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Materials separated from spent fuel
Actinides: Lighter elements:
Reprocessed uranium Fission products
Plutonium Activation products
Minor actinides Cladding
Nuclear reprocessing separates components of spent nuclear fuel.Reprocessing serves multiple purposes, whose relative importance has changed over time:
material disposition
plutonium, minor actinides, reprocessed uranium fission in fast, fusion, or subcritical reactor
reprocessed uranium, cladding, filters less stringent storage as low-level waste
long-lived fission and activation products nuclear transmutation or geological repository
medium-lived fission products 137Cs and 90Sr medium-term storage as high-level waste
useful radionuclides and noble metals industrial and medical uses


The first large-scale nuclear reactors were built during World War II. These reactors were designed for the production of plutonium for use in nuclear weapons. The only reprocessing required, therefore, was the extraction of the plutonium (free of fission-product contamination) from the spent natural uranium fuel. In 1943, several methods were proposed for separating the relatively small quantity of plutonium from the uranium and fission products. The first method selected, a precipitation process called the Bismuth Phosphate process, was developed and tested at the Oak Ridge National Laboratorymarker (ORNL) in the 1943-1945 period to produce quantities of plutonium for evaluation and use in weapons programs. ORNL produced the first macroscopic quantities (grams) of separated plutonium with these processes.

The Bismuth Phosphate process was first operated on a large scale at the Hanford Sitemarker, in the latter part of 1944. It was successful for plutonium separation in the emergency situation existing then, but it had a significant weakness: the inability to recover uranium.

The first successful solvent extraction process for the recovery of pure uranium and plutonium was developed at ORNL in 1949. The PUREX process is the current method of extraction. Separation plants were also constructed at Savannah River Sitemarker and a smaller plant at West Valley, New Yorkmarker which closed by 1972 because of its inability to meet new regulatory requirements.

Reprocessing of civilian fuel has long been employed in Europe, at the COGEMA La Hague sitemarker in France, the Sellafieldmarker 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, fear of nuclear weapons proliferation (especially after Indiamarker 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 April 7, 1977 , President Jimmy Carter banned the reprocessing of commercial reactor spent nuclear fuel. The key issue driving this policy was the serious threat of nuclear weapons proliferation by diversion of plutonium from the civilian fuel cycle, and to encourage other nations to follow the USA lead.. After that, only countries that already had large investments in reprocessing infrastructure continued to reprocess spent nuclear fuel. President Reagan lifted the ban in 1981, but did not provide the substantial subsidy that would have been necessary to start up commercial reprocessing.

In March 1999, the U.S. Department of Energy (DOE) reversed its own policy and signed a contract with a consortium of Duke Energy, COGEMA, and Stone & Webster (DCS) to design and operate a Mixed Oxide fuel fabrication facility. Site preparation at the Savannah River Site (South Carolina) began in October 2005.

The Global Nuclear Energy Partnership, announced by the secretary of the Department of Energy, Samuel Bodman, on February 6, 2006, is a plan to form an international partnership to reprocess spent nuclear fuel in a way that renders the plutonium in it usable for nuclear fuel but not for nuclear weapons. The Department of Energy reversed these plans in July of 2009, under the Obama administration.

Water and organic solvents


PUREX, the current standard method, is an acronym standing for Plutonium and Uranium Recovery by EXtraction. The PUREX process is a liquid-liquid extraction method used to reprocess spent nuclear fuel, in order to extract uranium and plutonium, independent of each other, from the fission products. This is the most developed and widely used process in the industry at present.When used on fuel from commercial power reactors the plutonium extracted typically contains too much Pu-240 to be useful in a nuclear weapon. However, reactors that are capable of refuelling frequently can be used to produce weapon-grade plutonium, which can later be recovered using PUREX. Because of this, PUREX chemicals are monitored.

Modifications of PUREX


The PUREX process can be modified to make a UREX (URanium EXtraction) process which could be used to save space inside high level nuclear waste disposal sites, such as the Yucca Mountain nuclear waste repositorymarker, by removing the uranium which makes up the vast majority of the mass and volume of used fuel and recycling it as reprocessed uranium.

The UREX process is a PUREX process which has been modified to prevent the plutonium from being extracted. This can be done by adding a plutonium reductant before the first metal extraction step. In the UREX process, ~99.9% of the Uranium and >95% of Technetium are separated from each other and the other fission products and actinides. The key is the addition of acetohydroxamic acid (AHA) to the extraction and scrub sections of the process. The addition of AHA greatly diminishes the extractability of Plutonium and Neptunium, providing greater proliferation resistance than with the plutonium extraction stage of the PUREX process.


