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(updated May 2010)
Uranium found in nature consists largely of two isotopes, U-235 and U-238. The production of energy in nuclear reactors is from the 'fission' or splitting of the U-235 atoms, a process which releases energy in the form of heat. U-235 is the main fissile isotope of uranium.
Natural uranium contains 0.7% of the U-235 isotope. The remaining 99.3% is mostly the U-238 isotope which does not contribute directly to the fission process (though it does so indirectly by the formation of fissile isotopes of plutonium).
Uranium-235 and U-238 are chemically identical, but differ in their physical properties, particularly their mass. The nucleus of the U-235 atom contains 92 protons and 143 neutrons, giving an atomic mass of 235 units. The U-238 nucleus also has 92 protons but has 146 neutrons - three more than U-235, and therefore has a mass of 238 units.
The difference in mass between U-235 and U-238 allows the isotopes to be separated and makes it possible to increase or "enrich" the percentage of U-235. All present enrichment processes, directly or indirectly, make use of this small mass difference.
Some reactors, for example the Canadian-designed Candu and the British Magnox reactors, use natural uranium as their fuel. Most present day reactors (Light Water Reactors or LWRs) use enriched uranium where the proportion of the U-235 isotope has been increased from 0.7% to about 3% or up to 5%. (For comparison, uranium used for nuclear weapons would have to be enriched in plants specially designed to produce at least 90% U-235.)
International Enrichment Centres, Multilateral approaches
Following proposals from the International Atomic Energy Agency (IAEA) and Russia, and in connection with the US-led Global Nuclear Energy Partnership (GNEP), there are moves to establish international uranium enrichment centres. These are one kind of multilateral nuclear approaches (MNA) called for by IAEA. Part of the motivation for international centres is to bring all new enrichment capacity, and perhaps eventually all enrichment, under international control as a non-proliferation measure. Precisely what "control" means remains to be defined, and will not be uniform in all situations. But having ownership and operation shared among a number of countries at least means that there is a level of international scrutiny which is unlikely in a strictly government-owned and -operated national facility.
The first of these international centres is the International Uranium Enrichment Centre (IUEC) at Angarsk in Siberia, with Kazakh, Armenian and Ukrainian equity so far. The centre is to provide assured supplies of low-enriched uranium for power reactors to new nuclear power states and those with small nuclear programs, giving them equity in the project, but without allowing them access to the enrichment technology. Russia will maintain majority ownership, and in February 2007 the IUEC was entered into the list of Russian nuclear facilities eligible for implementation of IAEA safeguards. The USA has expressed support for the IUEC at Angarsk. IUEC will sell both enrichment services (SWU) and enriched uranium product.
In some respects this is very similar to what pertains now with the Eurodif set-up, where a single large enrichment plant in France with five owners (France - 60%, Italy, Spain, Belgium and Iran) is operated under IAEA safeguards by the host country without giving participants any access to the technology - simply some entitlement to share of the product, and even that is constrained in the case of Iran. The French Atomic Energy Commission proposed that the new Georges Besse II plant which replaces Eurodif should be open to international partnerships on a similar basis, and minor shares in the Areva subsidiary operating company Societe d'Enrichissement du Tricastin (SET) have so far been sold to GDF Suez, a Japanese partnership, and Korea Hydro and Nuclear Power (KHNP) - total 10%.
The three-nation Urenco set-up is also similar though with more plants in different countries - UK, Netherlands and Germany, and here the technology is not available to host countries or accessible to other equity holders. Like Russia with IUEC, Urenco (owned by the UK and Netherlands host governments plus E.On and RWE in Germany) has made it plain that if its technology is used in international centres it will not be accessible. Its new plant is in the USA, where the host government has regulatory but not management control.
A new Areva plant in the USA has no ownership diversity beyond that of Areva itself, so will be essentially a French plant on US territory. The only other major enrichment plant in the Western world is USEC's very old one, in the USA.
