separation technologies==
===Water and organic solvents===
====PUREX====
{{Main|Physics: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 [[Physics:Nuclear fuel|nuclear fuel]], to extract [[Chemistry:Uranium|uranium]] and [[Chemistry:Plutonium|plutonium]], independent of each other, from the [[Physics: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 [[Engineering: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====
=====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 [[Physics:Nuclear waste|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 [[Physics:Reprocessed uranium|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 [[Chemistry:Technetium|technetium]] are separated from each other and the other fission products and [[Chemistry:Actinide|actinide]]s. The key is the addition of [[Chemistry:Acetohydroxamic acid|acetohydroxamic acid]] (AHA) to the extraction and scrub sections of the process. The addition of AHA greatly diminishes the extractability of plutonium and [[Chemistry:Neptunium|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 [[Organization:Argonne National Laboratory|Argonne National Laboratory]] and is designed to remove the transuranic metals (Am/Cm) from waste. The idea is that by lowering the [[Physics: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 [[Chemistry:Solvation|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 [[Chemistry:Carbon|carbon]], [[Chemistry:Hydrogen|hydrogen]], [[Chemistry:Nitrogen|nitrogen]], and [[Chemistry:Oxygen|oxygen]]. Such an organic waste can be burned without the formation of acidic gases which could contribute to [[Earth:Acid rain|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.<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 [[Chemistry:Lanthanide|lanthanide]]s and trivalent minor [[Chemistry:Actinide|actinide]]s should be removed from the PUREX [[Chemistry:Raffinate|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 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 [[Place:Czech Republic|Czech Republic]]; it is designed to completely remove the most troublesome [[Physics:Radioisotopes|radioisotopes]] (Sr, Cs and minor actinides) from the raffinate remaining after the extraction of uranium and plutonium from used [[Physics:Nuclear fuel|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=0 |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 [[Chemistry:Caesium|caesium]] and [[Chemistry:Strontium|strontium]] with [[Chemistry:Polyethylene glycol|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 [[Chemistry:Cobalt|cobalt]] [[Chemistry:Carborane|carborane]] anion (known as chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and the [[Chemistry:Diluent|diluent]] is a polar aromatic such as [[Chemistry:Nitrobenzene|nitrobenzene]]. Other diluents such as ''meta''-nitrobenzotri[[Chemistry:Fluoride|fluoride]] and phenyl trifluoromethyl [[Chemistry:Sulfone|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=0}}</ref> have been suggested as well.

==== Electrochemical and Ion Exchange methods ====
An exotic method using [[Chemistry:Electrochemistry|electrochemistry]] and [[Chemistry:Ion exchange|ion exchange]] in [[Chemistry:Ammonium|ammonium]] [[Chemistry:Carbonate|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 Corp.}}</ref>

====Obsolete methods====
===== Bismuth phosphate =====
The [[Physics:Bismuth phosphate process|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 [[Chemistry:Plutonium|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 [[Chemistry:Nitric acid|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 [[Chemistry:Phosphoric acid|phosphoric acid]] to form the bismuth phosphate. The plutonium was [[Chemistry: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 [[Chemistry:Potassium permanganate|potassium permanganate]]) to produce PuO<sub>2</sub><sup>2+</sup>. The plutonium was maintained in the +6 [[Chemistry:Oxidation state|oxidation state]] by addition of a [[Chemistry:Dichromate|dichromate]] salt.

The bismuth phosphate was next re-precipitated, leaving the plutonium in solution, and an iron(II) salt (such as [[Chemistry:Ferrous sulfate|ferrous sulfate]]) was added. The plutonium was again re-precipitated using a bismuth phosphate carrier and a combination of [[Chemistry:Lanthanum|lanthanum]] salts and [[Chemistry:Fluoride|fluoride]] added, forming a solid lanthanum fluoride carrier for the plutonium. Addition of an [[Chemistry:Alkali|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 [[Chemistry:Methyl isobutyl ketone|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 ([[Chemistry:Aluminium nitrate|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 [[Astronomy:Hanford Site#Separation facilities|Hanford plant T]] 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 [[Chemistry:Nitrate|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 [[Chemistry:Acetate|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 a never used in the West; the idea is to dissolve the fuel in [[Chemistry:Nitric acid|nitric acid]], alter the oxidation state of the plutonium, and then add [[Chemistry:Acetic acid|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.

==Alternatives to PUREX==
As there are some downsides to 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.<ref>{{cite web |title=Processing of Used Nuclear Fuel |url=https://world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/processing-of-used-nuclear-fuel.aspx |publisher=World Nuclear Association |access-date=4 October 2022 |archive-url=https://web.archive.org/web/20220928003928/https://www.world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/processing-of-used-nuclear-fuel.aspx |archive-date=28 September 2022 |language=en |date=December 2020 |url-status=live}}</ref>

=== Pyroprocessing ===

[[File:Ifr concept.jpg|thumb|upright=1.3|The most developed, though commercially unfielded, alternative reprocessing method, is Pyroprocessing,<ref>{{cite journal|author=L.C. Walters|date=18 September 1998|title=Thirty years of fuels and materials information from EBR-II|journal=Journal of Nuclear Materials|volume=270|issue=1|pages=39–48|doi=10.1016/S0022-3115(98)00760-0|bibcode=1999JNuM..270...39W|url=https://zenodo.org/record/1259635|access-date=17 March 2021|archive-date=31 January 2021|archive-url=https://web.archive.org/web/20210131115425/https://zenodo.org/record/1259635|url-status=live}}</ref> suggested as part of the depicted metallic-fueled, Integral fast reactor (IFR) a sodium fast reactor concept of the 1990s.  After the spent fuel is dissolved in molten salt, all of the recyclable actinides, consisting largely of plutonium and uranium though with important minor constituents, are extracted using electrorefining/electrowinning.  The resulting mixture keeps the plutonium at all times in an unseparated gamma and alpha emitting actinide form, that is also mildly self-protecting in theft scenarios.<ref>{{cite web|url=https://www.aps.org/units/fps/newsletters/2004/july/hannum.html|archive-url=https://web.archive.org/web/20200805184720/https://www.aps.org/units/fps/newsletters/2004/july/hannum.html |archive-date=2020-08-05|title=APS - Physics and Society Newsletter - July 2004 - PUREX AND PYRO ARE NOT THE SAME|access-date=2023-08-09}} PUREX and PYRO are not the same, Hannum, Marsh, Stanford.</ref>]]
[[Chemistry:Pyroprocessing|Pyroprocessing]] is a generic term for high-temperature methods. Solvents are [[Chemistry:Molten salt|molten salt]]s (e.g. LiCl&nbsp;+&nbsp;KCl or LiF&nbsp;+&nbsp;CaF<sub>2</sub>) and molten metals (e.g. cadmium, bismuth, magnesium) rather than water and organic compounds. Electrorefining, [[Distillation|distillation]], and solvent-solvent extraction are common steps.

