Processing Used Nuclear Fuel

Over the last 50 years, the principal reason for reprocessing used fuel has been to recover unused uranium and plutonium 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 disposed of as high-level waste to about one-fifth. In addition, the level of radioactivity in the debris from reprocessing is much smaller and, after nearly 100 years, falls much more rapidly than in the used fuel itself.

In the last decade, interest has grown in recovering all long-lived actinides together (i.e., with plutonium) to recycle them in fast reactors. Hence, they end up as short-lived fission products. Two factors drive this policy: reducing the long-term radioactivity in high-level wastes and reducing the possibility of diverting plutonium 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, almost 90,000 tonnes (of 290,000 t discharged) of used fuel from commercial power reactors has been reprocessed. Annual reprocessing capacity is now some 4000 tonnes per year for standard oxide fuels, but not all are operational.

Between now and 2030, some 400,000 tonnes of used fuel will be generated worldwide, including 60,000 t in North America and 69,000 t in Europe.

World commercial reprocessing capacity

(tons per year)
LWR fuel France, La Hague
UK, Sellafield (THORP)
Russia, Ozersk (Mayak)
Japan (Rokkasho)
Total (approx)
Other nuclear fuels UK, Sellafield (Magnox)
Total (approx)
Total civil capacity


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 generally have less than 1% U-235 (typically about 0.5% U-235) and 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 formed mainly through the alpha decay of Pu-236, and its concentration of it peaks after about ten years of storage.

The U-236 isotope is a neutron absorber present in much more significant amounts, typically 0.4% to 0.6% – more with higher burn-up – which means that if reprocessed uranium is used for fresh fuel in a conventional reactor, it must be enriched significantly more (e.g., up to one-tenth more) 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 energy 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 fission products and transuranic traces. This will affect its suitability for recycling as blend material or via enrichment. Over 2002-06 USEC successfully cleaned up 7400 tonnes of technetium-contaminated uranium from the US Department of Energy.

Most of the separated uranium (RepU) remains in storage. However, its conversion and re-enrichment (in the UK, Russia, and the Netherlands) and its reuse in fresh fuel have been demonstrated. Some 16,000 tonnes of RepU from Magnox reactors in the UK have been used to make about 1650 tonnes of enriched AGR fuel. Over 8000 tonnes of RepU have been recycled into nuclear power plants in Belgium, France, Germany, and Switzerland. In Japan, the figure is around 335 tonnes in tests. In India, about 250 t of RepU has been recycled into PHWRs. Allowing for impurities affecting its treatment and use, RepU value has been assessed as nearly 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 the plutonium is due to the increasing proportion of non-fissile isotopes and minor actinides and the depletion of fissile plutonium isotopes. Whether this plutonium is separated independently or with other actinides is a significant policy issue relevant to reprocessing (see the section on Reprocessing policies below).

Most 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 page on 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 plutonium 70 15,000
Ex-military high-enriched uranium 230 70,000


A great deal of hydrometallurgical reprocessing has been going on since the 1940s, initially for military purposes, to recover plutonium for weapons (from low burn-up used fuel, which has been in a reactor for only a few months). In the UK, metal fuel elements from the Magnox generation gas-cooled commercial reactors have been reprocessed at Sellafield for about 50 years. The 1500 t/yr Magnox reprocessing plant undertaking 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, and the 900 t/yr Thermal Oxide Reprocessing Plant (THORP) at Sellafield was commissioned in 1994.

No civil reprocessing plants are operating in the USA, though three have been built. The first, a 300 t/yr plant at West Valley, New York, was used successfully from 1966-72. However, escalating regulation required uneconomic plant modifications 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. It was declared inoperable in 1974. The third was a 1500 t/yr plant at Barnwell, South Carolina, which was aborted due to a 1977 change in government policy that ruled out all US civilian reprocessing as one facet of US non-proliferation policy. The USA has over 250 plant years of reprocessing operational experience, the vast majority at government-operated defense plants since the 1940s.