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 (TRansUranic EXtraction) process. TRUEX was invented in the USA by Argonne National Laboratory and is designed to remove the transuranic metals (Am/Cm) from waste. The idea is that by lowering the alpha activity of the waste, the majority of the waste can then be disposed of with greater ease. In common with PUREX this process operates by a solvation mechanism.


As an alternative to TRUEX, an extraction process using a malondiamide has been devised. The DIAMEX (DIAMideEXtraction) process has the advantage of avoiding the formation of organic waste which contains elements other than Carbon, Hydrogen, Nitrogen, and Oxygen. Such an organic waste can be burned without the formation of acidic gases which could contribute to acid rain. The DIAMEX process is being worked on in Europe by the French CEA. The process is sufficiently mature that an industrial plant could be constructed with the existing knowledge of the process. In common with PUREX this process operates by a solvation mechanism.


Selective ActiNide EXtraction. As part of the management of minor actinides it has been proposed that the lanthanides and trivalent minor actinides should be removed from the PUREX raffinate by a process such as DIAMEX or TRUEX. In order to allow the actinides such as americium to be either reused in industrial sources or used as fuel the lanthanides must be removed. The lanthanides have large neutron cross sections and hence they would poison a neutron driven nuclear reaction. To date the extraction system for the SANEX process has not been defined, but currently several different research groups are working towards a process. For instance the French CEA is working on a bis-triazinyl pyridine (BTP) based process.Other systems such as the dithiophosphinic acids are being worked on by some other workers.


The UNiversal EX'traction process was developed in Russiamarker and the Czech Republicmarker; it is designed to completely remove the most troublesome radioisotopes (Sr, Cs and minor actinides) from the raffinate remaining after the extraction of uranium and plutonium from used nuclear fuel. The chemistry is based upon the interaction of caesium and strontium with poly ethylene oxide (poly ethylene glycol) and a cobalt carborane anion (known as chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and the diluent is a polar aromatic such as nitrobenzene. Other dilents such as meta-nitrobenzotrifluoride and phenyl trifluoromethyl sulfone have been suggested as well.

Electrochemical methods

An exotic method using electrochemistry and ion exchange in ammonium carbonate has been reported.

Obsolete methods

Bismuth phosphate

The bismuth phosphate process is a very old process which adds lots of material to the final highly active waste. It was replaced by solvent extraction processes. The process was designed to extract plutonium from aluminium-clad uranium metal fuel. The fuel was declad by boiling it in caustic soda. After decladding, the uranium metal was dissolved in nitric acid. The plutonium at this point is in the +4 oxidation state. It was then precipitated by the addition of bismuth nitrate and phosphoric acid to form the bismuth phosphate. The plutonium was coprecipitated with this. The supernatant liquid (containing many of the fission products) was separated from the solid. The precipitate was then dissolved in nitric acid before the addition of an oxidant such as potassium permanganate which converted the plutonium to PuO22+ (Pu VI), then a dichromate salt was added to maintain the plutonium in the +6 oxidation state. The bismuth phosphate was then re-precipitated leaving the plutonium in solution. Then an iron (II) salt such as ferrous sulfate was added and the plutonium re-precipitated again using a bismuth phosphate carrier precipitate. Then lanthanum salts and fluoride were added to create solid lanthanum fluoride which acted as a carrier for the Pu. This was converted to the oxide by the action of a base. The lanthanum plutonium oxide was then collected and extracted with nitric acid to form plutonium nitrate.

Hexone or Redox

This is a liquid-liquid extraction process which uses methyl isobutyl ketone as the extractant. The extraction is by a solvation mechanism. This process has the disadvantage of requiring the use of a salting out reagent (aluminium nitrate) to increase the nitrate concentration in the aqueous phase to obtain a reasonable distribution ratio (D value). Also hexone is degraded by concentrated nitric acid. This process has been replaced by PUREX.

Pu4+ + 4NO3- + 2S --> [Pu(NO3)4S2]

Butex, β,β'-dibutyoxydiethyl ether

A process based on a solvation extraction process using the triether extractant named above. This process has the disadvantage of requiring the use of a salting out reagent (aluminium nitrate) to increase the nitrate concentration in the aqueous phase to obtain a reasonable distribution ratio. This process was used at Windscalemarker many years ago. This process has been replaced by PUREX.