These processes are not currently in significant use worldwide, but they have been pioneered at [[Organization:Argonne National Laboratory|Argonne National Laboratory]]<ref>{{cite web |url=http://www.ne.anl.gov/nce/pyroprocessing/ |title=Pyroprocessing Development |publisher=Argonne National Laboratory |access-date=6 June 2016 |archive-date=24 June 2016 |archive-url=https://web.archive.org/web/20160624220930/http://www.ne.anl.gov/nce/pyroprocessing/ |url-status=live }}</ref><ref>{{cite web
 |year=2012
 |title=Pyroprocessing Technologies: Recycling used nuclear fuel for a sustainable energy future
 |url=http://www.ne.anl.gov/pdfs/12_Pyroprocessing_bro_5_12_v14[6].pdf
 |archive-url=https://web.archive.org/web/20130219051536/http://www.ne.anl.gov/pdfs/12_Pyroprocessing_bro_5_12_v14%5B6%5D.pdf
 |url-status=dead
 |archive-date=19 February 2013
|pages=7
 |publisher=[[Organization:Argonne National Laboratory|Argonne National Laboratory]]
 |access-date=6 June 2016
}}</ref> with current research also taking place at CRIEPI in Japan, the Nuclear Research Institute of Řež in Czech Republic, [[Organization:Indira Gandhi Centre for Atomic Research|Indira Gandhi Centre for Atomic Research]] in India and [[Organization:Korea Atomic Energy Research Institute|KAERI]] in South Korea.<ref>{{cite web|url=https://www.oecd-nea.org/pt/iempt9/Nimes_Presentations/INOUE.pdf|title=An Overview of CRIEPI Pyroprocessing Activities|author=T. Inoue|access-date=20 May 2019|archive-date=13 July 2017|archive-url=https://web.archive.org/web/20170713211506/http://www.oecd-nea.org/pt/iempt9/Nimes_Presentations/INOUE.pdf|url-status=dead}}</ref><ref>Tulackova, R., et al. "Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle." Atom Indonesia 33.1 (2007): 47–59.</ref><ref>Nagarajan, K., et al. "Current status of pyrochemical reprocessing research in India." Nuclear Technology 162.2 (2008): 259–263.</ref><ref>Lee, Hansoo, et al. "Development of Pyro-processing Technology at KAERI." (2009).</ref>

====Advantages of pyroprocessing====
* The principles behind it are well understood, and no significant technical barriers exist to their adoption.<ref>{{cite web |url=http://www.inl.gov/technicalpublications/Documents/3931942.pdf |title=PYROPROCESSING PROGRESS AT IDAHO NATIONAL LABORATORY |date=September 2007 |publisher=Idaho National Laboratory article |url-status=dead |archive-url=https://web.archive.org/web/20110612014703/http://www.inl.gov/technicalpublications/Documents/3931942.pdf |archive-date=12 June 2011}}</ref>
* Readily applied to high-[[Engineering:Burnup|burnup]] spent fuel and requires little cooling time, since the [[Physics:Operating temperature|operating temperature]]s are high already.
* Does not use solvents containing hydrogen and carbon, which are [[Physics:Neutron moderator|neutron moderator]]s creating risk of [[Physics:Criticality accident|criticality accident]]s and can absorb the fission product [[Physics:Tritium|tritium]] and the [[Physics:Activation product|activation product]] [[Physics:Carbon-14|carbon-14]] in dilute solutions that cannot be separated later.
**Alternatively, voloxidation<ref name=advancedheadend /> (see below) can remove 99% of the tritium from used fuel and recover it in the form of a strong solution suitable for use as a supply of tritium.
* More compact than aqueous methods, allowing on-site reprocessing at the reactor site, which avoids transportation of spent fuel and its security issues, instead storing a much smaller volume of fission products on site as high-level waste until [[Physics:Nuclear decommissioning|decommissioning]]. For example, the Integral Fast Reactor and Molten Salt Reactor fuel cycles are based on on-site pyroprocessing.
* It can separate many or even all [[Chemistry:Actinide|actinide]]s at once and produce highly radioactive fuel which is harder to manipulate for theft or making nuclear weapons. (However, the difficulty has been questioned.<ref>{{cite web|url=http://www.princeton.edu/sgs/publications/sgs/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|access-date=25 April 2011|archive-date=26 March 2013|archive-url=https://web.archive.org/web/20130326063648/http://www.princeton.edu/sgs/publications/sgs/pdf/13_3%20Kang%20vonhippel.pdf|url-status=live}}</ref>) In contrast the PUREX process was designed to separate plutonium only for weapons, and it also leaves the [[Chemistry:Minor actinide|minor actinide]]s ([[Chemistry:Americium|americium]] and [[Chemistry:Curium|curium]]) behind, producing waste with more long-lived radioactivity.
* Most of the radioactivity in roughly 10<sup>2</sup> to 10<sup>5</sup> years after the use of the nuclear fuel is produced by the actinides, since there are no fission products with half-lives in this range. These actinides can fuel [[Physics:Fast reactor|fast reactor]]s, so extracting and reusing (fissioning) them increases energy production per kg of fuel, as well as reducing the long-term radioactivity of the wastes.
* [[Physics:Fluoride volatility|Fluoride volatility]] (see below) produces salts that can readily be used in molten salt reprocessing such as pyroprocessing
* The ability to process "fresh" spent fuel reduces the needs for [[Engineering:Spent fuel pool|spent fuel pool]]s (even if the recovered short lived radionuclides are "only" sent to storage, that still requires less space as the bulk of the mass, uranium, can be stored separately from them). Uranium – even higher specific activity [[Physics:Reprocessed uranium|reprocessed uranium]] – does not need cooling for safe storage.
* Short lived radionuclides can be recovered from "fresh" spent fuel allowing either their direct use in industry science or medicine or the recovery of their decay products without contamination by other isotopes (for example: ruthenium in spent fuel decays to rhodium all isotopes of which other than {{chem|103|Rh}} further decay to stable [[Physics:Isotopes of palladium|isotopes of palladium]]. Palladium derived from the decay of fission ruthenium and rhodium will be nonradioactive, but fission Palladium contains significant contamination with long-lived {{chem|107|Pd| link=Palladium-107}}. Ruthenium-107 and rhodium-107 both have half lives on the order of minutes and decay to palladium-107 before reprocessing under most circumstances)
* Possible fuels for [[Physics:Radioisotope thermoelectric generator|radioisotope thermoelectric generator]]s (RTGs) that are mostly decayed in spent fuel, that has significantly aged, can be recovered in sufficient quantities to make their use worthwhile. Examples include materials with half lives around two years such as {{chem|134|Cs| link=Caesium-134}}, {{chem|125|Sb| link=Antimony-125}}, {{chem|147|Pm| link=Promethium-147}}. While those would perhaps not be suitable for lengthy space missions, they can be used to replace [[Engineering:Diesel generator|diesel generator]]s in off-grid locations where refueling is possible once a year.{{efn|a radioisotope with a two year half life will retain 0.5^0.5 or over 70% of its power after a year - all those isotopes have half lives longer than two years and would thus retain even more power. Even if the yearly refueling window were to be missed, over half the power would still remain for the second refueling window}} Antimony would be particularly interesting because it forms a stable alloy with lead and can thus be transformed relatively easily into a partially self-shielding and chemically inert form. Shorter lived RTG fuels present the further benefit of reducing the risk of [[Physics:Orphan source|orphan source]]s as the activity will decline relatively quickly if no refueling is undertaken.

====Disadvantages of pyroprocessing====
* 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 pyrometallurgical systems, although there could be if the [[Engineering:Generation IV reactor|Generation IV reactor]] programs become reality.
* The used salt from pyroprocessing is less suitable for conversion into glass than the waste materials produced by the PUREX process.
* If the goal is to reduce the longevity of spent nuclear fuel in burner reactors, then better recovery rates of the minor actinides need to be achieved.
* Working with "fresh" spent fuel requires more shielding and better ways to deal with heat production than working with "aged" spent fuel does. If the facilities are built in such a way as to ''require'' high specific activity material, they cannot handle older "legacy waste" except blended with fresh spent fuel

===Electrolysis===
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 the cathode out of the molten salt solution before the other elements.<ref>Morss, L. R. The chemistry of the actinide and transactinide elements. Eds. Lester R. Morss, et al. Vol. 1. Dordrecht: Springer, 2006.</ref>

[[File:Electrorefining technology anl gov.jpg|thumb|Experimental electro refinement cell at Argonne National Laboratory]]

====PYRO-A and -B for IFR====
These processes were developed by [[Organization:Argonne National Laboratory|Argonne National Laboratory]] and used in the Integral Fast Reactor project.