The Global Laser Enrichment project which may proceed to build a commercial plant in the USA has shareholding from companies based in three countries: USA (51%), Canada (24%) and Japan (25%).
CONVERSION
Uranium leaves the mine as the concentrate of a stable oxide known as U3O8 or as a peroxide,. It still contains some impurities and prior to enrichment has to be further refined before being converted to uranium hexafluoride (UF6), commonly referred to as 'hex'.
Conversion plants are operating commercially in USA, Canada, France, UK, Russia and China.
Conversion of uranium oxide to UF6 is achieved by a dry fluoride volatility process in the USA, while all other converters use a wet process.
World Primary Conversion capacity
WNA Market Report 2009 * operating capacity estimated at 12,000 to 18,000 tU/yr
After initial refining, which may involve the production of uranyl nitrate, uranium trioxide is reduced in a kiln by hydrogen to uranium dioxide. This is then reacted in another kiln with hydrogen fluoride (HF) to form uranium tetrafluoride. The tetrafluoride is then fed into a fluidised bed reactor with gaseous fluorine to produce UF6. The alternative wet process involves making the tetrafluoride from uranium oxide by a wet process, using aqueous HF.
In detail:
In the dry process, uranium oxide concentrates are first calcined (heated strongly) to drive off some impurities, then agglomerated and crushed.
For the wet process, the concentrate is dissolved in nitric acid. The resulting solution of uranyl nitrate UO2(NO3)2.6H2O is fed into a countercurrent solvent extraction process, using tributyl phosphate dissolved in kerosene or dodecane. The uranium is collected by the organic extractant, from which it can be washed out by dilute nitric acid solution and then concentrated by evaporation. The solution is then calcined in a fluidised bed reactor to produce UO3 (or UO2 if heated sufficiently).
Purified U3O8 from the dry process and purified uranium oxide UO3 from the wet process are then reduced in a kiln by hydrogen to UO2:
U3O8 + 2H2 ===> 3UO2 + 2H2O deltaH = -109 kJ/mole
or UO3 + H2 ===> UO2 + H2O deltaH = -109 kJ/mole
This reduced oxide is then reacted in another kiln with gaseous hydrogen fluoride (HF) to form uranium tetrafluoride (UF4), though in some places this is made with aqueous HF by a wet process:
UO2 + 4HF ===> UF4 + 2H2O deltaH = -176 kJ/mole
The tetrafluoride is then fed into a fluidised bed reactor or flame tower with gaseous fluorine to produce uranium hexafluoride, UF6. Hexafluoride ("hex") is condensed and stored.
UF4 + F2 ===> UF6
Removal of impurities takes place at each step.
The UF6, particularly if moist, is highly corrosive. When warm it is a gas, suitable for use in the enrichment process. At lower temperature and under moderate pressure, the UF6 can be liquefied. The liquid is run into specially designed steel shipping cylinders which are thick walled and weigh over 15 tonnes when full. As it cools, the liquid UF6 within the cylinder becomes a white crystalline solid and is shipped in this form.
The siting, environmental and security management of a conversion plant is subject to the regulations that are in effect for any chemical processing plant involving fluorine-based chemicals.
ENRICHMENT
A number of enrichment processes have been demonstrated historically or in the laboratory but only two, the gaseous diffusion process and the centrifuge process, are operating on a commercial scale. In both of these, UF6 gas is used as the feed material. Molecules of UF6 with U-235 atoms are about one percent lighter than the rest, and this difference in mass is the basis of both processes. Isotope separation is a physical process.*
*One chemical process has been demonstrated to pilot plant stage but not used. The French Chemex process exploited a very slight difference in the two isotopes' propensity to change valency in oxidation/reduction, utilising aqueous (III valency) and organic (IV) phases.
Large commercial enrichment plants are in operation in France, Germany, Netherlands, UK, USA, and Russia, with smaller plants elsewhere. New centrifuge plants are being built in France and USA. Several plants are adding capacity.