In France, a 400 t/yr reprocessing plant operated for metal fuels from gas-cooled reactors at Marcoule until 1997. At La Hague, iron fuel reprocessing 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 for storing reprocessed uranium (RepU) for up to 250 years as a strategic reserve. Currently, reprocessing 1150 tonnes of EDF-used fuel per year produces 8.5 tonnes of plutonium (immediately recycled as 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. Still, it is currently uneconomic due to conversion costing three times as much as fresh uranium and enrichment needing to be separate because of U-232 and U-236 impurities. The presence of the gamma-emitting U-232 requires shielding and should be handled in dedicated facilities. The presence of the neutron-absorbing U-236 isotope means that a higher level of enrichment is needed compared with fresh uranium.

The plutonium is immediately recycled via the dedicated Melox mixed oxide (MOX) fuel fabrication plant. The reprocessing output in France is coordinated with MOX plant input to avoid building up plutonium stocks. Suppose plutonium is stored for some years; the level of americium-241 is. In that case, the isotope used in household smoke detectors will accumulate, making it challenging to handle through a MOX plant due to the elevated gamma radioactivity levels.

India has a 100 t/yr oxide fuel plant operating at Tarapur with others at Kalpakkam and Trombay, and Japan is starting up a central (800 t/yr) plant at Rokkasho while having had most of its used fuel reprocessed in Europe meanwhile. It has a small (90 t/yr) reprocessing plant 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 together
  • Separate U, Pu+actinides, and certain fission products

In today’s reactors, reprocessed uranium (RepU) must be enriched, whereas plutonium goes straight to mixed oxide (MOX) fuel fabrication. This situation has two perceived problems: the separated plutonium is a potential 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.

As minor actinides are not destroyed, recycling through light water reactors delivers only part of the potential waste management benefit. For the future, the focus is on removing the actinides from the final waste and burning them with recycled uranium and plutonium in fast neutron reactors. (The longer-lived fission products may also be separated from the trash and transmuted in some other way.) Hence the combination of reprocessing followed by recycling in today’s reactors should be seen as an interim phase of nuclear power development, pending the widespread use of fast neutron reactors.

All but one of the six Generation IV reactors being developed have closed fuel cycles that recycle all the actinides. Although US policy has been to avoid reprocessing, the US budget process 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 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.”

In February 2006, the US government announced the Global Nuclear Energy Partnership (GNEP). It would “work with other nations possessing advanced nuclear technologies to develop new proliferation-resistant recycling technologies to produce more energy, reduce waste and minimize proliferation concerns.” GNEP’s goals included reducing US dependence on imported fossil fuels and building a new generation of nuclear power plants in the USA. Two significant new elements in the strategy were new reprocessing technologies at advanced recycling centers, which separate all transuranic elements together (and not plutonium on its own) ­starting with the UREX+ process (see the section on Developments of PUREX below), and ‘advanced burner reactors’ to consume the result of this while generating power.

GE Hitachi Nuclear Energy (GEH) is developing this concept by combining electrometallurgical separation (see the 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, fission products, waste, and finally, actinides, including plutonium.

In mid-2006, a report by the Boston Consulting Group for Areva based on proprietary Areva information showed that recycling used fuel in the USA using the COEX aqueous process (see Developments of PUREX below) would be economically competitive with direct disposal of used fuel. A $12 billion, 2500 t/yr plant was considered, with a total capital expenditure of $16 billion for all related aspects. This would have the benefit of significantly reducing demand for space at the planned Yucca Mountain repository.

Boston Consulting Group gave four reasons for reconsidering the US used fuel strategy, which has been applied since 1977:

  • Cost estimates for direct disposal at Yucca Mountain had risen sharply, 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 now a lot of experience with civil reprocessing.

Soon after, the US Department of Energy said it might start the GNEP program using reprocessing technologies that “do not require further development of any substantial nature,” such as COEX, while others were further developed. It also flagged detailed siting studies on the feasibility of this accelerated “development and deployment of advanced recycling technologies by proceeding with commercial-scale demonstration facilities.”


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 – which may be used for producing Pu-238 for thermo-electric generators for spacecraft – 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 before re-enrichment and the plutonium straight to MOX fuel fabrication.

Alternatively, some small amount of recovered uranium can be left with the plutonium and sent to the MOX plant so the plutonium is never separated. 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 is removed from high-level waste, containing about 3% of the used fuel in 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 incorporating the dry material into borosilicate glass, then stored pending disposal. In principle, any compact, stable, insoluble solid is satisfactory for removal.


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 the fission products 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% of iodine and technetium recoveries, respectively. The same research effort has demonstrated the separation of cesium.

