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• Used nuclear fuel has long been reprocessed to extract fissile materials for recycling and to reduce the volume of high-level wastes.
• Increasingly the focus is on processing to separate (partition) individual constituents either for segregated waste disposal or for transmutation of long-lived isotopes into shorter-lived ones.
• Pyro-processing is one technique being promoted.

Reprocessing of used fuel

Over the last fifty years the principal reason for reprocessing has been to recover unused uranium and plutonium in the used fuel elements. A secondary reason is to reduce the volume of material to be disposed of as high-level waste. In addition, the level of radioactivity in such 'light' waste after about 100 years falls much more rapidly than in used fuel itself.

In the last decade interest has grown in separating ('partitioning') individual radionuclides both to reduce long-term radioactivity in residual wastes and to be able to transmute separated long-lived radionuclides into shorter-lived ones.

Reprocessing to recover uranium and plutonium avoids the wastage of a valuable resource because most of the used fuel (uranium at less than 1% U-235 and a little plutonium) can be recycled as fresh fuel, saving some 30% of the natural uranium otherwise required. It also avoids leaving the plutonium in the used fuel, where in a century or two the built-in radiological protection will have diminished, possibly allowing it to be recovered for weapons.

So far, more than 80,000 tonnes of used fuel from commercial power reactors has been reprocessed for U & Pu recovery, and annual capacity is now almost 5000 tonnes per year.

Used fuel assemblies removed from a reactor are very radioactive and produce heat. They are therefore put into large tanks or "ponds" of water which cool them, and three metres of water over them shields the radiation. Here they remain, either at the reactor site or at the reprocessing plant, for a number of years as the level of radioactivity decreases considerably. For most types of fuel, reprocessing occurs anything from 5 to 25 years after reactor discharge.

All commercial reprocessing plants use the well-proven hydrometallurgical PUREX process. This involves dissolving the fuel elements in concentrated nitric acid. Chemical separation of uranium and plutonium is then undertaken by solvent extraction steps. (Neptunium can also be recovered if required, and maybe used for producing Pu-238 for thermo-electric generators for spacecraft.) The Pu and U can be returned to the input side of the fuel cycle - the uranium to the conversion plant prior to re-enrichment and the plutonium straight to fuel fabrication .

The remaining liquid after Pu and U are removed is high-level waste, containing about 3% of the used fuel in the form of fission products and minor actinides (Np, Am, Cm). It is highly radioactive and continues to generate a lot of heat. It is conditioned by calcining and incorporation of the dry material into borosilicate glass, then stored pending disposal. In principle any compact, stable, insoluble solid is satisfactory for disposal.

Another version of PUREX has the minor actinides (americium, neptunium, caesium) being separated in a second aqueous stage and then directed to an accelerator-driven system cycling with pyroprocessing for transmutation (see later sections). The waste stream then contains largely fission products.

The PUREX process may also be modified to enable recovery of iodine by volatilisation and of technetium by electrolysis. French CEA research has shown 95% and 90% recoveries respectively. The same research effort has demonstrated separation of caesium.

Recently (1999) a variation of PUREX has been proposed by the US Dept of Energy for civil wastes. In this, only uranium is recovered (hence UREX process) for disposal as low-level waste, with iodine and technetium also being recovered at the head end. The residual is treated by pyroprocessing to recover transuranic elements (actinides) including plutonium for transmutation. The fission products become the main high-level waste. A major goal of this system is to keep the plutonium with the other transuranics which are destroyed by transmutation. Nevertheless, one version also has Pu recovered for recycling commercially as fuel, as in Europe, but at present contrary to US policy.

History

A great deal of reprocessing has been going on since the 1940s, mainly for military purposes, to recover plutonium for weapons. In the UK, metal fuel elements from the first generation gas-cooled commercial reactors have been reprocessed at Sellafield for about 40 years. The 1500 t/yr plant has been successfully developed to keep abreast of evolving safety, hygiene and other regulatory standards. From 1969 to 1973 oxide fuels were also reprocessed, using part of the plant modified for the purpose. A new 900 t/yr thermal oxide reprocessing plant (THORP) was commissioned in 1994 and the corresponding MOX plant in 2001.

