An important characteristic of nuclear energy is that used fuel may be reprocessed to recover fissile and fertile materials in order to provide fresh fuel for existing and future nuclear power plants. Several European countries, Russia and Japan have had a policy to reprocess used nuclear fuel, although government policies in some other countries have not yet fully addressed reprocessing technologies.
Over the last fifty years the principal reason for reprocessing used fuel has been to recover unused uranium (U) and plutonium (Pu) in the used fuel elements and thereby close the fuel cycle, gaining some 25% more energy from the original uranium in the process and thus contributing to energy security. A secondary reason is to reduce the volume of material to be disposed of as high-level waste (HLW) to about one fifth. In addition, the level of radioactivity in in the waste from reprocessing is much smaller and after about 100 years falls much more rapidly than in used fuel itself.
|World Commercial Reprocessing Capacity (tonnes per year)|
|LWR fuel:||France, La Hague||1700|
|UK, Sellafield (THORP)||900|
|Russia, Ozersk (Mayak)||400|
|Other nuclear fuels:||UK, Sellafield||1500|
|Total civil capacity||5550|
|Sources: OECD/NEA 2004 Nuclear Energy Data, Nuclear Eng. International handbook 2004.|
In the last decade interest has grown in recovering all long-lived actinides together (i.e. with plutonium) so as to recycle them in fast reactors so that they end up as short-lived fission products. This policy is driven by two factors: reducing the long-term radioactivity in high-level wastes, and reducing the possibility of plutonium being diverted from civil use - thereby increasing proliferation resistance of the fuel cycle. If used fuel is not reprocessed, then in a century or two the built-in radiological protection will have diminished, allowing the plutonium to be recovered for illicit use (though it is unsuitable for weapons due to the non-fissile isotopes present).
Reprocessing used fuel to recover uranium (as reprocessed uranium, or RepU) and plutonium (Pu) avoids the wastage of a valuable resource. Most of it - about 96% - is uranium, of which less than 1% is the fissile U-235 (often 0.4-0.8%); and up to 1% is plutonium. Both can be recycled as fresh fuel, saving up to 30% of the natural uranium otherwise required. The materials potentially available for recycling (but locked up in stored used fuel) could conceivably run the US reactor fleet of about 100 GWe for almost 30 years with no new uranium input.
So far, some 90,000 tonnes of used fuel from commercial nuclear power reactors has been reprocessed. Annual reprocessing capacity is now some 4000 tonnes per year for normal oxide fuels, but not all of it is operational.
Between now and 2030 some 400,000 tonnes of used fuel is expected to be generated worldwide, including 60,000 t in North America and 69,000 t in Europe.
Products of reprocessing
The composition of reprocessed uranium (RepU) depends on the initial enrichment and the time the fuel has been in the reactor, but it is mostly U-238. It will normally have less than 1% U-235 (typically about 0.5% U-235) and also smaller amounts of U-232 and U-236 created in the reactor. The U-232, though only in trace amounts, has daughter nuclides which are strong gamma-emitters, making the material difficult to handle. However, once in the reactor, U-232 is no problem (it captures a neutron and becomes fissile U-233). It is largely formed through alpha decay of Pu-236, and the concentration of it peaks after about ten years of storage.
The U-236 isotope is a neutron absorber present in much larger amounts, typically 0.4% to 0.6% - more with higher burn-up, which means that if reprocessed uranium is used for fresh fuel it must be enriched significantly more (eg one tenth) than is required for natural uranium. Thus RepU from low-burn-up fuel is more likely to be suitable for re-enrichment, while that from high burn-up fuel is best used for blending or MOX fuel fabrication.
The other minor uranium isotopes are U-233 (fissile), U-234 (from original ore, enriched with U-235, fertile), and U-237 (short half-life beta emitter). None of these affects the use of handling of the reprocessed uranium significantly. In the future, laser enrichment techniques may be able to remove these isotopes.
Reprocessed uranium (especially from earlier military reprocessing) may also be contaminated with traces of fission products and transuranics. This will affect its suitability for recycling either as blend material or via enrichment. Over 2002-06 USEC successfully cleaned up 7400 tonnes (t) of technetium-contaminated uranium from the US Department of Energy.