The US Department of Energy is developing the UREX+ processes under the Global Nuclear Energy Partnership (GNEP) program (see page on Global Nuclear Energy Partnership). Only uranium is initially recovered for recycling, and the residual is treated to recover plutonium with other transuranic. The fission products then comprise most of the high-level waste. The central feature of this system is to increase proliferation resistance by keeping the plutonium with other transuranic – all of which are then destroyed by recycling in fast reactors.

Several variations of UREX+ have been developed, with the differences being in how the plutonium is combined with various minor actinides and lanthanide and non-lanthanide fission products are combined or separated. UREX+1a combines plutonium with three minor actinides, but this gives rise to problems in fuel fabrication due to americium being volatile and curium being a neutron emitter. Therefore, remote fuel fabrication facilities would be required, leading to high fuel fabrication costs and significant technological development. An alternative process, UREX+3, is therefore being considered. This leaves only neptunium with the plutonium, and the result is closer to a conventional MOX fuel. However, it is less proliferation-resistant than UREX+1a.

Energy Solutions holds the rights to PUREX in the USA and has developed NUEX, which separates uranium and all transuranic (including plutonium) with fission products separately. NUEX is similar to UREX+1a but has more flexibility in the separations process.

Areva and CEA have developed three processes based on extensive French experience with PUREX:

  • The COEX process is based on the co-extraction and co-precipitation of uranium and plutonium (and usually 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 allowing high MOX performance for light-water and fast reactors. COEX may have from 20 to 80% uranium in the product; the baseline is 50%.
  • The DIAMEX-SANEX processes involve elective separation of long-lived radionuclides (focusing on Am and Cm separation) from short-lived fission products. This can be implemented with COEX following a break of U-Pu-Np. 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). Still, it separates minor actinides and lanthanides from the short-lived fission products. The uranium, plutonium, and minor actinides fuel Generation IV fast neutron reactors; the lanthanides become waste. It has been demonstrated at ATALANTE and La Hague since 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 (see Transmutation section below) and subsequently Japan’s Monju.

Initial work is at ATALANTE at Marcoule, which started operation in 1992 to consolidate reprocessing and recycling research from three other sites but is still under development. By 2012, it is expected to have demonstrated GANEX and fabrication of oxide fuel pins combining U, Pu, Am, Np & Cm. Then work will proceed at La Hague on partitioning and fabricating minor actinide-bearing fuels without the curium. From 2020 these will be irradiated in the Monju fast reactor in Japan.

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 – is to make the driver fuel for the Generation IV reactor planned to be built by CEA by 2020.
  • One is to produce fuel assemblies containing minor actinides for testing in Japan’s Monju fast reactor and France’s Generation IV fast reactor.

The long-term goal is to have a technology validated for industrial deployment of Generation IV fast reactors by about 2040. The present La Hague plant will be due for replacement at this stage.

Another alternative reprocessing technology 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, plutonium, and minor actinides complexing with TBP.

A new reprocessing technology is part of the reduced-moderation water reactor (RMWR) concept. This is the fluoride volatility process, developed in the 1980s, coupled with solvent extraction for plutonium to give Hitachi’s Fluorex process. In this, 90-92% of the uranium in the used fuel is volatilized as UF6, then purified for enrichment or storage. The residual is put through a Purex circuit separating fission products and minor actinides, leaving the unseparated U-Pu mix (about 4:1) into MOX fuel.


Several factors give rise to a more sophisticated view of reprocessing today, and 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 the prime objective is to separate long-lived radionuclides from short-lived ones. Secondly, new fuel cycles, such as those for fast neutron reactors (including a lead-cooled one), fused salt reactors, and the possible advent of accelerator-driven systems, require a new approach to reprocessing. Here the focus is on electrolytic processes (‘preprocessing) in a molten salt bath. The term ‘electrometallurgical’ is also increasingly used to refer to this in the USA.

The primary 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-129, technetium-99, caesium-135, and strontium-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 separate different isotopes of any particular element.

Efficient separation methods are needed to achieve low residuals of long-lived radionuclides in conditioned wastes and high purities of individual separated ones for use in transmutation targets or for commercial purposes (e.g., americium for household smoke detectors). If transmutation targets are not of high purity, then the transmutation results will be uncertain. In particular fertile uranium isotopes (e.g., U-238) in a transmutation target with slow neutrons will generate different radiotoxic transuranic isotopes through neutron capture.