'''PYRO-A''' is a means of separating actinides (elements within the [[Chemistry:Actinide|actinide]] family, generally heavier than U-235) from non-actinides. The spent fuel is placed in an [[Physics:Anode|anode]] [[Engineering:Basket|basket]] which is immersed in a molten salt electrolyte. An electric 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 [[Chemistry:Cadmium|cadmium]] cathode. Many of the fission products (such as [[Chemistry:Caesium|caesium]], [[Chemistry:Zirconium|zirconium]] and [[Chemistry:Strontium|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 |access-date=22 June 2009 |url-status=dead |archive-url=https://web.archive.org/web/20090225154203/http://criepi.denken.or.jp/en/e_publication/pdf/den363.pdf |archive-date=25 February 2009 }}</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=12 December 2001|access-date=22 June 2009|archive-date=5 September 2009|archive-url=https://web.archive.org/web/20090905052041/http://www.nea.fr/html/pt/docs/iem/madrid00/Proceedings/Paper56.pdf|url-status=dead}}</ref><ref>R. Tulackova (Zvejskova), K. Chuchvalcova Bimova, P. Soucek, F. Lisy [http://www.nea.fr/html/pt/iempt8/abstracts/Abstracts/Session_II/zvejskova.ppt Study of Electrochemical Processes for Separation of the Actinides and Lanthanides in Molten Fluoride Media] {{Webarchive|url=https://web.archive.org/web/20090905052044/http://www.nea.fr/html/pt/iempt8/abstracts/Abstracts/Session_II/zvejskova.ppt |date=5 September 2009 }} (PPT file). Nuclear Research Institute Rez plc, Czech Republic</ref> As alternatives to the molten cadmium electrode it is possible to use a molten [[Chemistry:Bismuth|bismuth]] cathode, or a solid aluminium cathode.<ref>[http://www.nea.fr/html/pt/docs/iem/jeju02/session2/SessionII-06.pdf Electrochemical Behaviours of Lanthanide Fluorides in the Electrolysis System with LiF-NaF-KF Salt] {{Webarchive|url=https://web.archive.org/web/20090905052049/http://www.nea.fr/html/pt/docs/iem/jeju02/session2/SessionII-06.pdf |date=5 September 2009 }}. (PDF) . Retrieved 10 December 2011.</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 {{Dead link|date=February 2022}}</ref> -->

As an alternative to electrowinning, the wanted metal can be isolated by using a molten [[Chemistry:Alloy|alloy]] of an electropositive metal and a less reactive metal.<ref>[https://web.archive.org/web/20050123045408/http://www.merck.de/servlet/PB/show/1332930/10.Molten%20Salts%20Lanthanides.pdf Ionic Liquids/Molten Salts and Lanthanides/Actinides Reference List]. Merck.de. Retrieved 10 December 2011.</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 of, years.<ref>{{cite web
 | url= http://www.ne.doe.gov/AFCI/neAFCI.html
 | title= Advanced Fuel Cycle Initiative
 | publisher= U.S. Department of Energy
 | access-date= 3 May 2008
| url-status= dead
 | archive-url= https://web.archive.org/web/20120510084252/http://www.ne.doe.gov/AFCI/neAFCI.html
 | archive-date= 10 May 2012 }}</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 to be burned efficiently. In a thermal neutron spectrum, the concentrations of several heavy actinides ([[Chemistry:Curium|curium]]-242 and [[Physics:Plutonium-240|plutonium-240]]) can become quite high, creating fuel that 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 is free from uranium and contains recovered transuranics in an inert matrix such as metallic [[Chemistry:Zirconium|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 [[Chemistry:Ozone|ozone]] to [[Chemistry:Uranium trioxide|uranium trioxide]] with decomposition by heating back to [[Chemistry:Triuranium octoxide|triuranium octoxide]].<ref name="advancedheadend"/> A major purpose is to capture [[Physics:Tritium|tritium]] as tritiated water vapor before further processing where it would be difficult to retain 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 [[Chemistry:Iodine|iodine]], [[Chemistry:Technetium|technetium]], and [[Physics:Carbon-14|carbon-14]]. Voloxidation also breaks up the fuel or increases its surface area to enhance penetration of reagents in following reprocessing steps.

==== Advantages ====
* The process is simple and requires no complex machinery or chemicals above and beyond that required in all reprocessing ([[Physics:Hot cell|hot cell]]s, remote handling equipment)
* Products such as [[Physics:Krypton-85|krypton-85]] or tritium, as well as [[Chemistry:Xenon|xenon]] (whose isotope are either stable, [[Physics:Xenon-136|very nearly stable]], or quickly decay), can be recovered and sold for use in industry, science or medicine 
* Driving off volatile fission products allows for safer storage in interim storage or [[Earth:Deep geological repository|deep geological repository]]
* Nuclear proliferation risks are low as no separation of plutonium occurs
* Radioactive material is not chemically mobilized beyond what should be accounted for in long-term storage anyway. Substances that are inert as native elements or oxides remain so
* The product can be used as fuel in a CANDU reactor or even downblended with similarly treated spent CANDU fuel if too much fissile material is left in the spent fuel.
* The resulting product can be further processed by any of the other processes mentioned above and below. Removal of volatile fission products means that transportation becomes slightly easier compared to spent fuel with damaged or removed cladding
* All volatile products of concern (while helium will be present in the spent fuel, there won't be any radioactive [[Physics:Isotopes of helium|isotopes of helium]]) can in principle be recovered in a [[Physics:Cold trap|cold trap]] cooled by [[Chemistry:Liquid nitrogen|liquid nitrogen]] (temperature: {{convert|77|K}} or lower). However, this requires significant amounts of cooling to counteract the effect of decay heat from radioactive volatiles like krypton-85. Tritium will be present in the form of [[Chemistry:Tritiated water|tritiated water]], which is a solid at the temperature of liquid nitrogen.
* Technetium heptoxide can be removed as a gas by heating above its boiling point of {{convert|392.6|K}} which reduces the issues presented by Technetium contamination in processes like fluoride volatility or PUREX; [[Chemistry:Ruthenium tetroxide|ruthenium tetroxide]] (gaseous above {{convert|313.1|K}}) can likewise be removed from the spent fuel and recovered for sale or disposal

====Disadvantages====
* Further processing is needed if the resulting product is to be used for re-enrichment or fabrication of MOX-fuel
* If volatile fission products escape to the environment this presents a radiation hazard, mostly due to {{chem|129|I|link=Iodine-129}}, Tritium and {{chem|85|Kr|link=Krypton-85}}. Their safe recovery and storage requires further equipment.
* An [[Chemistry:Oxidizing agent|oxidizing agent]] / [[Chemistry:Reducing agent|reducing agent]] has to be used for reduction/oxidation steps whose recovery can be difficult, energy consuming or both

===Volatilization in isolation===
Simply heating spent oxide fuel in an inert atmosphere or vacuum at a temperature between {{convert|700|°C}} and {{convert|1000|°C}} as a first reprocessing step can remove several volatile elements, including caesium whose isotope [[Physics:Caesium-137|caesium-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 [[Physics:Strontium-90|strontium-90]], which has a similar half-life).
The estimated overall mass balance for 20,000&nbsp;g of processed fuel with 2,000&nbsp;g of cladding is:<ref>{{cite web |url=http://web.mac.com/mosb1000/iWeb/Bob's%20Site/Examples_files/Sr_Design_Rpt.pdf |title=Removal of caesium from spent nuclear fuel destined for the electrorefiner fuel treatment process |author=Wolverton, Daren |publisher=University of Idaho (dissertation?) |date=11 May 2005 |display-authors=0 |url-status=dead |archive-url=https://web.archive.org/web/20071129121201/http://web.mac.com/mosb1000/iWeb/Bob%27s%20Site/Examples_files/Sr_Design_Rpt.pdf |archive-date=29 November 2007}}</ref>

