Editing Nuclear reprocessing (section)

==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 [[MOX fuel#Americium content|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>]]
[[Pyroprocessing]] is a generic term for high-temperature methods. Solvents are [[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]], and solvent-solvent extraction are common steps.

These processes are not currently in significant use worldwide, but they have been pioneered at [[Argonne National Laboratory]]<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=[[Argonne National Laboratory]]
 |access-date=6 June 2016
}}</ref> with current research also being developed in Russia,<ref>{{Cite journal |last=Gutorova |first=S. V. |last2=Logunov |first2=M. V. |last3=Voroshilov |first3=Yu. A. |last4=Babain |first4=V. A. |last5=Shadrin |first5=A. Yu. |last6=Podoynitsyn |first6=S. V. |last7=Kharitonov |first7=O. V. |last8=Firsova |first8=L. A. |last9=Kozlitin |first9=E. A. |last10=Ustynyuk |first10=Yu. A. |last11=Lemport |first11=P. S. |last12=Nenajdenko |first12=V. G. |last13=Voronina |first13=A. V. |last14=Volkovich |first14=V. A. |last15=Polovov |first15=I. B. |date=2024-12-01 |title=Modern Trends in Spent Nuclear Fuel Reprocessing and Waste Fractionation |url=https://link.springer.com/article/10.1134/S1070363224150015 |journal=Russian Journal of General Chemistry |language=en |volume=94 |issue=2 |pages=S243–S430 |doi=10.1134/S1070363224150015 |issn=1608-3350}}</ref> as well, taking place at [[Central Research Institute of Electric Power Industry|CRIEPI]] in Japan, the Nuclear Research Institute of [[Řež]] in Czech Republic, [[Indira Gandhi Centre for Atomic Research]] in India and [[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-[[burnup]] spent fuel and requires little cooling time, since the [[operating temperature]]s are high already.
* Does not use solvents containing hydrogen and carbon, which are [[neutron moderator]]s creating risk of [[criticality accident]]s and can absorb the [[fission product]] [[tritium]] and the [[activation product]] [[carbon-14]] in dilute solutions that cannot be separated later.
**Alternatively, voloxidation<ref name=advancedheadend /> (see [[#Voloxidation|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 product]]s on site as [[high-level waste]] until [[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 [[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 [[minor actinide]]s ([[americium]] and [[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 [[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.
* [[Fluoride volatility]] (see [[#Fluoride volatility|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 [[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 [[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 [[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 [[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 [[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 [[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 [[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 potential]]s 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 [[Argonne National Laboratory]] and used in the [[Integral Fast Reactor]] project.

'''PYRO-A''' is a means of separating actinides (elements within the [[actinide]] family, generally heavier than U-235) from non-actinides. The spent fuel is placed in an [[anode]] [[basket]] which is immersed in a molten salt electrolyte. An electric current is applied, causing the uranium metal (or sometimes oxide, depending on the spent fuel) to plate out on a solid metal cathode while the other actinides (and the rare earths) can be absorbed into a liquid [[cadmium]] cathode. Many of the fission products (such as [[caesium]], [[zirconium]] and [[strontium]]) remain in the salt.<ref>{{cite web|url=http://criepi.denken.or.jp/en/e_publication/pdf/den363.pdf |publisher=CRIEPI News |title=Development of pyro-process fuel cell technology |date=July 2002 |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 [[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]] [[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= [[United States Department of Energy|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 material|fertile]], though many of these materials would require a [[fast breeder reactor]] to be burned efficiently. In a [[thermal neutron]] spectrum, the concentrations of several heavy actinides ([[curium]]-242 and [[plutonium-240]]) can become quite high, creating fuel that 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 [[transuranic]]s in an inert matrix such as metallic [[zirconium]]. In the PYRO-B processing of such fuel, an [[electrorefining]] step is used to separate the residual transuranic elements from the fission products and recycle the transuranics to the reactor for fissioning. Newly generated technetium and iodine are extracted for incorporation into transmutation targets, and the other fission products are sent to waste.

===Voloxidation===
Voloxidation (for ''volumetric oxidation'' or ''volatile oxidation''<ref>{{cite journal|doi=10.1021/acsomega.4c00029 |title=Elucidating the Composition and Structure of Uranium Oxide Powders Produced via NO<sub>2</sub> Voloxidation |date=2024 |last1=Peruski |first1=Kathryn M. |last2=Spano |first2=Tyler L. |last3=Vick |first3=Matthew C. |last4=Cobble |first4=Chase |last5=Greaney |first5=Allison T. |last6=McFarlane |first6=Joanna |journal=ACS Omega |volume=9 |issue=9 |pages=10979–10991 |pmid=38463331 |pmc=10918823 }}</ref>) involves heating oxide fuel with oxygen, sometimes with alternating oxidation and reduction, or alternating oxidation by [[ozone]] to [[uranium trioxide]] with decomposition by heating back to [[triuranium octoxide]].<ref name="advancedheadend"/> A major purpose is to capture [[tritium]] as tritiated water vapor before further processing where it would be difficult to retain the tritium. Tritium is a difficult contaminant to remove from aqueous solution, as it cannot be separated from water except by isotope separation. However, tritium is also a valuable product used in industry science and [[nuclear weapons]], so recovery of a stream of hydrogen or water with a high tritium content can make targeted recovery economically worthwhile. Other volatile elements leave the fuel and must be recovered, especially [[iodine]], [[technetium]], and [[carbon-14]]. Voloxidation also breaks up 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 ([[hot cell]]s, [[remote handling]] equipment)
* Products such as [[krypton-85]] or tritium, as well as [[xenon]] (whose isotope are either stable, [[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 [[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 [[isotopes of helium]]) can in principle be recovered in a [[cold trap]] cooled by [[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 [[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|583.8|K}} which reduces the issues presented by technetium contamination in processes like fluoride volatility or PUREX; [[ruthenium tetroxide]] (gaseous above {{convert|129.6|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 [[oxidizing 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 [[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 [[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=etal |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 !![[Zeolite]]<br />filter!!Carbon<br />filter!!Particle<br />filters
|-
|[[Palladium]]||28||14||14|| ||
|-
|[[Tellurium]]||10||5||5|| ||
|-
|[[Molybdenum]]||70|| ||70|| ||
|-
|[[Caesium]]||46|| ||46|| ||
|-
|[[Rubidium]]||8|| ||8|| ||
|-
|Silver||2|| ||2|| ||
|-
|[[Iodine]]||4|| || ||4||
|-
|Cladding||2000||2000|| || ||
|-
|[[Uranium]]||19218||19218|| || ||?
|-
|Others||614||614|| || ||?
|-
|Total||22000||21851||145||4||0
|}