World Enrichment capacity - operational and planned (thousand SWU/yr)
source: OECD NEA (2006) Nuclear Energy Data, WNA Market Report 2009.* diffusion Urenco reached 12,200 at the end of 2009. Including its US plant, it expects to reach 15,000 in 2012, and 18,000 by 2015.'Other' includes Resende in Brazil, Kahutab in Pakistan, Rattehallib in India and Natanz in Iran.
'Other' includes Resende in Brazil, Kahutab in Pakistan, Rattehallib in India and Natanz in Iran.
Enrichment accounts for almost half of the cost of nuclear fuel and about 5% of the total cost of the electricity generated. It can also account for the main greenhouse gas impact from the nuclear fuel cycle if the electricity used for enrichment is generated from coal. However, it still only amounts to 0.1% of the carbon dioxide from equivalent coal-fired electricity generation if modern gas centrifuge plants are used, or up to 3% in a worst case situation.
The utilities which buy uranium from the mines need a fixed quantity of enriched uranium in order to fabricate the fuel to be loaded into their reactors. The quantity of uranium they must supply to the enrichment company is determined by the enrichment level required (% U-235) and the tails assay (also % U-235). This is the contracted or transactional tails assay, and determines how much natural uranium must be supplied to create a quantity of Enriched Uranium Product (EUP) - a lower tails assay means that more enrichment services (notably energy) are to be applied. The enricher, however, has some flexibility in respect to the operational tails assay at the plant. If the operational tails assay is lower than the contracted/transactional, the enricher can set aside some surplus natural uranium, which he is free to sell (either as natural uranium or as EUP) on his own account. This is known as underfeeding. The opposite situation, where the operational tails assay is higher, requires the enricher to supplement the natural uranium supplied by the utility with some of his own - this is called overfeeding. In either case, the enricher will base his decision on his plant economics together with uranium and energy prices.
The trend in enrichment technology is to retire obsolete diffusion plants:
Gaseous diffusion process
Commercial uranium enrichment was first carried out by the diffusion process in the USA. It has since been used in Russia, UK, France, China and Argentina as well. It is a very energy-intensive process, requiring about 2400 kWh per SWU. USEC says that electricity accounts for 70% of the production cost at its Paducah plant.
Today only the USA and France use the process on any significant scale. The remaining large USEC plant in the USA was originally developed for weapons programs and has a capacity of some 8 million SWU per year. At Tricastin, in southern France, a more modern diffusion plant with a capacity of 10.8 million kg SWU per year has been operating since 1979 (see photo above). This plant can produce enough 3.7% enriched uranium a year to fuel some ninety 1000 MWe nuclear reactors.
At present the gaseous diffusion process accounts for about 25% of world enrichment capacity. However, though they have proved durable and reliable, most gaseous diffusion plants are now nearing the end of their design life and the focus is on centrifuge enrichment technology which is replacing them.
The diffusion process involves forcing uranium hexafluoride gas under pressure through a series of porous membranes or diaphragms. As U-235 molecules are lighter than the U-238 molecules they move faster and have a slightly better chance of passing through the pores in the membrane. The UF6 which diffuses through the membrane is thus slightly enriched, while the gas which did not pass through is depleted in U-235.
This process is repeated many times in a series of diffusion stages called a cascade. Each stage consists of a compressor, a diffuser and a heat exchanger to remove the heat of compression. The enriched UF6 product is withdrawn from one end of the cascade and the depleted UF6 is removed at the other end. The gas must be processed through some 1400 stages to obtain a product with a concentration of 3% to 4% U-235 .
Centrifuge process
The gas centrifuge process was first demonstrated in the 1940s but was shelved in favour of the simpler diffusion process. It was then developed and brought on stream in the 1960s as the second-generation enrichment technology. It is economic on a smaller scale, eg under 2 million SWU/yr, which enables staged development of larger plants. It is much more energy-efficient than diffusion, requiring only about 50-60 kWh per SWU.
