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Nuclear reprocessing

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Nuclear reprocessing separates any usable elements (e.g., uranium and plutonium) from fission products and other materials in used nuclear reactor fuels. Usually the goal is to place these elements in new mixed oxide fuel (MOX), but some reprocessing is done to obtain plutonium for weapons. It is the process that closes the loop in the nuclear fuel cycle.

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 from 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 ORNL in the 1943-1945 period to produce quantities of plutonium for evaluation and use in weapons programs.

The Bismuth Phosphate process was first operated on a large scale at Hanford, Washington, in the latter part of 1944. It was successful for plutonium separation in the emergency situation existing then, but it had a significant weakness, namely, 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 Argonne National Laboratory (ANL) soon after World War II. It was this process, named PUREX, (see below) which became the current standard method.

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. Other nations, have not followed suit and continued to reprocess spent nuclear fuel.

However in March 1999, the U.S. Department of Energy (DOE) 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. [1]

Spent nuclear fuel

Spent low enriched uranium fuel contains:

  • 3% of the mass consists of fission products of 235U (also indirect products in the decay chain), nuclear poisons considered radioactive waste or separated further for various industrial and medical uses. The fission products include every element from zinc through to the lanthanides, much of the fission yield is concentrated in two peaks, one in the second transition row (Zr, Mo, Tc, Ru, Rh, Pd) while the other is later in the periodic table (I, Xe, Cs, Ba, La, Ce, Nd). Many of the fission products are either non radioactive or only shortly lived radioisotopes. But a considerable number are medium to long lived radioisotopes such as 90Sr, 137Cs, 99Tc and 129I.
  • 1% of the mass is 239Pu and 240Pu resulting from conversion of 238U, which may either be considered a useful by-product, or as dangerous and inconvenient waste. One of the main concerns regarding nuclear proliferation is to prevent this plutonium from being used by states other than those already established as Nuclear Weapons States, to produce nuclear weapons. If the reactor has been used normally, the plutonium is reactor-grade, not weapon-grade: it contains much 240Pu and less than 80% 239Pu, which makes it less suitable, but not impossible, to use in a weapon [2]. If the irradiation period has been short then the plutonium is weapon-grade (more than 80%, up to 93%).
  • 96% of the mass is the remaining uranium: most of the original 238U and a little 235U. Usually 235U would be less than 0.83% of the mass.
  • Traces of the minor actinides. In present in used reactor fuel are the minor actinides, these are actinides other than uranium and plutonium. These include americium and curium. The amount formed depends greatly upon the nature of the fuel used and the conditions under which it was used. For instance the use of MOX fuel (239Pu in a 238U matrix) is likely to lead to the production of more 241Am than the use of a uranium/thorium based fuel (233U in a 232Th matrix). Also present as a minor actinide is 237Np, this neptunium isotope is fissile but also can be converted into 238Pu by neutron bombardment.

For natural uranium fuel: Fissile component starts at 0.71% 235U concentration in natural uranium). At discharge, total fissile component still 0.50% (0.23% 235U, 0.27% fissile 239Pu, 241Pu) Fuel is discharged not because it is fully used-up, but because the neutron-absorbing fission products have built up and the fuel then becomes significantly less able to sustain a nuclear reaction.

Some natural uranium fuels use chemically active cladding, such as Magnox, and need to be reprocessed because long-term storage and disposal is difficult [3].

For highly enriched fuels used in marine reactors and research reactors the isotope inventory will vary based on in-core fuel management and reactor operating conditions.

Old Methods which are no longer used

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 co-precipitated 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 a 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. [4]

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) is required to increase the nitrate concentration in the aqueous phase to obtain a resonable distribution ratio (D value). Also hexone is degraded by concentrated nitric acid. This process has been replaced by PUREX.[5][6]

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) is required to increase the nitrate concentration in the aqueous phase to obtain a resonable distribution ratio. This process was used at Windscale many years ago. This process has been replaced by PUREX.

Current methods which are in use

File:Extracteduraniumcomplex.jpg
The complex formed from a uranyl ion, two nitrates and two molecules of triethyl phosphate this is very similar to the uranium complex which is in the organic phase in the PUREX process

PUREX

Overview of PUREX

This process can be used to recover weapon-grade materials from spent nuclear reactor fuel, and as such, its component chemicals are monitored. 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.

The chemical process

The irradiated fuel is first dissolved into nitric acid. After the dissolution step it is normal to remove the fine insoluble solids because otherwise they will disturb the solvent extraction process by altering the liquid liquid interface. It is known that the presence of a fine solid can stabilize an emulsion. Emulsions are often referred to as third phases in the solvent extraction community.

An organic solvent composed of 30% tributyl phosphate (TBP) in odorless kerosene (or hydrogenated propylene trimer) is used to recover the uranium and plutonium; the fission products remain in the aqueous nitric phase. Once separated from the fission products, further processing allows separation of the heavier plutonium from the uranium. The PUREX extraction process uses a 'solvation' liquid-liquid extraction process in which a complex is formed between the tributyl phosphate and the extracted actinides. The extraction is favoured by a high nitric acid concentration and the back extraction (stripping) is favoured by a low nitric acid concentration. For the plutonium back extraction it is possible to use redox stripping in which the oxidation state of the plutonium is lowered by the action of a reducing agent.