{| class="wikitable"
|-
! !!Input !!Residue !![[Chemistry:Zeolite|Zeolite]]<br />filter!!Carbon<br />filter!!Particle<br />filters
|-
|[[Chemistry:Palladium|Palladium]]||28||14||14|| ||
|-
|[[Chemistry:Tellurium|Tellurium]]||10||5||5|| ||
|-
|[[Chemistry:Molybdenum|Molybdenum]]||70|| ||70|| ||
|-
|[[Chemistry:Caesium|Caesium]]||46|| ||46|| ||
|-
|[[Chemistry:Rubidium|Rubidium]]||8|| ||8|| ||
|-
|Silver||2|| ||2|| ||
|-
|[[Chemistry:Iodine|Iodine]]||4|| || ||4||
|-
|Cladding||2000||2000|| || ||
|-
|[[Chemistry:Uranium|Uranium]]||19218||19218|| || ||?
|-
|Others||614||614|| || ||?
|-
|Total||22000||21851||145||4||0
|}

====Advantages====
* Requires no chemical processes at all
* Can in theory be done "self heating" via the [[Physics:Decay heat|decay heat]] of sufficiently "fresh" spent fuel
* [[Physics:Caesium-137|Caesium-137]] has uses in [[Physics:Food irradiation|food irradiation]] and can be used to power [[Physics:Radioisotope thermoelectric generator|radioisotope thermoelectric generator]]s. However, its contamination with stable {{chem|133|Cs}} and long lived {{chem|135|Cs}} reduces efficiency of such uses while contamination with {{chem|134|Cs}} in relatively fresh spent fuel makes the curve of overall radiation and heat output much steeper until most of the {{chem|134|Cs}} has decayed
* Can potentially recover elements like [[Chemistry:Ruthenium|ruthenium]] whose ruthenate ion is particularly troublesome in PUREX and which has no isotopes significantly longer lived than a year, allowing possible recovery of the metal for use
* A "third phase recovery" can be added to the process if substances that melt but don't vaporize at the temperatures involved are drained to a container for liquid effluents and allowed to re-solidify. To avoid contamination with low-boiling products which melt at low temperatures, a melt plug could be used to open the container for liquid effluents only once a certain temperature is reached by the liquid phase.
* Strontium, which is present in the form of the particularly troublesome mid-lived fission product {{chem|90|Sr}} is liquid above {{convert|1050|K}}. However, [[Chemistry:Strontium oxide|Strontium oxide]] remains solid below {{convert|2804|K}} and if strontium oxide is to be recovered with other liquid effluents, it has to be reduced to the native metal before the heating step. Both Strontium and Strontium oxide form soluble [[Chemistry:Strontium hydroxide|Strontium hydroxide]] and hydrogen upon contact with water, which can be used to separate them from non-soluble parts of the spent fuel.
* As there are little to no chemical changes in the spent fuel, any chemical reprocessing methods can be used following this process

====Disadvantages====
* At temperatures above {{convert|1000|K}} the native metal form of several [[Chemistry:Actinide|actinide]]s, including [[Chemistry:Neptunium|neptunium]] (melting point: {{convert|912|K}}) and [[Chemistry:Plutonium|plutonium]] (melting point: {{convert|912.5|K}}), are molten. This could be used to recover a liquid phase, raising proliferation concerns, given that uranium metal remains a solid until {{convert|1405.3|K}}. While neptunium and plutonium cannot be easily separated from each other by different melting points, their differing solubility in water can be used to separate them.
* If "nuclear self heating" is employed, the spent fuel with have much higher [[Physics:Specific activity|specific activity]], heat production and radiation release. If an external heat source is used, significant amounts of external power are needed, which mostly go to heat the uranium.
* Heating and cooling the vacuum chamber and/or the piping and vessels to collect volatile effluents induces [[Physics:Thermal stress|thermal stress]]. This combines with radiation damage to material and possibly [[Physics:Neutron embrittlement|neutron embrittlement]] if [[Physics:Neutron source|neutron source]]s such as [[Physics:Californium-252|californium-252]] are present to a significant extent.
* In the commonly used oxide fuel, some elements will be present both as oxides and as native elements. Depending on their chemical state, they may end up in either the volatalized stream or in the residue stream. If an element is present in both states to a significant degree, separation of that element may be impossible without converting it all to one chemical state or the other
* The temperatures involved are much higher than the melting point of lead ({{convert|600.61|K}}) which can present issues with radiation shielding if lead is employed as a shielding material
* If filters are used to recover volatile fission products, those become low- to intermediate level waste.

=== Fluoride volatility ===
{{Main|Physics:Fluoride volatility}}
[[File:Fission yield volatile 2.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 or their oxides 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→Rb, Sr-90→Zr, Ru-106→Pd, Sb-125→Te, Cs-137→Ba, Ce-144→Nd, Sm-151→Eu, Eu-155→Gd visible.]]
In the fluoride volatility process, [[Chemistry:Fluorine|fluorine]] is reacted with the fuel. Fluorine is so much more reactive than even [[Chemistry:Oxygen|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 [[Chemistry:Uranium|uranium]], which makes up the bulk of the fuel, is converted to [[Chemistry:Uranium hexafluoride|uranium hexafluoride]], the form of uranium used in uranium enrichment, which has a very low boiling point. [[Chemistry:Technetium|Technetium]], the main [[Physics:Long-lived fission product|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|fractional distillation]] or selective [[Chemistry:Redox|reduction]]. [[Chemistry:Uranium hexafluoride|Uranium hexafluoride]] and [[Chemistry:Technetium hexafluoride|technetium hexafluoride]] have very similar boiling points and vapor pressures, which makes complete separation more difficult.

Many of the fission products volatilized are the same ones volatilized in non-fluorinated, higher-temperature volatilization, such as [[Chemistry:Iodine|iodine]], [[Chemistry:Tellurium|tellurium]] and [[Chemistry:Molybdenum|molybdenum]]; notable differences are that [[Chemistry:Technetium|technetium]] is volatilized, but [[Chemistry:Caesium|caesium]] is not.

Some transuranium elements such as [[Chemistry:Plutonium|plutonium]], [[Chemistry:Neptunium|neptunium]] and [[Chemistry:Americium|americium]] can form volatile fluorides, but these compounds are not stable when the fluorine partial pressure is decreased.<ref>{{Cite book| url = https://books.google.com/books?id=SJOE00whg44C&pg=PA66| title = The radiochemistry of nuclear power plants with light water reactors| isbn = 978-3-11-013242-7| author = Neeb, Karl-Heinz| publisher = Walter de Gruyter| year = 1997| access-date = 29 November 2021| archive-date = 25 January 2022| archive-url = https://web.archive.org/web/20220125095656/https://books.google.com/books?id=SJOE00whg44C&pg=PA66| url-status = live}}</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, [[Chemistry:Neptunium|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 [[Chemistry:Ruthenium hexafluoride|ruthenium hexafluoride]] is relatively stable and volatile.

Distillation of the residue at higher temperatures can separate lower-boiling [[Chemistry:Transition metal|transition metal]] fluorides and [[Chemistry:Alkali metal|alkali metal]] (Cs, Rb) fluorides from higher-boiling [[Chemistry:Lanthanide|lanthanide]] and [[Chemistry:Alkaline earth metal|alkaline earth metal]] (Sr, Ba) and [[Chemistry:Yttrium|yttrium]] fluorides. The temperatures involved are much higher, but can be lowered somewhat by distilling in a vacuum. If a carrier salt like [[Chemistry:Lithium fluoride|lithium fluoride]] or [[Chemistry:Sodium fluoride|sodium fluoride]] is being used as a solvent, high-temperature distillation is a way to separate the carrier salt for reuse.