In the USA, no civil reprocessing plants are now operating, though three have been built. The first, a 300 t/yr plant at West Valley, NY, was operated successfully from 1966-72. However, escalating regulation required plant modifications which were deemed uneconomic, and the plant was shut down. The second was a 300 t/yr plant built at Morris, Illinois, incorporating new technology which, although proven on a pilot-scale, failed to work successfully in the production plant. The third was a 1500 t/yr plant at Barnwell, South Carolina, which was aborted due to a change in government policy which ruled out all US civilian reprocessing as one facet of US non-proliferation policy. In all, the USA has over 250 plant-years of reprocessing operational experience, the vast majority being at government-operated defence plants since the 1940s.

In France one 400 t/yr reprocessing plant is operating for metal fuels from gas-cooled reactors at Marcoule. At La Hague, reprocessing of oxide fuels has been done since 1976, and two 800 t/yr plants are now operating. India has a 100 t/yr oxide fuel plant operating at Tarapur with others at Kalpakkam and Trombay, and Japan is building a major plant at Rokkasho while having had most of its used fuel reprocessed in Europe meanwhile. It has had a small (100 t/yr) plant operating. Russia has a 400 t/yr oxide fuel reprocessing plant at Ozersk (Chelyabinsk).

In France EdF has made provision to store reprocessed uranium (RepU) for up to 250 years as a strategic reserve. Currently, reprocessing of 1150 tonnes of EdF used fuel per year produces 8.5 tonnes of plutonium (immediately recycled as mixed oxide - MOX - fuel) and 815 tonnes of RepU. Of this about 650 tonnes is converted into stable oxide form for storage. EdF has demonstrated the use of RepU in its 900 MWe power plants, but it is currently uneconomic due to conversion costing three times as much as that for fresh uranium, and enrichment needing to be separate because of U-232 and U-236 impurities (the former gives rise to gamma radiation, the latter means higher enrichment is required).

Partitioning goals

Several factors give rise to a more sophisticated view of reprocessing today, and use of the term partitioning reflects this. First, new management methods for high and intermediate-level nuclear wastes are under consideration, notably partitioning-transmutation (P&T;) and partitioning-conditioning (P&C;), where long-lived radionuclides are the prime objectives to separate out. Secondly, new fuel cycles such as those for fast neutron reactors (including a lead-cooled one) and fused salt reactors, and the possible advent of accelerator-driven systems, require a new approach to reprocessing. Here the focus is on pyrometallurgical processes ('pyroprocessing') in a molten salt bath, with electrochemical separation.

The main radionuclides targeted for separation for P&T; or P&C; are the actinides neptunium, americium and curium (along with U & Pu), and the fission products iodine (I-129), technetium (Tc-99), caesium (Cs-135) and strontium (Sr-90). Removal of the latter two significantly reduces the heat load of residual conditioned wastes. In Japan, platinum group metals are also targeted, for commercial recovery. Of course any chemical process will not discriminate different isotopes of any element.

Efficient separation methods are needed to achieve low residuals of long-lived radionuclides in conditioned wastes and high purities of individual separated ones in transmutation targets . Otherwise any transmutation effort is a random process with uncertain results. In particular one does not want fertile U isotopes in a transmutation target or it will generate further highly radiotoxic transuranic isotopes.

Achieving effective full separation for any transmutation program is likely to mean pyroprocessing of residuals from the PUREX or similar aqueous processes.

A BNFL-Cogema study in 2001 reported that 99% removal of actinides, Tc-99 & I-129 would be necessary to justify the effort in reducing the radiological load in a waste repository. A U.S. study identified a goal of 99.9% removal of the actinides and 95% removal of technetium and iodine. In any event, the balance between added cost and societal benefits is the subject of considerable debate.

