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Nuclear reprocessing separates any usable elements (e.g., uranium and plutonium) from fission products and other materials in spent nuclear reactor fuels. Usually the goal is to recycle the reprocessed uranium or place these elements in new mixed oxide fuel (MOX), but some reprocessing is done to obtain plutonium for weapons. It is the process that partially closes the loop in the nuclear fuel cycle.
Use of breeder reactors combined with reprocessing could extend the usefulness of mined uranium by more than 60 times. [1]
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 both uranium and plutonium in decontaminated form was developed at ORNL in 1949. It was this process, named PUREX, (see below) which became the current standard method. 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 March 1977, fear of nuclear weapons proliferation (especially after India demonstrated nuclear weapons capabilities using reprocessing technology) led President Jimmy Carter to issue a Presidential directive to indefinitely suspend the commercial reprocessing and recycling of plutonium in the U.S. After that, only countries that already had large investments in reprocessing infrastructure continued to reprocess spent nuclear fuel.
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.
PUREX, the current main method
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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 completely developed and widely used process in the industry at present. If used on fuel from commercial power reactors, plutonium extracted using PUREX typically contain 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]
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. [3]
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 disadvantge 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]
Butex, β,β'-dibutyoxydiethyl ether
A process based on a solvation extraction process using the triether extractant named above. This process has the disadvantge 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.
Aqueous methods under development
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.
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.
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.
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][7][8] Other systems such as the dithiophosphinic acids are being worked on by some other workers.
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. [6][7] The chemistry is based upon the interaction of cesium 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 [8]have been suggested as well.
Electrochemical method in aqueous alkali
An exotic method using electrochemistry and ion exchange in ammonium carbonate has been reported.[9]
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. 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.
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 an advantage as it might first appear as it is possible to treat normal oxide fuel using a process called Voloxidation[10] 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. This is generally considered a fairly desirable property. 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.
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 cesium, zirconium and strontium) remain in the salt. [11][12][13] As alternatives to the moltern cadmium electrode it is possible to use a molten bismuth cathode, or a solid aluminium cathode [14]
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.[15]
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. [9]
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
Voloxidation (for volumetric oxidation) involves heating oxide fuel with oxygen, sometimes with alternating oxidation and reduction steps, or alternating oxidation by ozone to uranium trioxide with decomposition back to triuranium octoxide [16]. The most common 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 can and should be recovered, especially iodine and carbon-14.
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 cesium 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 20000 grams of processed fuel with 2000 grams of cladding is: [17]
Tritium is not mentioned in this paper.
Fluoride volatility
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Image:Fission yield volatile.png
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, possibly after pretreatment with hydrogen fluoride. The uranium is converted to uranium hexafluoride, the form of uranium used in uranium enrichment, which has a very low boiling point and can be distilled away from the remainder. A few other elements also form similarly volatile hexafluorides, pentafluorides, or heptafluorides, and must be separated from the uranium hexafluoride by fractional distillation or selective reduction.
Some of the same elements are volatilized as in the non-fluoridated volatilization process described in the last section, such as iodine, tellurium and molybdenum. Some notable differences are that technetium is volatilized, but cesium is not and remains in the molten mixture of fluoride salts.
Plutonium can form volatile fluorides, but not as readily as uranium, so its degree of separation from uranium may vary, depending on the specifics of the process. Americium and curium do not form volatile fluorides and will remain with the alkaline fission products. Some noble metals may not form fluorides at all, but remain in metallic form.
Distillation of the residue at higher temperatures is a possibility for separation of 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, so this may not be as convenient as the preceding low-temperature distillation; however, if a carrier salt like lithium fluoride or sodium fluoride is used, high-temperature distillation may be a way to separate the carrier salt for reuse.
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[10], 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, cesium are more soluble than those of uranium, neptunium, plutonium, and zirconium.
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 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 radioactive 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% [18], though it still takes over a thousand years for the level of radioactivity to approach that of natural uranium [19]. 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 nuclear reprocessing sites
See also
References
- ^ Supply of Uranium: Nuclear issues briefing paper #75. Uranium Information Centre,. Retrieved on 2007-06-17.
- ^ Plutonium Recovery from Spent Fuel Reprocessing by Nuclear Fuel Services at West Valley, New York from 1966 to 1972. U.S. Department of Energy (February 1996). Retrieved on 2007-06-17.
- ^ Gerber, Michelle. The plutonium production story at the Hanford Site: processes and facilities history (WHC-MR-0521)(excerpts). Department of Energy.
- ^ Seaborg, Glenn T. et al (1960-08-23). Method for separation of plutonium from uranium and fission products by solvent extraction. U.S. Patent and Trademark Office.
- ^ L.W. Gray (1999-04-15). From separations to reconstitution--a short history of plutonium in the U.S. and Russia (UCRL-JC-133802). Lawrence Livermore National Laboratory preprint.
- ^ U.S.-Russia Team Makes Treating Nuclear Waste Easier. U.S. embassy press release(?) (2001-12-19). Retrieved on 2007-06-14.
- ^ J. Banaee et al. (2001-09-01). INTEC High-Level Waste Studies Universal Solvent Extraction Feasibility Study. INEEL Technical report.
- ^ J.D. Law et al. (2001-03-01). Flowsheet testing of the universal solvent extraction process for the simultaneous separation of cesium, strontium, and the actinides from dissolved INEEL calcine. WM 2001 conference proceedings. Retrieved on 2006-06-17.
- ^ Asanuma, Noriko, et al.. "Andodic dissociation of UO2 pellet containing simulated fission products in ammonium carbonate solution". Journal of Nuclear Science and Technology 43: 255-262.
- ^ a b
- ^ Development of pyro-process fuel cell technology. CRIEPI News (July 2002).
- ^ Masatoshi Iizuka (2001-12-12). Development of plutonium recovery process by molten salt electrorefining with liquid cadmium cathode. Proceedings of the 6th information exchange meeting on actinide and fission product partitioning and transmutation (Madrid, Spain).
- ^ [1]
- ^ [2]
- ^ [3]
- ^ (page 17 of [4])
- ^ Wolverton, Daren et al. (2005-05-11). Removal of cesium from spent nuclear fuel destined for the electrorefiner fuel treatment process. University of Idaho (dissertation?).
- ^ [5]
- ^ Radioactive wastes: myths and realities. World Nuclear Association (2006-06). Retrieved on 2007-06-14.
- 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.
External links
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