The centrifuge process has been deployed at a commercial level in Russia, and in Europe by Urenco, an industrial group formed by British, German and Dutch governments. Russia's four plants at Seversk, Zelenogorsk, Angarsk and Novouralsk account for some 40% of world capacity. Urenco operates enrichment plants in UK, Netherlands and Germany and is building one in the USA.
In Japan, JNC and JNFL operate small centrifuge plants, the capacity of JNFL's at Rokkasho was planned to be 1.5 million SWU/yr. China has two small centrifuge plants imported from Russia. One at Lanzhou is 0.5 million SWU/yr and the other main one at Hanzhun is operating at 0.5 million SWU/yr and is being doubled in size. Brazil has a small plant which is being developed to 0.2 million SWU/yr. Pakistan has developed centrifuge enrichment technology, and this appears to have been sold to North Korea. Iran has sophisticated centrifuge technology which is being commissioned.
In both France and the USA plants with centrifuge technology are now being built to replace ageing diffusion plants, not least because they are more economical to operate. As noted, a centrifuge plant requires as little as 50 kWh/SWU power (Urenco at Capenhurst, UK, input 62.3 kWh/SWU for the whole plant in 2001-02, including infrastructure and capital works).
The EUR 3 billion French plant operated by Areva - Georges Besse II - is startiing commercial operation and will ramp up to full capacity of 7.5 million SWU/yr in 2016.
The $1.5 billion National Enrichment Facility in New Mexico, USA is using the same 6th generation Urenco technology and first production is expected in 2010, with full initial capacity of 3 million SWU/yr being reached in 2013, and 5.9 million SWU/yr being reached in 2015.
Following this, Areva is building a $2 billion, 3.3 million SWU/yr Eagle Rock plant at Idaho Falls, USA which it expects to commence operation in 2014, ramping up to full production in 2019. In 2009 it applied for doubling in capacity to 6.6 million SWU/yr.
USEC has been building its American Centrifuge Plant in Piketon, Ohio, on the same Portsmouth site where the DOE's experimental plant operated in the 1980s as the culmination of a very major R&D program. Operation from 2012 was envisaged, at a cost of $3.5 billion. It is designed to have an initial annual capacity of 3.8 million SWU, though its licence application is for 7 million SWU to allow for expansion. Authorisation for enrichment up to 10% was sought - most enrichment plants operate up to 5% U-235 product, which is becoming a serious constraint as reactor fuel burnup increases. A demonstration cascade started up in September 2007 with about 20 prototype machines, and a lead cascade of commercial centrifuges started operation in March 2010. However the whole project was largely halted in July 2009 pending further finance, with $1.7 billion having been spent since May 2007. In March 2010 the DOE made $45 million available to USEC for continued development.
Like the diffusion process, the centrifuge process uses UF6 gas as its feed and makes use of the slight difference in mass between U-235 and U-238. The gas is fed into a series of vacuum tubes, each containing a rotor 3 to 5 metres tall and 20 cm diameter.* When the rotors are spun rapidly, at 50,000 to 70,000 rpm, the heavier molecules with U-238 increase in concentration towards the cylinder's outer edge. There is a corresponding increase in concentration of U-235 molecules near the centre. The countercurrent flow set up by a thermal gradient enables enriched product to be drawn off axially, heavier molecules at one end and lighter ones at the other.
* USEC's American Centrifuges are more than 12 m tall and 40-50 cm diameter. The Russian centrifuges are said to be only about one metre tall.
The enriched gas forms part of the feed for the next stages while the depleted UF6 gas goes back to the previous stage. Eventually enriched and depleted uranium are drawn from the cascade at the desired assays.
To obtain efficient separation of the two isotopes, centrifuges rotate at very high speeds, with the outer wall of the spinning cylinder moving at between 400 and 500 metres per second to give a million times the acceleration of gravity.