[[Engineering:Molten salt reactor|Molten salt reactor]] designs carry out fluoride volatility reprocessing continuously or at frequent intervals. The goal is to return [[Chemistry:Actinide|actinide]]s to the molten fuel mixture for eventual fission, while removing fission products that are [[Chemistry:Neutron poison|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-[[Chemistry:Valence|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 {{convert|331|°C}} and {{convert|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 |title=Advanced Head-End Processing of Spent Fuel: A Progress Report |author=Guillermo D. Del Cul |work=2005 ANS annual meeting |publisher=[[Organization:Oak Ridge National Laboratory|Oak Ridge National Laboratory]], U.S. DOE |access-date=3 May 2008 |display-authors=0 |url-status=dead |archive-url=https://web.archive.org/web/20060307211536/http://www.ornl.gov/~webworks/cppr/y2001/pres/123514.pdf |archive-date=7 March 2006}}</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 [[Chemistry:Americium|americium]], [[Chemistry:Curium|curium]], lanthanides, [[Chemistry:Strontium|strontium]], [[Chemistry:Caesium|caesium]] are more soluble than those of [[Chemistry:Uranium|uranium]], [[Chemistry:Neptunium|neptunium]], [[Chemistry:Plutonium|plutonium]], and [[Chemistry:Zirconium|zirconium]].

====Advantages of halogen volatility====
* Chlorine (and to a lesser extent fluorine<ref>{{cite web |title=Fluorine |url=https://www.essentialchemicalindustry.org/chemicals/fluorine.html |website=essentialchemicalindustry.org |access-date=4 October 2022 |archive-url=https://web.archive.org/web/20220425062623/https://www.essentialchemicalindustry.org/chemicals/fluorine.html |archive-date=25 April 2022 |date=10 October 2016 |url-status=live}}</ref>) is a readily available industrial chemical that is produced in mass quantity<ref>{{cite web |title=Chlorine Manufacturing Industry in the US |url=https://www.ibisworld.com/united-states/market-research-reports/chlorine-manufacturing-industry/ |website=ibisworld.com |access-date=4 October 2022 |archive-url=https://web.archive.org/web/20220223185750/https://www.ibisworld.com/united-states/market-research-reports/chlorine-manufacturing-industry/ |archive-date=2022-02-23 |language=en-US |date=28 Jun 2022 |url-status=live}}</ref>
* Fractional distillation allows many elements to be separated from each other in a single step or iterative repetition of the same step
* Uranium will be produced directly as [[Chemistry:Uranium hexafluoride|Uranium hexafluoride]], the form used in enrichment
* Many volatile fluorides and chlorides are volatile at relatively moderate temperatures reducing thermal stress. This is especially important as the boiling point of uranium hexafluoride is below that of water, allowing to conserve energy in the separation of high boiling fission products (or their fluorides) from one another as this can take place in the absence of uranium, which makes up the bulk of the mass
* Some fluorides and chlorides melt at relatively low temperatures allowing a "liquid phase separation" if desired. Those low melting salts could be further processed by molten salt electrolysis.
* Fluorides and chlorides differ in water solubility depending on the cation. This can be used to separate them by aqueous solution. However, some fluorides violently react with water, which has to be taken into account.

====Disadvantages of halogen volatility====
* Many compounds of fluorine or chlorine as well as the native elements themselves are toxic, corrosive and react violently with air, water or both
* [[Chemistry:Uranium hexafluoride|Uranium hexafluoride]] and [[Chemistry:Technetium hexafluoride|Technetium hexafluoride]] have very similar boiling points ({{convert|329.6|K}} and {{convert|328.4|K}} respectively), making it hard to completely separate them from one another by distillation.
* Fractional distillation as used in petroleum refining requires large facilities and huge amounts of energy. To process tons of uranium would require similarly large facilities as processing tons of petroleum - however, unlike petroleum refineries, the entire process would have to take place inside radiation shielding and there would have to be provisions made to prevent leaks of volatile, poisonous and radioactive fluorides
* [[Chemistry:Plutonium hexafluoride|Plutonium hexafluoride]] boils at {{convert|335|K}} this means that any facility capable of separating uranium hexafluoride from Technetium hexafluoride is capable of separating plutonium hexafluoride from either, raising proliferation concerns
* The presence of [[Physics:Alpha decay|alpha emitters]] induces some (α,n) reactions in fluorine, producing both radioactive {{chem|22|Na|link=sodium-22}} and neutrons.<ref>{{cite journal|url=https://www.sciencedirect.com/science/article/abs/pii/S0969806X23001640|title=Neutron and gamma-ray signatures for the control of alpha-emitting materials in uranium production: A Nedis2m-MCNP6 simulation|date=2023 |doi=10.1016/j.radphyschem.2023.110919 |access-date=2023-08-09 |last1=Vlaskin |first1=Gennady N. |last2=Bedenko |first2=Sergey V. |last3=Polozkov |first3=Sergey D. |last4=Ghal-Eh |first4=Nima |last5=Rahmani |first5=Faezeh |journal=Radiation Physics and Chemistry |volume=208 |page=110919 |s2cid=257588532 }}</ref> This effect can be reduced by separating alpha emitters and fluorine as fast as feasible. Interactions between chlorine's two stable isotopes {{chem|35|Cl|link=chlorine-35}} and {{chem|37|Cl|link=chlorine-37}} on the one hand and alpha particles on the other are of lesser concern as they do not have as high a cross section and do not produce neutrons or long lived radionuclides.<ref>[https://www.oecd-nea.org/janisweb/book/alphas/Cl35/MT4/renderer/1082 Dead link] {{Dead link|date=September 2022}}</ref>
* If carbon is present in the spent fuel it'll form halogenated hydrocarbons which are extremely potent [[Chemistry:Greenhouse gas|greenhouse gas]]es, and hard to chemically decompose. Some of those are toxic as well.

===Radioanalytical separations===
To determine 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).<ref>{{cite journal | last1 = Kabay | first1 = N. | last2 = Cortina | first2 = J.L. | last3 = Trochimczuk | first3 = A. | last4 = Streat | first4 = M. | year = 2010 | title = Solvent-impregnated resins (SIRs) – Methods of preparation and their applications | journal = React. Funct. Polym. | volume = 70 | issue = 8| pages = 484–496 | doi=10.1016/j.reactfunctpolym.2010.01.005| hdl = 2117/10365 }}</ref>

== Economics ==
The relative economics of reprocessing-waste disposal and interim storage-direct disposal was the focus of much debate over the first decade of the 2000s. Studies<ref>{{cite web
 |url         = http://www.inl.gov/technicalpublications/Documents/4536700.pdf
 |title       = Advanced Fuel Cycle Cost Basis
 |publisher   = Idaho National Laboratory, United States Department of Energy
 |access-date  = 19 December 2010
|url-status     = dead
 |archive-url  = https://web.archive.org/web/20111128060259/http://www.inl.gov/technicalpublications/Documents/4536700.pdf
 |archive-date = 28 November 2011}}</ref>
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 [[Engineering:Breeder reactor|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 agreed that under then-current economic conditions the reprocessing-recycle option is the more costly one.<ref>{{cite web|url=http://www.wise-uranium.org/nfccr.html|title=Recycled Nuclear Fuel Cost Calculator|website=www.wise-uranium.org|access-date=4 April 2018|archive-date=16 April 2013|archive-url=https://archive.today/20130416055108/http://www.wise-uranium.org/nfccr.html|url-status=live}}</ref> While the [[Chemistry:Uranium market|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 the fuel supply" via MOX fuel, breeder reactors or even the [[Physics:Thorium fuel cycle|thorium fuel cycle]] could become more attractive.  However, if uranium prices remain low, reprocessing will remain less attractive.{{citation needed|date=May 2022}}

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&nbsp;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= [[Organization:World Nuclear Association|World Nuclear Association]]
 | access-date= 3 May 2008
 | archive-date= 27 February 2013
 | archive-url= https://web.archive.org/web/20130227083958/http://world-nuclear.org/how/wastemanag.html
 | url-status= live
 }}</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=June 2006 |access-date=3 May 2008 |archive-date=2 March 2013 |archive-url=https://web.archive.org/web/20130302080054/http://www.world-nuclear.org/info/inf103.html |url-status=live }}</ref> However the level of transuranic elements, including [[Physics:Plutonium-239|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.

