Information Papers

Processing of Used Nuclear Fuel for Recycle 

(April 2008)

The availability of recyclable fissile and fertile materials able to provide fresh fuel for existing and future nuclear power plants is a key, nearly unique, characteristic of nuclear energy. In many countries government policies have not yet addressed the various aspects of this feature.

Over the last fifty 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 to be disposed of as high-level waste to about one fifth.  In addition, the level of radioactivity in such 'light' waste is much smaller and after about 100 years falls much more rapidly than in used fuel itself.  Reprocessing has been the government policy in many European countries, Russia and Japan.

In the last decade interest has grown in separating ('partitioning') individual radionuclides or groups of them both to reduce long-term radioactivity in residual wastes and to be able to transmute separated long-lived radionuclides into shorter-lived ones, mostly by fission. Starting in 2005 this interest in more fully closing the fuel cycle has grown and became more public, driven by concerns about long-term resource utilisation and proliferation resistance.

Reprocessing used fuel* to recover uranium (U, as RepU) and plutonium (Pu) avoids the wastage of a valuable resource. Most of it - about 96% - is uranium at less than 1% U-235 (often 0.4 - 0.8%), and up to 1% is plutonium. Both can be recycled as fresh fuel, saving some 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.

* Used fuel from light water reactors (at normal US burn-up) contains approximately:
95.6% uranium (U-232 0.1-0.3%, U-234 0.1-0.3%, U-235 0.5-1.0%, U-236 ).4-0.7%, balance: U-238)
2.9% stable fission products
0.9% plutonium
0.3% cesium & strontium (fission products)
0.1% iodine and technetium (fission products)
0.1% other long-lived fission products
0.1% minor actinides (americium, curium, neptunium)

Reprocessing also avoids leaving the plutonium in the used fuel, where in a century or two the built-in radiological protection will have diminished, allowing it to be recovered for illicit use (though it is unsuitable for weapons due to the non-fissile isotopes present).

 

World Commercial Reprocessing Capacity

(tonnes per year)
LWR fuel: France, La Hague
1700
UK, Sellafield (THORP)
900
Russia, Ozersk (Mayak)
400
Japan (Rokkasho)
800
total approx
3800
Other nuclear fuels: UK, Sellafield
1500
India
275
total approx
1750
Total civil capacity  
5550
Sources: OECD/NEA 2006 Nuclear Energy Data, Nuclear Eng. International handbook 2007.
 

So far, almost 90,000 tonnes (of 290,000 t discharged) of used fuel from commercial power reactors has been reprocessed for U & Pu recovery. Annual reprocessing capacity is now some 4000 tonnes per year for normal oxide fuels, but not all of it is operational.

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.  Higher assay material (c 0.9% U-235) from THORP reprocessing will be used in the next Sizewell B reload.  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.

* since Magnox fuel was not enriched in the first place, this is actually known as Magnox depleted uranium (MDU).  It assayed about 0.4% U-235 and was converted to UF6, enriched to 0.7% at BNFL's Capenhurst diffusion plant and then to 2.6% to 3.4% at Urenco's centrifuge plant.  Until the mid 1990s some 60% of all AGR fuel was made from MDU and it amounted to about 1650 tonnes of LEU.  Recycling of MDU was discontinued in 1996 due to economic factors.

Most of the separated plutonium is used almost immediately in mixed oxide (MOX) fuel. 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 isotopes.

Reprocessing policies

Conceptually reprocessing can take several courses, separating certain elements from the remainder, which becomes high-level waste:

* Minor actinides are americium and curium (95 & 96 in periodic table), sometimes also neptunium (93). (The major actinides are plutonium - #94 and uranium - #92)

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.

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.

All but one of the six Generation IV reactors being developed have closed fuel cycles which recycle all the actinides. Although US policy has been to avoid reprocessing, the US budget process for 2006 includes $50 million to develop a plan for "integrated spent fuel recycling facilities", and a program to achieve this with fast reactors will apparently be a major US budget request the following year.

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."

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* 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, though 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 10 years of storage. 

* Typical value today about 0.3% U-235.

The U-236 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 (eg 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 fuel is best used for blending or MOX 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 of technetium-contaminated uranium from the US Department of Energy. 

