continued
Reprocessing policies
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 actinidese .
* Separate U, Pu+Np, Am+Cm.
* Separate U+Pu all together.
* Separate U, Pu+actinides, certain fission products.
In today's reactors, reprocessed uranium (RepU) needs to 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 there is no destruction of minor actinides, 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 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.) Hence the combination of reprocessing followed by recycling in today's reactors should be seen as an interim phase of nuclear power development, pending widespread use of fast neutron reactors.
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 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 statement4 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) through which it would "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 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 centres, which separate all transuranic elements together (and not plutonium on its own) *starting with the UREX+ process (see 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 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 report5 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 (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 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 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 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 of this accelerated "development and deployment of advanced recycling technologies by proceeding with commercial-scale demonstration facilities."
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 – 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 prior to re-enrichment and the plutonium straight to MOX 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 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.
Developments of PUREX
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 volatilisation, 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.
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). In these, only uranium is recovered initially for recycle and the residual is treated to recover plutonium with other transuranics. 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 transuranics - 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 a neutron emitter. Remote fuel fabrication facilities would therefore be required, leading to high fuel fabrication costs and requiring 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 then all transuranics (including plutonium) together, 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 on the basis of extensive French experience with PUREX:
* The COEX process based on 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, 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, following separation 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), 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 being demonstrated at ATALANTE and 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 (see Transmutation section below) and subsequently in Japan's Monju.
Initial work is at ATALANTEf 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 fabrication of minor actinide-bearing fuels without the curium. From 2020 these will be irradiated in the Monju fast reactor, 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 - 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 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.
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 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) and fused salt reactors, and the possible advent of accelerator-driven systems, require a new approach to reprocessing. Here the focus is on electrolytic processes ('pyroprocessing') in a molten salt bath. 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-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 transumation targets are not of high purity then the results of transmutation will be uncertain. In particular fertile uranium isotopes (e.g. U-238) in a transmutation target with slow neutrons will generate further radiotoxic transuranic isotopes through neutron capture.
Achieving effective full separation for any transmutation program is likely to 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.
Electrometallurgical 'pyroprocessing'
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: volatilisation; liquid-liquid extraction using immiscible metal-metal phases or metal-salt phases; electrolytic separation in molten salt; and fractional crystallisation. They are generally based on the use of either fused 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 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 is by 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 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 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 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 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 (eg 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 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 for the processing and recycle of used 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 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 chloride bath from which uranium is recovered electrolytically to concentrate the actinides, which are then removed together (with some remaining fission products). The latter product 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. Development of this process is at the heart of US-South Korean nuclear cooperation, and is central to the renewal of the bilateral US-South Korean nuclear cooperation agreement in March 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. 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.
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 transuranics (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 removal of fission products. A full 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 breeding ratio of 0.8.
Transmutation
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 spallationg. 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 page 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.
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.
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 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. In March 2010, this technique began a trial period after fuel bundles containing recovered uranium from used fuel were inserted into China's Qinshan Phase III unit 1.6
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 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 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.
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.
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 Korea's Wolsong CANDU units indicate that both the front- and rear-loading techniques could be used with some plant modification.
k got from above 2 reports.....