On 25 October 2011 a commission of the Japanese Atomic Energy Commission revealed during a meeting calculations about the costs of recycling nuclear fuel for power generation. These costs could be twice the costs of direct geological disposal of spent fuel: the cost of extracting plutonium and handling spent fuel was estimated at 1.98 to 2.14 yen per kilowatt-hour of electricity generated. Discarding the spent fuel as waste would cost only 1 to 1.35 yen per kilowatt-hour.<ref>NHK-world (26 October 2011) [http://www3.nhk.or.jp/daily/english/26_12.html Nuclear fuel recycling costs] {{webarchive |url=https://web.archive.org/web/20110810181931/http://www3.nhk.or.jp/daily/english/26_12.html |date=10 August 2011 }}</ref><ref>JAIF (26 October 2011) [https://web.archive.org/web/20130510210349/http://www.jaif.or.jp/english/news_images/pdf/ENGNEWS01_1319605053P.pdf Nuclear fuel recycling costs]</ref>

In July 2004 Japanese newspapers reported that the Japanese Government had estimated the costs of disposing radioactive waste, contradicting claims four months earlier that no such estimates had been made. The cost of non-reprocessing options was estimated to be between a quarter and a third ($5.5–7.9&nbsp;billion) of the cost of reprocessing ($24.7&nbsp;billion). At the end of the year 2011 it became clear that Masaya Yasui, who had been director of the Nuclear Power Policy Planning Division in 2004, had instructed his subordinate in April 2004 to conceal the data. The fact that the data were deliberately concealed obliged the ministry to re-investigate the case and to reconsider whether to punish the officials involved.<ref>{{cite news|title=Cover-up of estimated costs to dispose of radioactive waste raises serious questions|url=http://www.fissilematerials.org/blog/2012/01/japanese_mislead_about_sp.html|access-date=8 January 2012|newspaper=The Mainichi Daily News|date=2 January 2012|archive-date=27 February 2021|archive-url=https://web.archive.org/web/20210227121506/http://fissilematerials.org/blog/2012/01/japanese_mislead_about_sp.html|url-status=live}}</ref><ref>{{cite web|last=Mycle|first=Schneider|title=Japanese mislead about spent fuel reprocessing costs|date=2 January 2012|url=http://www.fissilematerials.org/blog/2012/01/japanese_mislead_about_sp.html|publisher=International Panel on Fissile Materials|access-date=8 January 2012|archive-date=27 February 2021|archive-url=https://web.archive.org/web/20210227121506/http://fissilematerials.org/blog/2012/01/japanese_mislead_about_sp.html|url-status=live}}</ref>

