Editing Nuclear reprocessing (section)

=== 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