Pyro-processing

Pyrometallurgical processing ('pyroprocessing') to separate nuclides from a radioactive waste stream involve several techniques: 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.

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. However, such processes are at an early stage of development compared with hydrometallurgical processes already operational.

Separating (partitioning) the actinides contained in a fused salt bath involves electrodeposition on a cathode, extraction between the salt bath and a molten metal (eg Li), or oxide precipitation from the salt bath.

Many pyroprocessing techniques are at an early stage of development, and only one has been licensed for use on a significant scale. This is the US IFR process developed by Argonne National Laboratory and used for pyroprocessing the used fuel from EBR-II experimental fast reactor which ran from 1963-1994. This application is essentially a P&C; process, because neither plutonium nor other fissile transuranics are recovered for recycle. The process is used to facilitate the disposal of a fuel that could not otherwise be sent directly to a geologic repository. The uranium metal fuel is dissolved in LiCl+KCl molten bath, the U is deposited on a solid cathode, while the stainless steel cladding and noble metal fission products remain in the anode and are consolidated by melting to form a durable metallic waste form. The transuranics and fission products in salt are then incorporated into a zeolite matrix which is hot pressed into a ceramic composite waste. The highly-enriched uranium recovered from the EBR-II driver fuel is down-blended to less than 20% enrichment and stored for possible future use.

The PYRO-A process, being developed at Argonne to follow the UREX process, is a pyrochemical process for the separation of transuranic elements and fission products contained in the oxide powder resulting from denitration of the UREX raffinate. The nitrates in the residual raffinate acid solution are converted to oxides, which are then reduced electrochemically in a LiCl-Li₂O molten salt bath. The more chemically active fission products (e.g., Cs, Sr) are not reduced and remain in the salt. The metallic product is electrorefined in the same salt bath to separate the transuranic elements on a solid cathode from the rest of the fission products. The salt bearing the separated fission products is then mixed with a zeolite to immobilize the fission productsin a ceramic composite waste form. The cathode deposit of transuranic elements is then processed to remove any adhering salt and is formed into ingots for subsequent fabrication of transmutation targets.

Another pyrochemical process, the PYRO-B process, has been developed for the processing and recycle 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.

Transmutation

Transmutation of one radionuclide into another is achieved by neutron bombardment in a nuclear reactor or accelerator-driven device. A high-energy proton beam hitting a heavy metal target produces a shower of neutrons by spallation. The neutrons can cause fission in a subcritical fuel assembly, but unlike a conventional reactor, fission ceases when the accelerator is turned off. The fuel may be uranium, plutonium or thorium, possibly mixed with long-lived wastes from conventional reactors. See also paper on Accelerator-driven Nuclear Energy.

The objective is to change long-lived actinides and fission products into significantly shorter-lived nuclides. The goal is to have wastes which become radiologically innocuous in only a few hundred years.

Some radiotoxic nuclides, such as Pu-239 and the long-lived fission products Tc-99 and I-129, can be transmuted (fissioned, in the case of Pu-239) with thermal (slow) neutrons. The minor actinides Np, Am and Cm (as well as the higher isotopes of plutonium), all highly radiotoxic, are much more readily destroyed by fissioning in a fast neutron energy spectrum, where they can also contribute to the generation of power.

With repeated recycle in a transmutation system, the radiotoxicity of used nuclear fuel can be reduced to the point that, after a decay period of less than 1000 years, it is less toxic than the uranium ore originally used to produce the fuel. The need for a waste repository is certainly not eliminated, but the hazard posed by the disposed waste materials is greatly reduced.

Main sources:

Madic, C. 2000, Overview of the hydrometallurgical and pyrometallurgical processes É. for partitioning high-level nuclear wastes, inActinide and Fission Product Partitioning and Transmutation, Madrid, OECD/NEA.
Laidler, J.J. 2000, Pyrochemical separations technologies envisioned for the US accelerator transmutation waste system, OECD/NEA workshop proceedings:Pyrochemical Separations; - also personal communication.
NuclearFuel 15/10/01 & 31/1/05.