==List of sites==
{| class="wikitable sortable"
|-
! Country
! Reprocessing site
! Fuel type
! Procedure
! Reprocessing<br />capacity tHM/yr
! Commissioning<br />or operating period
|-
| {{flag|Belgium}} || Mol || LWR, MTR (Material test reactor) || || 80<ref name=EuroNuclear>{{cite web
 |url=http://www.euronuclear.org/info/encyclopedia/r/reprocessing-plants-ww.htm
 |title=Reprocessing plants, world-wide
 |publisher=[[Organization:European Nuclear Society|European Nuclear Society]]
 |access-date=29 July 2008
|archive-url=https://web.archive.org/web/20150622040852/http://www.euronuclear.org/info/encyclopedia/r/reprocessing-plants-ww.htm
 |archive-date=22 June 2015
|url-status=dead
 }}</ref> || style="background:#e21;"| 1966–1974<ref name=EuroNuclear/>
|-
|-
| {{flag|China}} || Plant 404<ref>[http://allthingsnuclear.org/post/2702753168/china-and-reprocessing-separating-fact-from-fiction All Things Nuclear • China and Reprocessing: Separating Fact from Fiction] {{Webarchive|url=https://web.archive.org/web/20110318072026/http://allthingsnuclear.org/post/2702753168/china-and-reprocessing-separating-fact-from-fiction |date=18 March 2011 }}. Allthingsnuclear.org (11 January 2011). Retrieved 10 December 2011.</ref> || || || 50 || style="background:#3a3;"| 2004
|-
| {{flag|Germany}} || Karlsruhe, WAK || LWR<ref name=Princeton>{{cite web
 |url=http://www.princeton.edu/~ota/disk1/1995/9530/953007.PDF
 |title=Civil Reprocessing Facilities
 |publisher=[[Organization:Princeton University|Princeton University]]
 |access-date=30 July 2008
 |archive-date=2 August 2020
 |archive-url=https://web.archive.org/web/20200802104946/http://www.princeton.edu/~ota/disk1/1995/9530/953007.PDF
 |url-status=live
 }}</ref>|| || 35<ref name=EuroNuclear/> || style="background:#e21;"| 1971–1990<ref name=EuroNuclear/>
|-
| {{flag|France}} || Marcoule, UP 1 || Military || || 1200<ref name=EuroNuclear/> || style="background:#e21;"| 1958<ref name=EuroNuclear/>-1997<ref>{{cite web
 |url=http://www.globalsecurity.org/wmd/world/france/marcoule.htm
 |title=Marcoule – Valrho
 |publisher=Global Security
 |access-date=30 July 2008
 |archive-date=23 September 2020
 |archive-url=https://web.archive.org/web/20200923182313/https://www.globalsecurity.org/wmd/world/france/marcoule.htm
 |url-status=live
 }}</ref>
|-
| {{flag|France}} || Marcoule, CEA APM || FBR || PUREX DIAMEX SANEX<ref>{{cite journal|url=http://www.jstage.jst.go.jp/article/jnst/44/3/44_410/_article |doi=10.3327/jnst.44.410 |title=Status of the French Research Program on Partitioning and Transmutation |author=Dominique Warin |year=2007 |journal=Journal of Nuclear Science and Technology |volume=44 |page=410 |issue=3 |doi-access=free }} </ref> || 6<ref name=Princeton/> || style="background:#3a3;"| 1988–present<ref name=Princeton/>
|-
| {{flag|France}} || La Hague, UP 2 || LWR<ref name=Princeton/>|| PUREX<ref name=FranceNucleaire>
{{cite web
 |url=http://www.francenuc.org/en_sites/lnorm_hague_e.htm
 |title=BASSE-NORMANDIE- LOWER NORMANDY, La Hague
 |publisher=France Nucleaire
 |access-date=31 July 2008
|url-status=dead
 |archive-url=https://web.archive.org/web/20110716141530/http://www.francenuc.org/en_sites/lnorm_hague_e.htm
 |archive-date=16 July 2011 }}</ref> || 900<ref name=EuroNuclear/> || style="background:#e21;"|1967–1974<ref name=EuroNuclear/>
|-
| {{flag|France}} || La Hague, UP 2–400 || LWR<ref name=Princeton/> || PUREX<ref name=FranceNucleaire/> || 400<ref name=EuroNuclear/> || style="background:#e21;"|1976–1990<ref name=EuroNuclear/>
|-
| {{flag|France}} || La Hague, UP 2–800 || LWR || PUREX<ref name=FranceNucleaire/> || 800<ref name=EuroNuclear/> || style="background:#3a3;"| 1990<ref name=EuroNuclear/>
|-
| {{flag|France}} || La Hague, UP 3 || LWR || PUREX<ref name=FranceNucleaire/> || 800<ref name=EuroNuclear/> || style="background:#3a3;"| 1990<ref name=EuroNuclear/>
|-
| {{flag|UK}} || Windscale, B204|| [[Engineering:Magnox|Magnox]], LWR || BUTEX || 750<ref name=EuroNuclear/> || style="background:#e21;" | 1956–1962,<ref name=EuroNuclear/> 1969-1973
|-
| {{flag|UK}} || Sellafield, Magnox Reprocessing Plant || Magnox,<ref name=Princeton/> LWR, FBR || PUREX || 1500<ref name=EuroNuclear/> || style="background:#e21;"| 1964<ref name=EuroNuclear/>-2022
|-
| {{flag|UK}} || [[Engineering:Dounreay|Dounreay]] || FBR<ref name=Princeton/> || || 8<ref name=EuroNuclear/> || style="background:#e21;"| 1980<ref name=EuroNuclear/>
|-
| {{flag|UK}} || THORP || AGR, LWR || PUREX || 900<ref name=EuroNuclear/><ref name=wnalist13/> || style="background:#e21;"| 1994<ref name=EuroNuclear/><ref name=wnalist13>{{cite web|url=http://www.world-nuclear.org/info/Nuclear-Fuel-Cycle/Fuel-Recycling/Processing-of-Used-Nuclear-Fuel/|title=Processing of Used Nuclear Fuel|date=September 2013|publisher=World Nuclear Association|access-date=5 December 2013|archive-date=23 January 2016|archive-url=https://web.archive.org/web/20160123112054/http://www.world-nuclear.org/info/Nuclear-Fuel-Cycle/Fuel-Recycling/Processing-of-Used-Nuclear-Fuel/|url-status=live}}</ref>-2018
|-
| {{flag|Italy}} || Rotondella || [[Physics:Thorium fuel cycle|Thorium]] || || 5<ref name=EuroNuclear/> || style="background:#e21;"| 1968<ref name=EuroNuclear/> (shutdown)
|-
| {{flag|India}} || [[Organization:Bhabha Atomic Research Centre|Trombay]] || Military || PUREX<ref>{{cite web
 |url=http://www.globalsecurity.org/wmd/world/india/kalpakkam.htm
 |title=CIRUS and DHRUVA Reactors, Trombay
 |publisher=Global Security
 |access-date=30 July 2008
 |archive-date=26 January 2021
 |archive-url=https://web.archive.org/web/20210126060010/https://www.globalsecurity.org/wmd/world/india/kalpakkam.htm
 |url-status=live
 }}</ref> || 60<ref name=EuroNuclear/> || style="background:#3a3;"| 1965<ref name=EuroNuclear/>
|-
| {{flag|India}} || Tarapur || PHWR || PUREX || 100<ref name=EuroNuclear/> || style="background:#3a3;"| 1982<ref name=EuroNuclear/>
|-
| {{flag|India}} || Kalpakkam || PHWR and FBTR || PUREX || 100<ref name=GlobalSecurityKARP>{{cite web
 |url=http://www.globalsecurity.org/wmd/world/india/kalpakkam.htm
 |title=Kalpakkam Atomic Reprocessing Plant [KARP]
 |publisher=Global Security
 |access-date=30 July 2008
 |archive-date=26 January 2021
 |archive-url=https://web.archive.org/web/20210126060010/https://www.globalsecurity.org/wmd/world/india/kalpakkam.htm
 |url-status=live
 }}</ref> || style="background:#3a3;"|1998<ref name=GlobalSecurityKARP/>
|-
| {{flag|India}} || Tarapur || PHWR || || 100<ref name="BusStd">[http://www.business-standard.com/india/news/pm-to-dedicate-tarapur-nuke-reprocessing-unit-next-week/420362/ PM to dedicate Tarapur nuke reprocessing unit next week] {{Webarchive|url=https://web.archive.org/web/20121009221549/http://www.business-standard.com/india/news/pm-to-dedicate-tarapur-nuke-reprocessing-unit-next-week/420362/ |date=9 October 2012 }}. Business-standard.com. Retrieved 10 December 2011.</ref> || style="background:#3a3;"| 2011<ref name="BusStd"/>
|-
| {{flag|Israel}} || [[Organization:Shimon Peres Negev Nuclear Research Center|Dimona]] || Military || || 60–100<ref name=GFMR2010/> || style="background:#3a3;"| ~1960–present
|-
| {{flag|Japan}} || Tokai || LWR<ref>{{cite web
 |url=http://www.globalsecurity.org/wmd/world/japan/tokai.htm
 |title=Tokai Reprocessing Plant (TRP)
 |publisher=Global Security
 |access-date=30 July 2008
 |archive-date=23 September 2020
 |archive-url=https://web.archive.org/web/20200923191606/https://www.globalsecurity.org/wmd/world/japan/tokai.htm
 |url-status=live
 }}</ref> || || 210<ref name=EuroNuclear/> || style="background:#e21;"| 1977<ref name=EuroNuclear/>-2006<ref name=kramer2012 >{{Cite journal | last1 = Kramer | first1 = D. | title = Is Japan ready to forgo nuclear reprocessing? | doi = 10.1063/PT.3.1469 | journal = Physics Today | volume = 65 | issue = 3 | pages = 25–42 | year = 2012 |bibcode = 2012PhT....65c..25K }}</ref>
|-
| {{flag|Japan}} || Rokkasho || LWR<ref name=Princeton/> || || 800<ref name=EuroNuclear/><ref name=wnalist13/> || style="background:#e21;"| under construction (2024)<ref name=wnn-20171228>{{cite news |url=http://www.world-nuclear-news.org/WR-Further-delay-to-completion-of-Rokkasho-facilities-2812174.html |title=Further delay to completion of Rokkasho facilities |publisher=World Nuclear News |date=28 December 2017 |access-date=28 December 2017 |archive-date=29 December 2017 |archive-url=https://web.archive.org/web/20171229112855/http://www.world-nuclear-news.org/WR-Further-delay-to-completion-of-Rokkasho-facilities-2812174.html |url-status=live }}</ref>
|-
| {{flag|Pakistan}} || [[Physics:Pakistan Atomic Research Reactor#PARR-III Reactor (New-Labs)|New Labs]], [[Place:Rawalpindi|Rawalpindi]] || Military/[[Physics:Nuclear fuel cycle#Plutonium cycle|Plutonium]]/[[Physics:Thorium fuel cycle|Thorium]] || || 80<ref>{{cite web |url=https://fas.org/nuke/guide/pakistan/facility/rawalpindi.htm |title=Rawalpindi / Nilhore |publisher=[[Organization:Federation of American Scientists|Federation of American Scientists]] |access-date= |archive-date=4 March 2016 |archive-url=https://web.archive.org/web/20160304061905/https://fas.org/nuke/guide/pakistan/facility/rawalpindi.htm |url-status=live }}</ref>|| style="background:#3a3;"| 1982–present
|-
|{{flag|Pakistan}} || Khushab Nuclear Complex, Atomic City of Pakistan||HWR/Military/[[Physics:Tritium|Tritium]] || || 22&nbsp;kg<ref>{{cite news |title=Pakistan's Indigenous Nuclear Reactor Starts Up| work=The Nation|date= 13 April 1998 }}</ref>|| style="background:#3a3;"| 1986–present
|-
| {{flag|Russia}} || Mayak Plant B || Military || || 400 || style="background:#e21;"| 1948-196?<ref>{{cite web
 |url=http://www.globalsecurity.org/wmd/world/russia/chelyabinsk-65_nuc.htm
 |title=Chelyabinsk-65
 |publisher=Global Security
 |access-date=29 July 2008
 |archive-date=3 September 2010
 |archive-url=https://web.archive.org/web/20100903191113/http://www.globalsecurity.org/wmd/world/russia/chelyabinsk-65_nuc.htm
 |url-status=live
 }}</ref>
|-
| {{flag|Russia}} || Mayak Plant BB, RT-1 || LWR<ref name=Princeton/> || PUREX + Np separation<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=Modernization and Enhancement of NMAC at the Mayak RT-1 Plant
 |publisher=INMM
 |author=S. Guardini|date=16 June 2003
|access-date=9 August 2008|display-authors=0}}</ref> || 400<ref name=EuroNuclear/><ref name=wnalist13/> || style="background:#3a3;"| 1978<ref name=EuroNuclear/>
|-
| {{flag|Russia}} || Tomsk-7 Radiochemical Plant || Military || || 6000<ref name=GFMR2010>{{cite web|title=Global Fissile Material Report 2010|publisher=International Panel on Fissile Materials|url=http://fissilematerials.org/library/gfmr10.pdf|archive-url=https://web.archive.org/web/20200424023733/http://fissilematerials.org/library/gfmr10.pdf |archive-date=2020-04-24}}</ref> || style="background:#3a3;"| 1956<ref>{{cite web
 |url=https://www.globalsecurity.org/wmd/world/russia/tomsk-7_nuc.htm
 |title=Tomsk-7 / Seversk
 |publisher=Global Security
 |access-date=1 June 2020
 |archive-date=29 July 2020
 |archive-url=https://web.archive.org/web/20200729033421/https://www.globalsecurity.org/wmd/world/russia/tomsk-7_nuc.htm
 |url-status=live
 }}</ref>
|-
| {{flag|Russia}} || Zheleznogorsk (Krasnoyarsk-26) || Military || || 3500<ref name=GFMR2010/> || style="background:#e21;"| 1964–~2010<ref>{{cite web
 |url=https://www.globalsecurity.org/wmd/world/russia/krasnoyarsk-26_nuc.htm
 |title=Krasnoyarsk-26 / Zheleznogorsk
 |publisher=Global Security
 |access-date=1 June 2020
 |archive-date=31 July 2020
 |archive-url=https://web.archive.org/web/20200731200132/https://www.globalsecurity.org/wmd/world/russia/krasnoyarsk-26_nuc.htm
 |url-status=live
 }}</ref>
|-
| {{flag|Russia}} || Zheleznogorsk, RT-2 || [[Engineering:VVER|VVER]] || || 800<ref name=EuroNuclear/> || style="background:#e21;"| under construction (2030)
|-
| {{flag|USA}}, {{wipe|Washington (state)}} || [[Astronomy:Hanford Site|Hanford Site]] || Military || bismuth phosphate, REDOX, PUREX || || style="background:#e21;"| 1944–1988<ref>{{cite web | url=http://www.hanford.gov/page.cfm/TPlant | title=T Plant overview | publisher=Department of energy | access-date=9 April 2011 | archive-date=18 March 2021 | archive-url=https://web.archive.org/web/20210318075818/https://www.hanford.gov/page.cfm/TPlant | url-status=live }}</ref>
|-
| {{flag|USA}}, SC || Savannah River Site || Military/LWR/HWR/Tritium ||PUREX, REDOX, THOREX, Np separation || 5000<ref name=Fernandez>{{cite web | url=http://www.c-n-t-a.com/srs50_files/133fernandez.pdf | author=LeVerne Fernandez | title=Savannah River Site Canyons—Nimble Behemoths of the Atomic Age (WSRC-MS-2000-00061) | access-date=9 April 2011 | archive-date=3 March 2016 | archive-url=https://web.archive.org/web/20160303190956/http://www.c-n-t-a.com/srs50_files/133fernandez.pdf | url-status=live }}</ref> || style="background:#e21;"| 1952–2002
|-
| {{flag|USA}}, NY || West Valley || LWR<ref name=Princeton/> || PUREX || 300<ref name=EuroNuclear/> || style="background:#e21;"| 1966–1972<ref name=EuroNuclear/>
|-
|{{flag|USA}}, NC
|Barnwell
|LWR
|PUREX
|1500
|never permitted to operate<ref>{{Citation|title=West Valley Demonstration Project|date=1 December 2018|url=https://en.wikipedia.org/w/index.php?title=West_Valley_Demonstration_Project&oldid=871534289|work=Wikipedia|language=en|access-date=13 April 2020|archive-date=25 January 2022|archive-url=https://web.archive.org/web/20220125095652/https://en.wikipedia.org/w/index.php?title=West_Valley_Demonstration_Project&oldid=871534289|url-status=live}}</ref>
|-
|{{flag|USA}}, ID
|[[Organization:Idaho National Laboratory|INL]]
|LWR
|PUREX
| –
|
|}