Although the capacity of a single centrifuge is much smaller than that of a single diffusion stage, its capability to separate isotopes is much greater. Centrifuge stages normally consist of a large number of centrifuges in parallel. Such stages are then arranged in cascade similarly to those for diffusion. In the centrifuge process, however, the number of stages may only be 10 to 20, instead of a thousand or more for diffusion.
Laser processes
Laser enrichment processes have been the focus of interest for some time. They are a possible third-generation technology promising lower energy inputs, lower capital costs and lower tails assays, hence significant economic advantages. None of these processes is yet ready for commercial use, though one is well advanced.
Development of the Atomic Vapour Laser Isotope Separation (AVLIS, and the French SILVA) began in the 1970s. In 1985 the US Government backed it as the new technology to replace its gaseous diffusion plants as they reached the end of their economic lives early in the 21st century. However, after some US$ 2 billion in R&D, it was abandoned in USA in favour of SILEX, a molecular process. French work on SILVA has now ceased, following a 4-year program to 2003 to prove the scientific and technical feasibility of the process. Some 200kg of 2.5% enriched uranium was produced in this.
Atomic vapour processes work on the principle of photo-ionisation, whereby a powerful laser is used to ionise particular atoms present in a vapour of uranium metal. (An electron can be ejected from an atom by light of a certain frequency. The laser techniques for uranium use frequencies which are tuned to ionise a U-235 atom but not a U-238 atom.) The positively-charged U-235 ions are then attracted to a negatively-charged plate and collected. Atomic laser techniques may also separate plutonium isotopes.
The main molecular processes which have been researched work on a principle of photo-dissociation of Uf6 to solid UF5+, using tuned laser radiation as above to break the molecular bond holding one of the six fluorine atoms to a U-235 atom. This then enables the ionized UF5 to be separated from the unaffected UF6 molecules containing U-238 atoms, hence achieving a separation of isotopes. Any process using UF6 fits more readily within the conventional fuel cycle than the atomic process.
The only remaining laser process on the world stage is SILEX, an Australian development which is molecular and utilises UF6. In 1996 USEC secured the rights to evaluate and develop SILEX for uranium (it is also useable for silicon and other elements) but relinquished these in 2003.
In 2006 GE Energy entered a partnership to develop the SILEX process. It provided for GE (now GE-Hitachi) to construct in the USA an engineering-scale test loop, then a pilot plant or lead cascade, which could be operating in 2012, and expanded to a full commercial plant. Apart from US$ 20 million upfront and subsequent payments, the license agreement will yield 7-12% royalties, the precise amount depending on how low the cost of deploying the commercial technology. GE referred to SILEX, which it has rebadged as Global Laser Enrichment (GLE), as "game-changing technology" with a "very high likelihood" of success. GE-Hitachi plans to complete the test loop program in 2010. The initial phase of this to April 2010 was successful in meeting performance criteria, so that engineering design for a commercial facility has commenced. GLE anticipates gleaning sufficient data from the test loop by the end of 2010 to decide whether to proceed with a full-scale commercial enrichment facility. In mid 2008 Cameco bought into the GLE project, paying $124 million for 24% share, alongside GE (51%) and Hitachi (25%).
In October 2007 the two largest US nuclear utilities, Exelon and Entergy, signed letters of intent to contract for uranium enrichment services from GLE. The utilities may also provide GLE with facility licensing and public acceptance support if needed for development of a commercial-scale GLE plant. GEH is operating the GLE test loop at Global Nuclear Fuel's Wilmington, North Carolina fuel fabrication facility - GNF is a partnership of GE, Toshiba, and Hitachi. In mid 2009 GEH submitted the last part of its licence application for this GLE plant, which will take the NRC 30 months to process, to December 2011. If the decision is made to proceed with construction (from early 2012), the GLE commercial production facility at Wilmington, North Carolina would have an annual capacity of 3.5 to 6 million separative work units (SWU).
Applications to silicon and zirconium are also being developed by Silex Systems near Sydney.
