Nuclear reprocessing

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Nuclear reprocessing separates components of spent nuclear fuel such as:

• Reprocessed uranium
• Plutonium
• Minor actinides
• Fission products
• Activation products
• Cladding

Reprocessing serves multiple purposes, whose relative importance has changed over time:

• Producing plutonium as fuel for nuclear weapons
• Recycling plutonium, uranium, and other actinides as fuel for breeder reactors, closing the nuclear fuel cycle and extending the energy extractable per original unit of uranium by a factor of more than 60.^[1]
• After breeder reactors were postponed, recycling plutonium once as MOX fuel for thermal reactors, extending energy extracted by only about 30%
• Management of nuclear waste radioactivity by separating:
  • actinides for destruction by nuclear fission
  • some long-lived fission products (⁹⁹Tc, ¹²⁹I) and activation products for destruction by nuclear transmutation by neutron capture
  • other long-lived fission products and activation products for very long term storage in a deep geological repository which may have limited capacity
  • bulky but low-level waste such as reprocessed uranium for less stringent disposal
  • shorter-term secure storage for decay of dangerous medium-lived radionuclides such as ¹³⁷Cs and ⁹⁰Sr
  • radionuclides for industrial uses

[edit] History

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 Laboratory (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 Site, 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 Site and a smaller plant at West Valley, New York which closed by 1972.^[2]

Processing of civilian fuel has long been employed in Europe (at the COGEMA La Hague site) and briefly at the West Valley Reprocessing Plant in the U.S.

In October 1976, fear 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. This was confirmed by President Jimmy Carter in 1977. 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 comprised of Duke Energy, COGEMA, and Stone & Webster (DCS) to design and operate a Mixed Oxide (MOX) fuel fabrication facility. Site preparation at the Savannah River Site (South Carolina) began in October of 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.

[edit] Aqueous / organic solvent methods

[edit] Obsolete methods

[edit] 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.^[3]

[edit] 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.^[4][5]

Pu⁴⁺ + 4NO₃^- + 2S --> [Pu(NO₃)₄S₂]

[edit] Butex, β,β'-dibutyoxydiethyl ether

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

[edit] PUREX, the current method

Main article: PUREX

PUREX 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.^{[citation needed]}

[edit] UREX

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 Yucca Mountain, 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.

[edit] 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 (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.

[edit] DIAMEX

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.

[edit] SANEX

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-triaiznyl pyridine (BTP) based process.^[6] Other systems such as the dithiophosphinic acids are being worked on by some other workers.

[edit] UNEX

This is the UNiversal EXtraction process which was developed in Russia and the Czech Republic, it is a process designed to remove all of the most troublesome (Sr, Cs and minor actinides) radioisotopes from the raffinates left after the extraction of uranium and plutonium from used nuclear fuel.^[7][8] 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^[9] have been suggested as well.

[edit] Electrochemical method in aqueous alkali

An exotic method using electrochemistry and ion exchange in ammonium carbonate has been reported.^[10]

[edit] Pyroprocessing

Pyroprocessing is a generic term for several kinds of Pyrometallurgical Reprocessing. These processes are not currently in significant use worldwide, but they have been researched and developed at Argonne National Laboratory and elsewhere. The principles behind them are well understood, and no significant technical barriers exist to their adoption^{[citation needed]}. The primary economic hurdle to widespread adoption is that 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.

Pyrometallurgical processing techniques involve several stages: volatilisation, liquid-liquid extraction using immiscible metal-metal phases or metal-salt phases, electrorefining in molten salt, fractional crystallisation, etc. They are generally based on the use of either fused (low-melting point) salts such as chlorides or fluorides (eg LiCl+KCl or LiF+CaF2) or fused metals such as cadmium, bismuth or aluminium. They are most readily applied to metal rather than oxide fuels.

[edit] Advantages and disadvantages

Advantages

• Pyroprocessing can readily be applied to high burn-up fuel and fuel which has had little cooling time, since the operating temperatures are high already.
• It does not use water. Water is problematic in nuclear chemistry for many reasons. First of all, it tends to serve as a moderator, and accelerate nuclear reactions. Secondly, it is easily contaminated, and not easily cleaned up, and it tends to evaporate, potentially taking tritium with it. This is not as large a disadvantage as it might first appear as it is possible to treat normal oxide fuel using a process called Voloxidation^[11] which removes 99% of the tritium from used fuel. The tritium can be recovered in the form of a strong solution which might be suitable for use as a supply of tritium for industrial applications.
• It separates out all actinides, and therefore produces fuel that is heavily spiked with heavy actinides, such as Plutonium (240+), and Curium 242. This does not prevent the fuel from being suitable for reactors, but it makes it hard to manipulate, steal, or make nuclear weapons from. (However, the difficulty has been questioned.^[12]) In contrast, the PUREX process can easily produce separated Uranium and Plutonium, and also tends to leave the remaining actinides (like Curium) behind, producing more dangerous nuclear waste.
• It is somewhat more efficient and considerably more compact than aqueous processing methods, allowing the possibility of on-site reprocessing of reactor wastes. This circumvents various transportation and security issues, allowing the reactor to simply store a small volume (perhaps a few percent of the original volume of the spent fuel) of fission product laced salt on site until decommissioning, when everything could be dealt with at once.
• Since pyrometalurgy recovers all the actinides, the remaining waste is not nearly as long lived as it would otherwise be. Most of the long term (past a couple hundred years) radioactivity produced by nuclear waste is produced by the actinides. These actinides can (mostly) be consumed by reactors as fuel, so extracting them from the waste and reinserting them into the reactor reduces the long term threat from the waste, and reduces the fuel needs of the reactor.

Disadvantages

• The used salt from pyro processing is not suitable for conversion into a glass in the same way as the raffinate from PUREX processing.

[edit] 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 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.^[13][14][15] As alternatives to the moltern cadmium electrode it is possible to use a molten bismuth cathode, or a solid aluminium cathode.^[16]

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.^[17]

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.^[18]

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.

[edit] Voloxidation

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.^[11] 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.

[edit] 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:^[19]

Tritium is not mentioned in this paper.

[edit] Fluoride volatility

Main article: Fluoride volatility

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 are volatile at very high temperatures. Yields at 10^0,1,2,3 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, 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.^[20] 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.

[edit] 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,^[11] 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.

[edit] Economics of reprocessing nuclear fuel

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 modelled 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%,^[21] though it still takes over a thousand years for the level of radioactivity to approach that of natural uranium.^[22] 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.

• Recycled Nuclear Fuel Cost Calculator designed by the WISE Uranium Project

[edit] List of nuclear reprocessing sites

[edit] See also

• Nuclear fuel cycle
• Nuclear breeder reactor
• Spent nuclear fuel shipping cask
• Global Nuclear Energy Partnership announced February, 2006
• Megatons to Megawatts Program

[edit] References

• 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.

[edit] External links

• Processing of Used Nuclear Fuel, World Nuclear Association
• PUREX Process, European Nuclear Society
• Mixed Oxide Fuel (MOX) - World Nuclear Association
• Disposal Options for Surplus Weapons-Usable Plutonium - Congressional Research Service Report for Congress
• Brief History of Fuel Reprocessing