Achieving effective complete separation for any transmutation program will likely mean electrolytic processing 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 US 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.


Electrolytic/electrometallurgical processing techniques (‘preprocessing) 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 the section on Recycled LWR uranium and used fuel in PHWRs below). Their significant development has possibly been in Russia, where they will be the mainstay of fully closing the fuel cycle by about 2020. There has been a particular emphasis on fast reactor fuels.

So-called pyro processing involves several stages, including volatilization; liquid-liquid extraction using immiscible metal-metal or metal-salt phases; electrolytic separation in molten salt; and fractional crystallization. They are generally based on using fused salts such as chlorides or fluorides (e.g., LiCl+KCl or LiF+CaF2) or combined metals such as cadmium, bismuth, or aluminum. They are most readily applied to metal rather than oxide fuels and are envisaged for Generation IV reactors’ fuels.

Electrometallurgical ‘pyroprocessing’ can readily be applied to high burn-up fuel and fuel with little cooling time since the operating temperatures are already high. However, such processes are at an early stage of development compared with hydrometallurgical processes already operational.

Separating (partitioning) the actinides in a fused salt bath is electrodeposition on a cathode. It involves all the positive ions without the possibility of chemical separation of heavy elements or nuts, such as in PUREX and its derivatives. This cathode product can then be used in a fast reactor.

So far, only one electrometallurgical technique has been licensed for use on a significant scale. This is the IFR (integral fast reactor) electrolytic process developed by Argonne National Laboratory in the USA and used for pyroprocessing the used fuel from the EBR-II experimental fast reactor from 1963-1994. This application is a partitioning-conditioning process because neither plutonium nor other transuranic are recovered for recycling. The method facilitates the disposal of fuel that could not otherwise be sent directly to a geologic repository. The used uranium metal fuel is dissolved in a 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 salt and are consolidated to form a stable metallic waste. The transuranic 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 future use.

The PYRO-A process, developed at Argonne to follow the UREX process, is a pyrochemical process for separating transuranic elements and fission products contained in the oxide powder resulting from denitration of the UREX raffinate. The residual raffinate acid solution nitrates 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 effect is electrorefining 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 products in 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.

The PYRO-B process has been developed to process and recycle used fuel from a transmuter (fast) reactor. A typical transmuter fuel is uranium-free and contains recovered transuranic in an inert matrix such as metallic zirconium. In the PYRO-B processing of such power, an electrorefining step separates the residual transuranic elements from the fission products and recycles the transuranic 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.

The KAERI advanced spent fuel conditioning process (ACP) involves separating uranium, transuranic, including plutonium, and fission products, including lanthanides. It utilizes a high-temperature lithium-potassium chloride bath from which uranium is recovered electrolytically to concentrate the actinides, which are then removed together (with some remaining fission products). The latter effect is then fabricated into fast reactor fuel without further treatment. The process is intrinsically proliferation-resistant because it is so hot radiologically, and the curium provides a high level of spontaneous neutrons. It recycles about 95% of the used fuel. The development of this process is at the heart of US-South Korean nuclear cooperation. It is central to renewing the bilateral US-South Korean nuclear cooperation agreement in March 2014, so it 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. KAERI hopes the project will be expanded to an 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 Japan’s CRIEPI has rebuffed it due to government policy.

Russian pyro processing consists of three main stages: dissolution of the used nuclear fuel in molten salts, precipitation of plutonium dioxide or electrolytic deposition of uranium and plutonium dioxides from the melt, then processing of the material deposited on the cathode or precipitated at the bottom of the melt for granulated fuel production. The process recovers the cathode deposits without changing their chemical composition or redistributing the plutonium. All products are reprocessed to have complete recycling of plutonium, neptunium, americium, and curium, as well as uranium. Combined with vibropacking in fuel fabrication, this process will produce fuel for the BN-800 fast reactor. The technologies complement one another well and involve high levels of radioactivity throughout, making them self-protecting against diversion or misuse.

A pilot-scale pyroprocessing demonstration facility for fast reactor fuel has been developed by the Russian Institute of Atomic Reactors (RIAR) at Dimitrovgrad.