Most of the separated uranium (RepU) remains in storage, though its conversion and re-enrichment (in UK, Russia and Netherlands) has been demonstrated, along with its re-use in fresh fuel. Some 16,000 tonnes of RepU from Magnox reactors in UK has been used to make about 1650 tonnes of enriched AGR fuel. In Belgium, France, Germany and Switzerland over 8000 tonnes of RepU has been recycled into nuclear power plants. In Japan the figure is over 335 tonnes in tests and in India about 250 t of RepU has been recycled into PHWRs. Allowing for impurities affecting both its treatment and use, RepU value has been assessed as about half that of natural uranium.
Plutonium from reprocessing will have an isotopic concentration determined by the fuel burn-up level. The higher the burn-up levels, the less value is the plutonium, due to increasing proportion of non-fissile isotopes and minor actinides, and depletion of fissile plutonium isotopesc. Whether this plutonium is separated on its own or with other actinides is a major policy issue relevant to reprocessing (see section on Reprocessing policies below).
Most of the separated plutonium is used almost immediately in mixed oxide (MOX) fuel. World MOX production capacity is currently around 200 tonnes per year, nearly all of which is in France (see Mixed Oxide (MOX) Fuel).
Inventory of separated recyclable materials worldwide
|Quantity (tonnes)||Natural U equivalent (tonnes)|
|Plutonium from reprocessed fuel||320||60,000|
|Uranium from reprocessed fuel||45,000||50,000|
|Ex-military high-enriched uranium||230||70,000|
History of reprocessing
A great deal of hydrometallurgical reprocessing has been going on since the 1940s, originally for military purposes, to recover plutonium for weapons (from low burn-up used fuel, which has been in a reactor for only a very few months). In the UK, metal fuel elements from the first generation gas-cooled commercial reactors have been reprocessed at the Sellafield site for about 50 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.
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 reprocessing plant was a 300 t/yr plant built at Morris, Illinois that incorporated new technology that, although proven on a pilot-scale, failed to work successfully in the production plant. The third reprocessing plant was a 1500 t/yr plant at Barnwell, South Carolina whose operation was aborted due to a change in government policy that 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 acquired at government-operated defense plants since the 1940s.
In France, one 400 t/yr reprocessing plant operated for metal fuels from gas-cooled reactors at Marcoule until 1997. At La Hague, reprocessing of oxide fuels has been done since 1976, and two 800 t/yr plants are now operating, with an overall capacity of 1700 t/yr.
French utility 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 the process is currently uneconomic due to conversion costing three times as much as that for fresh uranium, and enrichment needing to be separate because of uranium-232 and uranium-236 impurities.
The plutonium is immediately recycled via the dedicated Melox mixed oxide (MOX) fuel fabrication plant. The reprocessing output in France is co-ordinated with MOX plant input, to avoid building up stocks of plutonium. If plutonium is stored for some years the level of americium-241, the isotope used in household smoke detectors, will accumulate and make it difficult to handle through a MOX plant due to the elevated levels of gamma radioactivity.
India has a 100 t/yr oxide fuel plant operating at Tarapur with others at Kalpakkam and Trombay, and Japan is starting up a major (800 t/yr) plant at Rokkasho while having had most of its used fuel reprocessed in Europe in the interim. Japan also has had a small (90 t/yr) plant operating at Tokai Mura. Russia has a 400 t/yr oxide fuel reprocessing plant at Ozersk (Chelyabinsk).
Conceptually, reprocessing can take several courses, separating certain elements from the remainder, which becomes high-level waste. Reprocessing options include:
- Separate U, Pu, (as today)
- Separate U, Pu+U (small amount of U)
- Separate U, Pu, minor actinides
- Separate U, Pu+Np, Am+Cm
- Separate U+Pu all together,
- Separate U, Pu+actinides, certain fission products.
In today's reactors, recycled uranium needs to be enriched, whereas plutonium goes straight to mixed oxide (MOX) fuel fabrication. This situation has two perceived problems: the separated plutonium is sometimes considered a proliferation risk, and the minor actinides remain in the separated waste, which means that its radioactivity is longer-lived than if it comprised fission products only.
For the future, the focus is on removing the actinides from the final waste and burning them with the recycled uranium and plutonium in fast neutron reactors. The longer-lived fission products may also be separated from the waste and transmuted in some other way.