About 2000 to 2500 tonnes per year of RepU is currently used for fuel.  The remainder is stored.  Reprocessing of 850 tonnes of French used LWR fuel per year (about 15 years after discharge) yields 810 tonnes of reprocessed uranium (RepU).  Of this about two thirds is converted into stable oxide form for storage. One third of the RepU is re-enriched and EdF has demonstrated its use in 900 MWe power reactors. 

The separated plutonium from reprocessing will have an isotopic concentration determined by the fuel burn-up level.  At anything over about 20 GWday/t burn-up this will be "reactor grade" and significantly different from weapons grade material.*

* Some figures for Oskarshamn-3: with 30 GWd/t burn-up: 69% Pu is fissile, 40 GWd/t 61% fissile, 50 GWd/t: 55% fissile and 60 GWD/t: 50% fissile.

Reprocessing the 850 tonnes of French used fuel per year produces 8.5 tonnes of plutonium, which is immediately recycled as 100 tonnes of MOX.  World MOX production capacity in 2008 is a little over 200 tonnes per year, mostly in France, and this utilises about one third of the annual amount separated in reprocessing plants.

In 2007 EdF said that the plutonium stored at Areva's La Hague from reprocessing could provide the start-up fuel for seven Generation IV fast reactors, with 15 tonnes for each.

Inventory of Separated Recyclable Materials

 

  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

 

Source: NEA 2007. 

Reprocessing today - PUREX

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, most at the reactor site or otherwise at a central storage facility 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).  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 MOX fuel fabrication. 

*  Plutonium URanium EXtraction

** may be used for producing Pu-238 for thermo-electric generators for spacecraft.

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 COEX ** process, developed in France as a "Generation III" process, but not yet in use (see next section). 

 ** CO-EXtraction of actinides

At Japan's new Rokkasho plant a modified PUREX process is used to achieve a similar result by recombining some uranium before co-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 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.

After reprocessing, the recovered uranium may be handled in a fuel fabrication plant (after re-enrichment)*, but the plutonium must be recycled via a dedicated mixed oxide (MOX) fuel fabrication plant. In France the reprocessing output is co-ordinated with MOX plant input, to avoid building up stocks of plutonium.**

* Recycled uranium must be handled in dedicated facilities because gamma-emitting uranium isotopes mean it requires shielding and neutron-absorbing isotopes mean that a higher level of enrichment is required compared with fresh uranium.

** 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.

Reprocessing of 850 tonnes of EdF used fuel per year (about 15 years after discharge) produces 8.5 tonnes of plutonium (immediately recycled as 100 tonnes of MOX) and 810 tonnes of reprocessed uranium (RepU). Of this about two thirds is converted into stable oxide form for storage. One third of the RepU is re-enriched at Pierrelatte and EdF has demonstrated its use of in 900 MWe power reactors.

Developments of PUREX

Another version of PUREX has the minor actinides (americium, neptunium, curium) 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 and its derivatives may also be supplemented to recover fission products iodine by volatilisation and technetium by electrolysis.  The US UREX+ process allows recovery of iodine and technetium at the head end.  French CEA research has shown 95% and 90% recoveries respectively.  The same research effort has demonstrated separation of caesium.

Energy Solutions Inc holds the rights to PUREX in the USA and has developed NUEX, which separates uranium and then all transuranics (including plutonium) together, with fission products separately.

Other variations of PUREX are being developed by the US Department of Energy for civil wastes.  In these, only uranium is recovered initially for recycle (hence UREX+ processes).  The residual is treated to recover plutonium with other transuranics for recycling in fast reactors.  The fission products then comprise most of the high-level waste.  The central feature of this system is to keep the plutonium with other transuranics which are destroyed by burning in a fast neutron reactor. 

A version of UREX+ was demonstrated with used fuel in 2005 at the Argonne National Laboratory, including separate recovery of technetium.  As of mid 2007 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 a neutron emitter.  UREX+3 leaves only neptunium with the plutonium and the result is closer to a conventional MOX fuel and hence more easily qualified by NRC.

Areva and CEA have developed three processes on the basis of extensive French experience with PUREX:

All three processes are to be assessed in 2012, so that two pilot plants can be built to demonstrate industrial scale potential:

History

 A great deal of reprocessing has been going on since the 1940s, originally 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 mixed oxide (MOX) fuel plant in 2001.

* from low burn-up used fuel, which has been in a reactor for only a very few months.

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. 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 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 a 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, giving overall capacity of 1700 t/yr.  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 meanwhile. It had a small (90 t/yr) plant operating at Tokai Mura. 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).