Degradation products of TBP

It is normal to extract both the uranium and plutonium from the majority of the fission products but it is the case that it is not possible to get an acceptable separation of the fission products from the actinide products with a single extraction cycle. The irradation of the tributyl phosphate / hydrocarbon mixture produces dibutyl hydrogen phosphate. This degradation product is able to act as an extraction agent for many metals, hence leading to the contamination of the product by fission products. Hence it is normal to use more than one extraction cycle, the first cycle is to lower the radioactivity level of the mixture so allowing the later extraction cycles to be kept cleaner (in terms of degradation products). Reference J.H. Burns, Inorganic Chemistry, 1983, 22, 1174.

The dialkyl hydrogen phosphates are able to form complexes with many metals, these include some polymeric metal complexes. The formation of these coordination polymers is one way in which fine solid can be formed in the process. While the cadmium concentration in both the fuel dissolution liquor and the PUREX raffinate is very low, the polymeric complex of cadmium of diethyl phosphate is shown as an example.

File:Cadmiumdobadz.jpg
The complex is formed from cadmium ions and diethyl phosphate ions

Here is the strucutre of a lanthanide complex of diethyl phosphate, unlike cadmium the concentration of neodynium in these mixtures formed from fuel is very high.

File:Neodyniumdobadz.jpg
The complex is formed neodynium ions and diethyl phosphate

Here is a mixed tributyl phosphate dibutyl phosphate complex of uranium, becuase the dibutyl phosphate ligands are acidic it will now be possible to extract uranium by a ion exchange liquid-liquid extraction mechanism rather than only by a solvation mechanism. This will potentially make the stripping of uranium with dilute nitric acid less effective.

The complex is formed formed from uranyl ions, two nitrates, two dibutyl phosphates and two molecules of tributyl phosphate
Extraction of technetium

In addition the uranium(VI) tributyl phosphate system is able to extract technetium as pertectinate through an ion pair extraction mechanism. Here is an example of a rhenium version of a uranium / technetium complex which is thought to be responsible for the extraction of technetium into the organic phase.

Here are two pictures of actinyl complexes of triphenyl phosphineoxide which have been crystalised with perrhenate. With the less highly charged neptunyl ion it is also possible to form a complex.

The complex is formed from a uranyl ion and three molecules of triphenyl phosphine oxide, the anions are in the first coordination sphere of the metal
The complex is formed formed from a neptunyl ion and four molecules of triphenyl phosphine oxide, the anions are separated from the metal centre

Reference G.H. John,I. May,M.J. Sarsfield,H.M. Steele,D. Collison,M. Helliwell and J.D. McKinney, Dalton Trans., 2004, 734.

The organic soluble complex

The nature of the organic soluble uranium complex has been the subject of some research, a series of complexes of uranium with nitrate and trialkyl phosphates and phosphine oxides have been characterised.

Possible methods for future use

Aqueous methods

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.

The UREX process is a PUREX process which has been modified to prevent the plutonium 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 this is a process which 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 EXraction. 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 has 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.

References:


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.[7][8] The chemistry is based upon the interaction of cesium and strontium with poly ethylene oxide (poly ethylene glycol)[9] 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 [10]have been suggested as well.

Non aqueous methods

Pyroprocessing is a generic term for Pyrometallurgy processes. These processes are not currently in significant use worldwide, but they have been researched and developed at INEEL, 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.

Advantages

  • 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 [11]which removes 99% of the tritium from used fuel. The tritium can be reovered 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 weapons grade 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 fission products are not nearly as long lived as they would otherwise be. Most of the long term (past a couple hundred years) radioactivity produced by nuclear waste is produced by the remaining 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

In nuclear engineering, PYRO-A refers to a type of nuclear reprocessing. More specifically, it is a means of separating actinides (elements within the actinide family, generally heavier than U-235) from non-actinides. The basic process is described here [12][13][14]. The principles are quite simple. The spent fuel is placed in an anode basket which is in contact with the molten salt. Then an electrical current is applied, the uranium will plate out as the conductive uranium dioxide 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. As alternatives to the moltern cadmium electrode it is possible to use a molten bismuth cathode[15], 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 of, 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.

Pyro-B

Another pyrochemical process, the PYRO-B process, has been developed for the processing and recycling of fuel from a transmuter reactor. 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.

Fluoride volatility

A related process exists in which the uranium is distilled away after conversion to uranium hexaflouride, this is the Fluoride volatility process.

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. Many approaches have been used and to a certain extent the approach taken has determined the outcome of the assessment. These studies model the total fuel cycle costs of a reprocessing-recycling system based on thermal recycling of plutonium 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 1000 years for the level of radioactivity to approach that of natural uranium [19].

list of nuclear reprocessing sites

See also

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.
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