All but one of the six Generation IV reactors being developed have closed fuel cycles with full actinide recycling. Although US policy has been to avoid reprocessing, the US budget for 2006 included $50 million to develop a plan for "integrated spent fuel recycling facilities", and a program to achieve this with fast reactors has become more explicit since.
In November 2005, the American Nuclear Society (ANS) released a position statement, saying that it:
"...believes that the development and deployment of advanced nuclear reactors based on fast-neutron fission technology is important to the sustainability, reliability and security of the world's long-term energy supply." This will enable "...extending by a hundred-fold the amount of energy extracted from the same amount of mined uranium."
The statement envisages on-site reprocessing of used fuel from fast reactors and says that:
"...virtually all long-lived heavy elements are eliminated during fast reactor operation, leaving a small amount of fission product waste which requires assured isolation from the environment for less than 500 years."
GE Hitachi Nuclear Energy (GEH) is developing this concept by combining electrometallurgical separation (see section on Electrometallurgical 'pyroprocessing' below) and burning the final product in one or more of its PRISM fast reactors on the same site. The first two stages of the separation remove uranium which is recycled to light water reactors, then fission products which are waste, and finally the actinides including plutonium.
In mid 2006 a report by the Boston Consulting Group for Areva showed that recycling used fuel in the USA could be economically competitive with direct disposal of used fuel. A $12 billion, 2500 t/yr plant was considered, with total capital expenditure of $16 billion for all related aspects. This would have the benefit of greatly reducing demand on space at the planned Yucca Mountain repository.
Boston Consulting Group gave four reasons for reconsidering US used fuel strategy which has applied since 1977:
- Cost estimates for direct disposal at Yucca Mountain had risen and capacity was limited (even if doubled)
- Increased US nuclear generation, potentially from 103 to 160 GWe
- The economics of reprocessing and associated waste disposal have improved
- There is considerable experience with civil reprocessing as of the year 2010.
Reprocessing Today - PUREX
All commercial reprocessing plants use the well-proven hydrometallurgical PUREX (Plutonium Uranium EXtraction) 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). 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.
Alternatively, some small amount of recovered uranium can be left with the plutonium which is sent to the MOX plant, so that the plutonium is never separated on its own. This is known as the COEX (co-extraction of actinides) process, developed in France as a 'Generation III' process, but not yet in use (see next section). Japan's new Rokkasho plant uses a modified PUREX process to achieve a similar result by recombining some uranium before denitration, with the main product being 50:50 mixed oxides.
In either case, the remaining liquid after Pu and U are removed is high-level waste, containing about three percent 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.
Developments of PUREX
Another version of PUREX separates the minor actinides (americium, neptunium, curium) in a second aqueous stage; they are then directed to an accelerator-driven system cycling with pyroprocessing for transmutation (see later sections). The waste stream is then composed largely of fission products.
A modified version of the PUREX that does not involve the isolation of a plutonium stream is the UREX (uranium extraction) process. This process can be supplemented to recover iodine by volatilization and technetium by electrolysis. Research at the French Atomic Energy Commission (Commissariat à l'énergie atomique, CEA) has shown the potential for 95% and 90% recoveries of iodine and technetium respectively. The same research effort has demonstrated separation of caesium.
Another variation of PUREX is being developed by the US Department of Energy for the processing of civil wastes. In this system, only uranium is recovered (hence a UREX or UREX+ process) initially for recycling or for disposal as low-level waste; iodine and technetium may also be recovered at the head end. The residual is treated to recover plutonium (or plutonium plus neptunium) for recycling in conventional reactors, and the other actinides for transmutation in fast reactors. The high-level waste is then composed mostly of fission products. The central feature of this system is to increase proliferation resistance by keeping the plutonium with other transuranics - all of which are then destroyed by recycling in fast reactors.
Areva and France's CEA have developed three processes on the basis of extensive French experience with PUREX:
- The COEX process based on co-extraction and co-precipitation of uranium and plutonium (and possibly neptunium) together as well as a pure uranium stream (eliminating any separation of plutonium on its own). It is close to near-term industrial deployment, and allows high MOX performance for both light-water and fast reactors. COEX may have from 20 to 80% uranium in the product, the baseline is 50%.