==See also==

* [[Physics:Nuclear fuel cycle|Nuclear fuel cycle]]
* [[Engineering:Breeder reactor|Breeder reactor]]
* Nuclear fusion-fission hybrid
* Spent nuclear fuel shipping cask
* Taylor Wilson's nuclear waste-fired small reactor
* Global Nuclear Energy Partnership announced February 2006

==References==
{{reflist|colwidth=30em}}

==Notes==
{{Notelist}}

==Further reading==
*{{cite journal
 | last1 = Williamson | first1 = M.A.
 | last2 = Willit | first2 = J.L.
 | title = Pyroprocessing Flowsheets for Recycling Used Nuclear Fuel
 | journal = Nuclear Engineering and Technology
 | volume = 43
 | issue = 4
 | pages = 329–334
 | year = 2011
 | url = http://www.ne.anl.gov/pdfs/ANL_Pyroprocessing_Flowsheets_Final.pdf
 | doi = 10.5516/NET.2011.43.4.329 | doi-access = free
 }}
* {{cite journal
 | last1 = Till | first1 = C.E.
 | last2 = Chang | first2 = Y.I
 | last3 = Hannum | first3 = W.H.
 | title = The integral fast reactor-an overview
 | journal = Progress in Nuclear Energy
 | volume = 31
 | issue = 1–2
 | pages = 3–11
 | year = 1997
 | doi = 10.1016/0149-1970(96)00001-7 }}
*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==
* [https://web.archive.org/web/20050404160048/http://world-nuclear.org/info/inf69.htm Processing of Used Nuclear Fuel], World Nuclear Association
*[https://web.archive.org/web/20160521105558/http://www.euronuclear.org/info/encyclopedia/p/purex-process.htm PUREX Process, European Nuclear Society]
*[https://web.archive.org/web/20060117004320/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
*[https://web.archive.org/web/20060323012601/http://mfnl.xjtu.edu.cn/gov-doe-ornl/RDF/history/03_FuelReprocessinHistory.html Brief History of Fuel Reprocessing]
*[https://web.archive.org/web/20100301163816/http://alsos.wlu.edu/qsearch.aspx?browse=science%2FReprocessing+Spent+Fuel Annotated bibliography for reprocessing spent nuclear fuel from the Alsos Digital Library for Nuclear Issues]

{{Nuclear technology}}

{{DEFAULTSORT:Nuclear Reprocessing}}
[[Category:Nuclear reprocessing| ]]

{{Sourceattribution|Nuclear reprocessing}}