GE Hitachi is designing an Advanced Recycling Centre (ARC) which integrates electrometallurgical processing with its PRISM fast reactors. The main feed is used fuel from light water reactors, and the three products are fission products, uranium, and transuranic (Np, Pu, Am, Cm), which become fuel for the fast reactors (with some of the uranium). The uranium can be re-enriched or used as fuel for Candu reactors. As the cladding reaches its exposure limits, used PRISM fuel is recycled after removing fission products. A complete commercial-scale ARC would comprise an electrometallurgical plant and three power blocks of 622 MWe each (six 311 MWe reactor modules), but a “full-scale building block” of ARC is a 50 t/yr electrometallurgical plant coupled to one 311 MWe reactor module, with a breeding ratio of 0.8.


Transmutation aims to change (long-lived) actinides into fission products and long-lived fission products into significantly shorter-lived nuclides. The goal is to have wastes that become radiologically innocuous in only a few hundred years. The need for a waste repository is not eliminated, but it can be smaller and more straightforward, and the hazard posed by the disposed waste materials is significantly reduced.

One radionuclide is transformed into another 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 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.

Fast neutrons mainly initiate transmutation. 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 complete transmutation in a light water reactor would take at least several decades, and recent research has focused on using 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.

One of the main functions of France’s Phenix fast neutron reactor in its last two years of operation was test burning fuel assemblies containing high concentrations of minor actinides. From mid-2007, it irradiated four fuel pins containing actinides from the US Department of Energy, two from the CEA, and two from the European Commission’s Institute for Transuranics.


Another approach to used nuclear fuel recycling is directing recycled uranium (referred to as RepU, reprocessed uranium) or used light water reactor (LWR) fuel into pressurized heavy water reactors (PHWRs). This may be directly using RepU, blending it with depleted uranium to give natural uranium equivalent, or using used PWR fuel in Candu reactors (DUPIC).

PHWRs (such as Candu reactors) typically use natural uranium as fuel, which has not undergone enrichment. They could operate fuelled by the uranium (and plutonium) that remains in used power from LWRs.

In unit 1 of the Qinshan Phase III plant in China, a demonstration commenced in March 2010 using fuel bundles with RepU from PWRs blended with depleted uranium to give natural uranium equivalent. Subject to supply from reprocessing plants, an entire core of natural uranium equivalent is envisaged.

AECL says using the RepU directly in Candus without blending it down is also possible.

With DUPIC, used fuel assemblies from LWRs would be dismantled and refabricated into fuel assemblies of 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 them into cylindrical bundles suitable for Candu geometry.

A dry reprocessing technology has been developed as an alternative method for the basic DUPIC process, which removes only the volatile fission products from the used LWR fuel mix. After the firing of the cladding, a thermal-mechanical process is used to reduce the LWR fuel pellet to 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 for actinides and plutonium.
  • Compared to other PWR-fuel recycling techniques, up to 25% more energy can be realized.
  • And a DUPIC fuel cycle could reduce a country’s need for used PWR fuel disposal by 70% while reducing fresh uranium requirements by 30%.

However, as noted above, used nuclear fuel is highly radioactive and generates heat. This high activity means the DUPIC manufacturing process must be carried out remotely behind heavy shielding. While these restrictions make the diversion of fissile materials much more complex and increase security, they also make the manufacturing process more complex than the original PWR fuel, which is barely radioactive before use.

A further complication is loading highly radioactive DUPIC fuel into the Candu reactor. Standard fuel handling systems are designed for the fuel to be hot and highly radioactive only after use. Still, it is thought that the used fuel path from the reactor to the cooling pond could be reversed 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.

Canada, which developed the Candu reactor, and South Korea, which hosts four Candu units as well as many PWRs, have initiated a bilateral joint research program to establish DUPIC, and the Korean Atomic Energy Research Institute (KAERI) has been implementing a comprehensive development program since 1992 to demonstrate the DUPIC fuel cycle concept.

KAERI believes that although it is too early to commercialize the DUPIC fuel cycle, the key technologies are in place to demonstrate the technique practically. Challenges that remain include the development of a technology to produce fuel pellets of the correct density, developing remote fabrication equipment, and handling 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 parts in February 2001. AECL is also able to manufacture DUPIC fuel elements.

Research is also underway on the reactor physics of DUPIC fuel and its impacts on safety systems.