- The DIAMEX-SANEX processes involving selective separation of long-lived radionuclides (with a focus on Am and Cm separation) from short-lived fission products. This can be implemented with COEX. U-Pu and minor actinides are recycled separately in Generation IV fast neutron reactors.
- The GANEX (grouped extraction of actinides) process co-precipitates some uranium with the plutonium (as with COEX), but then separates minor actinides and some lanthanides from the short-lived fission products. The uranium, plutonium and minor actinides together become fuel in Generation IV fast neutron reactors, the lanthanides become waste. It is to be demonstrated at La Hague from 2008 as part of a French-Japanese-US Global Actinide Cycle International Demonstration (GACID) with the product transmutation being initially in France's Phenix fast reactor and subsequently in Japan's Monju.
All three processes are to be assessed in 2012, so that two pilot plants can be built to demonstrate industrial scale potential:
- One - possibly based on COEX - to make the driver fuel for the Generation IV reactor planned to be built by CEA by 2020.
- One to produce fuel assemblies containing minor actinides for testing in Japan's Monju fast reactor and in France's Generation IV fast reactor.
In the longer term the goal is to have a technology validated for industrial deployment of Generation IV fast reactors about 2040, at which stage the present La Hague plant will be due for replacement.
Another alternative process being developed by Mitsubishi and Japanese R&D establishments is Super-DIREX (supercritical fluid direct extraction). This is designed to cope with uranium and MOX fuels from light water and fast reactors. The fuel fragments are dissolved in nitric acid with tributyl phosphate (TBP) and supercritical CO2, which results in uranium and plutonium complexing with TBP.
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 targets to be separated out of the wastes. Second, 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 term 'electrometallurgical' is also increasingly used to refer to this in the USA.
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 or for commercial purposes (eg americium for household smoke detectors). Otherwise, any transmutation effort is a random process with uncertain results. In particular, one does not want fertile uranium isotopes in a transmutation target with slow neutrons, or neutron capture will be the main action and will hence generate further radiotoxic transuranic isotopes.
Achieving effective full separation for any transmutation program is likely to mean pyroprocessing of residuals from the PUREX process or similar aqueous processes.
Electrolytic/ electrometallurgical processing techniques ('pyroprocessing') to separate nuclides from a radioactive waste stream have been under development in the US Department of Energy laboratories, notably Argonne, as well as by the Korea Atomic Energy Research Institute (KAERI) in conjunction with work on DUPIC (see section on DUPIC below).
So-called pyroprocessing involves several stages including: volatilization, liquid-liquid extraction using immiscible metal-metal phases or metal-salt phases, electrolytic separation in molten salt, and fractional crystallization. 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 and are envisaged for fuels from Generation IV reactors.
Electrometallurgical 'pyroprocessing' can readily be applied to high burn-up fuel and fuel which has had little cooling time, since the operating temperatures are already high. However, such processes are in the early stage of development compared with already-operational hydrometallurgical processes.
Separating (partitioning) the actinides contained in a fused salt bath involves electrodeposition on a cathode, so involves all the positive ions without the possibility of chemical separation of heavy elements such as in PUREX and its derivatives. This cathode product can then be used in a fast reactor.
So far, only one pyroprocessing technique has been licensed for use on a significant scale. This is the IFR (integral fast reactor) process developed by Argonne National Laboratory in the U.S., used for pyroprocessing the used fuel from EBR-II experimental fast reactor which ran from 1963-1994. This application is essentially a partitioning-conditioning process, because neither plutonium nor other transuranics are recovered for recycling. 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 KAERI advanced spent fuel conditioning process (ACP) involves separating uranium, transuranics including plutonium, and fission products including lanthanides. It utilises a high-temperature lithium-potassium cathode. Development of this process is at the heart of US-South Korean nuclear cooperation, and will be central to the renewal of the bilateral US-South Korean nuclear cooperation agreement in 2014, so is already receiving considerable attention in negotiations.
With US assistance through the International Nuclear Energy Research Initiative (I-NERI) program KAERI built the Advanced Spent Fuel Conditioning Process Facility (ACPF) at KAERI in 2005. KAERI hopes the project will be expanded to engineering scale by 2012, leading to the first stage of a Korea Advanced Pyroprocessing Facility (KAPF) starting in 2016 and becoming a commercial-scale demonstration plant in 2025.