Pyroprocessing is a generic term for high-temperature methods. Solvents are molten salts (e.g. LiCl+KCl or LiF+CaF2) 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 researched and developed at Argonne National Laboratorymarker and elsewhere.

  • The principles behind them are well understood, and no significant technical barriers exist to their adoption.
  • Readily applied to high-burnup spent fuel and requires little cooling time, since the operating temperatures are high already.
  • Does not use solvents containing hydrogen and carbon, which are neutron moderators creating risk of criticality accidents and can absorb the fission product tritium and the activation product carbon-14 in dilute solutions that cannot be separated later.
    • Alternatively, #Voloxidation 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 decommissioning. For example, the Integral Fast Reactor and Molten Salt Reactor fuel cycles are based on on-site pyroprocessing.
  • Can separate many or even all actinides at once and produce highly radioactive fuel which is harder to manipulate for theft or making nuclear weapons. (However, the difficulty has been questioned.) In contrast the PUREX process was designed to separate plutonium for weapons, and also leaves alkaline minor actinides (americium, curium) behind, producing waste with more long-lived radioactivity.
  • Most radioactivity in roughly 102~105 years after fuel use is produced by actinides, since there are no fission products with half-lives in this range. These actinides can fuel fast reactors, so extracting and reusing (fissioning) them reduces the long-term radioactivity of the waste and reduces fuel needs.

  • Reprocessing as a whole is not currently (2005) in favor, and places that do reprocess already have PUREX plants constructed. Consequently, there is little demand for new pyrometalurgical systems, although there could be if the Generation IV reactor programs become reality.
  • The used salt from pyroprocessing is less suitable for conversion into glass than the raffinate from PUREX processing.


PYRO-A and -B for IFR

These processes were developed by Argonne National Laboratorymarker and used in the Integral Fast Reactor project.

PYRO-A is a means of separating actinides (elements within the actinide family, generally heavier than U-235) from non-actinides. The spent fuel is placed in an anode basket which is immersed in a molten salt electrolyte. An electrical current is applied, causing the uranium metal (or sometimes oxide, depending on the spent fuel) to plate out on a solid metal cathode while the other actinides (and the rare earths) can be absorbed into a liquid cadmium cathode. Many of the fission products (such as caesium, zirconium and strontium) remain in the salt. As alternatives to the molten cadmium electrode it is possible to use a molten bismuth cathode, or a solid aluminium cathode.

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.

Since the majority of the long term radioactivity, and volume, of spent fuel comes from actinides, removing the actinides produces waste that is more compact, and not nearly as dangerous over the long term. The radioactivity of this waste will then drop to the level of various naturally occurring minerals and ores within a few hundred, rather than thousands, years.

The mixed actinides produced by pyrometallic processing can be used again as nuclear fuel, as they are virtually all either fissile, or fertile, though many of these materials would require a fast breeder reactor in order to be burned efficiently. In a thermal neutron spectrum, the concentrations of several heavy actinides (Curium-242 and Plutonium-240) can become quite high, creating fuel that is substantially different from the usual Uranium or mixed oxides (MOX) that most current reactors were designed to use.

Another pyrochemical process, the PYRO-B process, has been developed for the processing and recycling of fuel from a transmuter reactor ( A Fast breeder reactor designed to convert transuranic nuclear waste into fission products ). A typical transmuter fuel is free of uranium and contains recovered transuranics 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 (for volumetric oxidation) involves heating oxide fuel with oxygen, sometimes with alternating oxidation and reduction, or alternating oxidation by ozone to uranium trioxide with decomposition by heating back to triuranium octoxide. A major purpose is to capture tritium as tritiated water vapor before further processing where it would be difficult to retain the tritium. Other volatile elements leave the fuel and must be recovered, especially iodine, technetium, and carbon-14. Voloxidation also breaks up the fuel or increases its surface area to enhance penetration of reagents in following reprocessing steps.