====Advantages====
* Requires no chemical processes at all
* Can in theory be done "self heating" via the [[decay heat]] of sufficiently "fresh" spent fuel
* [[Caesium-137]] has uses in [[food irradiation]] and can be used to power [[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 [[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, [[Strontium oxide]] remains solid below {{convert|2804|K}} and if strontium oxide is to be recovered with other liquid effluents, it has to be [[reduction (chemistry)|reduced]] to the native metal before the heating step. Both Strontium and Strontium oxide form soluble [[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 [[actinide]]s, including [[neptunium]] (melting point: {{convert|912|K}}) and [[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 [[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 [[thermal stress]]. This combines with radiation damage to material and possibly [[neutron embrittlement]] if [[neutron source]]s such as [[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 level waste|low-]] to intermediate level waste.

=== Fluoride volatility ===
{{Main|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 [[Nuclear fission|fission]], not considering later [[neutron capture]], fraction of 100% not 200%. [[Beta decay]] [[Kr-85]]→[[Rubidium|Rb]], [[Sr-90]]→[[Zirconium|Zr]], [[Ru-106]]→[[Palladium|Pd]], [[Sb-125]]→[[Tellurium|Te]], [[Cs-137]]→[[Barium|Ba]], [[Ce-144]]→[[Neodymium|Nd]], [[Sm-151]]→[[Europium|Eu]], [[Eu-155]]→[[Gadolinium|Gd]] visible.]]
In the fluoride volatility process, [[fluorine]] is reacted with the fuel. Fluorine is so much more reactive than even [[oxygen]] that small particles of ground oxide fuel will burst into flame when dropped into a chamber full of fluorine. This is known as flame fluorination; the heat produced helps the reaction proceed. Most of the [[uranium]], which makes up the bulk of the fuel, is converted to [[uranium hexafluoride]], the form of uranium used in [[uranium enrichment]], which has a very low boiling point. [[Technetium]], the main [[long-lived fission product]], is also efficiently converted to its volatile hexafluoride. A few other elements also form similarly volatile hexafluorides, pentafluorides, or heptafluorides. The volatile fluorides can be separated from excess fluorine by condensation, then separated from each other by [[fractional distillation]] or selective [[redox|reduction]]. [[Uranium hexafluoride]] and [[technetium hexafluoride]] have very similar boiling points and vapor pressures, which makes complete separation more difficult.

Many of the [[fission product]]s volatilized are the same ones volatilized in non-fluorinated, higher-temperature volatilization, such as [[iodine]], [[tellurium]] and [[molybdenum]]; notable differences are that [[technetium]] is volatilized, but [[caesium]] is not.

Some transuranium elements such as [[plutonium]], [[neptunium]] and [[americium]] can form volatile fluorides, but these compounds are not stable when the fluorine partial pressure is decreased.<ref>{{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, [[neptunium]], and plutonium as volatile fluorides; however, some other [[minor actinides]] may not form volatile fluorides and instead remain with the alkaline fission products. Some [[noble metals]] may not form fluorides at all, but remain in metallic form; however [[ruthenium hexafluoride]] is relatively stable and volatile.

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

[[Molten salt reactor]] designs carry out fluoride volatility reprocessing continuously or at frequent intervals. The goal is to return [[actinide]]s to the molten fuel mixture for eventual fission, while removing [[fission product]]s that are [[neutron poison]]s, or that can be more securely stored outside the reactor core while awaiting eventual transfer to permanent storage.

====Chloride volatility and solubility====
Many of the elements that form volatile high-[[valence (chemistry)|valence]] fluorides will also form volatile high-valence chlorides. Chlorination and distillation is another possible method for separation. The sequence of separation may differ usefully from the sequence for fluorides; for example, [[zirconium tetrachloride]] and [[tin tetrachloride]] have relatively low boiling points of {{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=[[Oak Ridge National Laboratory]], U.S. DOE |access-date=3 May 2008 |display-authors=etal |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 [[americium]], [[curium]], [[lanthanides]], [[strontium]], [[caesium]] are more soluble than those of [[uranium]], [[neptunium]], [[plutonium]], and [[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 [[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
* [[Uranium hexafluoride]] and [[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 thousands of tons of uranium would require smaller facilities than processing billions of tons of petroleum {{mdash}} 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.
* [[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 [[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 |bibcode=2023RaPC..20810919V |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 [[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| bibcode = 2010RFPol..70..484K | hdl = 2117/10365 }}</ref>