South Korea has declined an approach from China to cooperate on electrolytic reprocessing, and it has been rebuffed by Japan's CRIEPI due to government policy.
The Russian Institute of Atomic Reactors (RIAR) at Dimitrovgrad has developed a pilot scale pyroprocessing demonstration facility for fast reactor fuel.
The objective of transumutation is to change (long-lived) actinides into fission products and long-lived fission products into significantly shorter-lived nuclides. The goal is to have wastes which become radiologically innocuous in only a few hundred years. The need for a waste repository is certainly not eliminated, but it can be smaller and simpler and the hazard posed by the disposed waste materials is greatly reduced.
Transmutation of one radionuclide into another is achieved by neutron bombardment in a nuclear reactor or accelerator-driven device. In the latter, 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 in a conventional nuclear power 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. For further information, see the article on Accelerator-driven nuclear energy.
Transmutation is mainly initiated by fast neutrons. Since these are more abundant in fast neutron reactors, such reactors are preferred for transmutation. 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. However, a 2001 BNFL-Cogema study found that full transmutation in a light water reactor would take at least several decades, and recent research has focused on use of fast reactors. 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.
Another approach to used nuclear fuel recycling which could be employed by some countries is DUPIC (direct use of used PWR fuel in CANDU reactors). CANDU reactors use as fuel natural uranium which has not undergone enrichment and so could theoretically operate fuelled by the uranium and plutonium that remains in used fuel from light water reactors.
With DUPIC, used fuel assemblies from light water reactors (LWRs) would be dismantled and refabricated into fuel assemblies the right shape for use in a CANDU reactor. This could be direct, involving only cutting the used LWR fuel rods to CANDU length (about 50 cm), resealing and reengineering into cylindrical bundles suitable for CANDU geometry.
Alternatively, a dry reprocessing technology has been developed which removes only the volatile fission products from the spent LWR fuel mix. After removal of the cladding, a thermal-mechanical process is used to reduce the used LWR fuel pellet to a powder. This could have more fresh natural uranium added, before being sintered and pressed into CANDU pellets.
The DUPIC technique has certain advantages:
- No materials are separated during the refabrication process. Uranium, plutonium, fission products and minor actinides are kept together in the fuel powder and bound together again in the DUPIC fuel bundles.
- A high net destruction rate can be achieved of actinides and plutonium.
- Up to 25% more energy can be realised compared to other PWR used fuel recycling techniques.
- And a DUPIC fuel cycle could reduce a country¹s need for used PWR fuel disposal by 70% while reducing fresh uranium requirements by 30%.
However, as noted above, used nuclear fuel exhibits residual radioactivity and generates heat. This means that the DUPIC manufacture process must be carried out remotely behind heavy shielding. While these restrictions make the diversion of fissile materials much more difficult and hence increase security, they also make the manufacture process more complex compared with that for the original PWR fuel, which is barely radioactive before use.
KAERI believes that although it is too early to commercialise the DUPIC fuel cycle, the key technologies are in place for a practical demonstration of the technique. Challenges which remain include the development of a technology to produce fuel pellets of the correct density, the development of remote fabrication equipment and the handling of the used PWR fuel. However, KAERI successfully manufactured DUPIC small fuel elements for irradiation tests inside the HANARO research reactor in April 2000 and fabricated full-size DUPIC elements in February 2001. AECL is also able to manufacture DUPIC fuel elements.
A further complication is the loading of highly radioactive DUPIC fuel into the CANDU reactor. Normal fuel handling systems are designed for the fuel to be hot and highly radioactive only after use, but it is thought that the used fuel path from the reactor to cooling pond could be reversed in order to load DUPIC fuel, and studies of South Korea's Wolsong CANDU units indicate that both the front- and rear-loading techniques could be used with some plant modification.
- WNA paper on Nuclear fuel reprocessing.
- Madic, C. 2000, Overview of the hydrometallurgical and pyrometallurgical processes Š. for partitioning high-level nuclear wastes, in Actinide 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.
- American Nuclear Society Nuclear News Sept 2005, statement Nov 2005.