Volatilization in isolation

Simply heating spent oxide fuel in an inert atmosphere or vacuum at a temperature between 700°C and 1000°C as a first reprocessing step can remove several volatile elements, including caesium whose isotope Cs-137 emits about half of the heat produced by the spent fuel over the following 100 years of cooling (however, most of the other half is from Sr-90 which remains).The estimated overall mass balance for 20,000 grams of processed fuel with 2,000 grams of cladding is:

Input Residue Zeolite



Palladium 28 14 14
Tellurium 10 5 5
Molybdenum 70 70
Caesium 46 46
Rubidium 8 8
Silver 2 2
Iodine 4 4
Cladding 2000 2000
Uranium 19218 19218 ?
Others 614 614 ?
Total 22000 21851 145 4 0
Tritium is not mentioned in this paper.

Fluoride volatility

In the fluoride volatility process, fluorine is reacted with the fuel. Fluorine is so much more reactive than even oxygen that small particles of ground oxide fuel will burst into flame when dropped into a chamber full of fluorine. This is known as flame fluorination; the heat produced helps the reaction proceed. Most of the uranium, which makes up the bulk of the fuel, is converted to uranium hexafluoride, the form of uranium used in uranium enrichment, which has a very low boiling point. Technetium, the main long-lived fission product, is also efficiently converted to its volatile hexafluoride. A few other elements also form similarly volatile hexafluorides, pentafluorides, or heptafluorides. The volatile fluorides can be separated from excess fluorine by condensation, then separated from each other by fractional distillation or selective reduction. Uranium hexafluoride and technetium hexafluoride have very similar boiling points and vapor pressures, which makes complete separation more difficult.

Many of the fission products 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. 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 actinides to the molten fuel mixture for eventual fission, while removing fission products that are neutron poisons, 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 fluorides will also form volatile high-valence chlorides. Chlorination and distillation is another possible method for separation. The sequence of separation may differ usefully from the sequence for fluorides; for example, zirconium tetrachloride and tin tetrachloride have relatively low boiling points of 331°C and 114.1°C. Chlorination has even been proposed as a method for removing zirconium fuel cladding, 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.


The relative economics of reprocessing-waste disposal and interim storage-direct disposal has been the focus of much debate over the past ten years. Studies have 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 fast breeder reactor cycle) and compare this to the total costs of an open fuel cycle with direct disposal. The range of results produced by these studies is very wide, but all are agreed that under current (2005) economic conditions the reprocessing-recycle option is the more costly.

If reprocessing is undertaken only to reduce the radioactivity level of spent fuel it should be taken into account that spent nuclear fuel becomes less radioactive over time. After 40 years its radioactivity drops by 99.9%, though it still takes over a thousand years for the level of radioactivity to approach that of natural uranium. However the level of transuranic elements,including plutonium-239, remains high for over 100,000 years, so if not reused as nuclear fuel, then those elements need secure disposal because of nuclear proliferation reasons as well as radiation hazard.

List of sites

Country Reprocessing site Fuel type Procedure Reprocessing

capacity tU/yr

or operating period
Mol LWR, MTR (Material test reactor) 80 1966-1974
Karlsruhe, WAK LWR 35 1971-1990
Marcoule, UP 1 Military 1,200 1958-1997
Marcoule, CEA APM FBR PUREX DIAMEX SANEX 6 1988- present
La Hague, UP 2 LWR PUREX 900 1967-1974
La Hague, UP 2-400 LWR PUREX 400 1976-1990
La Haguemarker, UP 2-800 LWR PUREX 800 1990
La Haguemarker, UP 3 LWR PUREX 800 1990
Windscale Magnox 1,000 1956-1962
Sellafieldmarker, B205 Magnox PUREX 1,500 1964
Dounreaymarker FBR 8 1980
THORPmarker LWR PUREX 1,200 1990
Rotondella Thorium 5 1968 (shutdown)
Kalpakkammarker Military 100 1998
Trombaymarker Military PUREX 60 1965
Tarapurmarker CANDU 100 1982
Tokaimura LWR 210 1977
Rokkashomarker LWR 800 2005
Mayak Plant B Military 400 1948-196?
Mayak Plant BB, RT-1 LWR PUREX + Np separation 400 1978
Zheleznogorskmarker (Krasnoyarsk-26), RT-2 WWER 1,500 under construction
, NYmarker West Valley LWR 300 1966-1972

See also


  1. Information from the World Nuclear Association about MOX
  2. Nuclear Fuel Reprocessing: U.S. Policy Development
  3. [1] [2] [3]
  5. Elecrochemical Behaviours of Lanthanide Fluorides in the Electrolysis System with LiF-NaF-KF Salt
  6. The Merck Group - Home


  • OECD Nuclear Energy Agency, 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.

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