US prospects

As noted above, the USA has considerable reprocessing experience but application of this to civil used fuel has been frustrated by political sensitivities motivated by proliferation concerns based on an understanding that reactor-grade plutonium is usable for weapons. Civil reprocessing was stopped in 1977.

In February 2006 the US government has announced a Global Nuclear Energy Partnership (GNEP) through which it "will work with other nations possessing advanced nuclear technologies to develop new proliferation-resistant recycling technologies in order to produce more energy, reduce waste and minimise proliferation concerns. GNEP goals include 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 are new reprocessing technologies which separate all transuranic elements together (and not Pu on its own) ­ starting with the UREX+ process, and Advanced Burner (fast) Reactors to consume the result of this while generating power.

In mid 2006 a report by the Boston Consulting Group for Areva and based on proprietary Areva information showed that recycling used fuel in the USA using the COEX aqueous process would 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 Yucca mountain repository.

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

Soon after this the US Department of Energy said that 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 on this accelerated "development and deployment of advanced recycling technologies by proceeding with commercial-scale demonstration facilities."

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 with slow neutrons, or neutron capture will be the main action and hence it will generate further 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 techniques ('pyroprocessing') to separate nuclides from a radioactive waste stream involve several stages: volatilisation, liquid-liquid extraction using immiscible metal-metal phases or metal-salt phases, electrorefining in molten salt, fractional crystallisation, etc. They are generally based on the use of either fused (low-melting point) salts such as chlorides or fluorides (eg LiCl+KCl or LiF+CaF2) or fused metals such as cadmium, bismuth or aluminium. They are most readily applied to metal rather than oxide fuels, and are envisaged for fuels from generation IV reactors.

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.

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 USA and 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 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. 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.

The PYRO-B process, has been developed for the processing and recycle of fuel from a transmuter (fast) 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.

The Russian Institute of Atomic Reactors (RIAR) at Dimitrovgrad has developed at pilot scale pyroprocessing for fast reactor fuel.

Transmutation

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 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 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.

Transmutation is mainly by fast neutrons. Since these are more abundant in a fast neutron reactor, such reactors are preferred for transmutation - and of course they can be fuelled with the recovered plutonium. (In a fast reactor americium and neptunium are also fissionable, so can contribute to the energy.) A 2001 BNFL-Cogema study said that full transmutation in a light water reactor would take at least several decades. Unfortunately prime contributors to the long-term radiological load in a repository are those which are most difficult to transmute.

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 a few hundred years, it is less radiotoxic than the uranium ore originally used to produce the fuel. 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.

DUPIC

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 (CANadian Deuterium Uranium) 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.

Under DUPIC, used LWR fuel assemblies would be dismantled and refabricated into fuel assemblies the right shape for use in a CANDU reactor. Some schemes would see the PWR fuel powderised without dissolution and mixed, possibly with the addition of more fresh natural uranium, before being sintered and pressed into CANDU pellets.

The DUPIC technique has certain advantages:

However, as noted above, used nuclear fuel is highly radioactive and generates heat. This high activity 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.

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 develop 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.

Challenges which remain include the development of a technology to produce fuel pellets of the correct high 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. Atomic Energy of Canada are also able to manufacture DUPIC fuel elements.

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

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 Korea¹s Wolsong CANDU units indicate that both the front- and rear-loading techniques could be used with some plant modification.

KAERI believe 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.

Main sources:
Madic, C. 2000, Overview of the hydrometallurgical and pyrometallurgical processes 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.
Park, J.J. et al, 2000,Technology and implementation of the DUPIC concept for spent nuclear fuel in the ROK
Yang, S.Y et al, 2006, The status and prospect of DUPIC fuel technology, Nuclear Engineering and Technology, Vol.38, 4.
McFarlane, H.F. 2004, Nuclear Fuel Reprocessing, in Encyclopedia of Energy, Elsevier.
Wood, Janet 2006, Should USA Reprocess? Nuclear Engineering Int'l Sept 2006.
OECD/NEA 2007, Management of Recyclable Fissile and Fertile Materials, NEA # 6107.

UK Nuclear Decommissioning Authority, 2007, Uranium and Plutonium: Macro-economic Study.

IAEA 2007, Management of Reprocessed Uranium - current status and future prospects, Tecdoc 1529.