Electromagnetic process
A very early endeavour was the electromagnetic isotope separation (EMIS) process. This was developed in the early 1940s in the Manhattan Project to make the highly enriched uranium used in the Hiroshima bomb, but was abandoned soon afterwards. However, it reappeared as the main thrust of Iraq's clandestine uranium enrichment program for weapons discovered in 1992. EMIS uses the same principles as a mass spectrometer (albeit on a much larger scale). Ions of uranium-238 and uranium-235 are separated because they describe arcs of different radii when they move through a magnetic field. The process is very energy-intensive - about ten times that of diffusion.
Aerodynamic processes
Two aerodynamic processes were brought to demonstration stage around the 1970s. One is the jet nozzle process, with demonstration plant built in Brazil, and the other the Helikon vortex tube process developed in South Africa. Neither is in use now, though the latter is the forerunner of new R&D. They depend on a high-speed gas stream bearing the UF6 being made to turn through a very small radius, causing a pressure gradient similar to that in a centrifuge. The light fraction can be extracted towards the centre and the heavy fraction on the outside. Thousands of stages are required to produce enriched product for a reactor. Both processes are energy-intensive - over 3000 kWh/SWU. The Helikon Z-plant in the early 1980s was not commercially oriented and had less than 500,000 SWU/yr capacity. It required some 10,000 kWh/SWU.
The Aerodynamic Separation Process (ASP) being developed by Klydon in South Africa employs similar stationary-wall centrifuges with UF6 injected tangentially. It is based on Helikon but pending regulatory authorisation it has not yet been tested on UF6 - only light isotopes such as silicon. However, extrapolating from results there it is expected to have an enrichment factor in each unit of 1.10 (cf 1.03 in Helikon) with about 1000 kWh/SWU and development of it is aiming for 1.15 enrichment factor and less than 500 kWh/SWU. Projections give an enrichment cost under $100/SWU, with this split evenly among capital, operation and energy input.
One chemical process has been demonstrated to pilot plant stage but not used. The French Chemex process exploited a very slight difference in the two isotopes' propensity to change valency in oxidation/reduction, utilising aqueous (III valency) and organic (IV) phases.
Enrichment of reprocessed uranium
In some countries used fuel is reprocessed to recover its uranium and plutonium, and to reduce the final volume of high-level wastes. The plutonium is normally recycled promptly into mixed-oxide (MOX) fuel, by mixing it with depleted uranium.
Where uranium recovered from reprocessing used nuclear fuel (RepU) is to be re-used, it needs to be converted and re-enriched. This is complicated by the presence of impurities and two new isotopes in particular: U-232 and U-236, which are formed by or following neutron capture in the reactor, and increase with higher burn-up levels. U-232 is largely a decay product of Pu-236, and increases with storage time in used fuel, peaking at about ten years. Both decay much more rapidly than U-235 and U-238, and one of the daughter products of U-232 emits very strong gamma radiation, which means that shielding is necessary in any plant handling material with more than very small traces of it. U-236 is a neutron absorber which impedes the chain reaction, and means that a higher level of U-235 enrichment is required in the product to compensate. Being lighter, both isotopes tend to concentrate in the enriched (rather than depleted) output, so reprocessed uranium which is re-enriched for fuel must be segregated from enriched fresh uranium. The presence of U-236 in particular means that most reprocessed uranium can be recycled only once - the main exception being in the UK with AGR fuel made from recycled Magnox uranium being reprocessed.
All these considerations mean that only RepU from low-enriched, low-burnup used fuel is normally recycled directly through an enrichment plant. For instance, some 16,000 tonnes of RepU from Magnox reactors* in UK has been used to make about 1650 tonnes of enriched AGR fuel, via two enrichment plants. Much smaller quantities have been used elsewhere, in France and Japan. Some re-enrichment, eg for Swiss, German and Russian fuel, is actually done by blending RepU with HEU.
* 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.
A laser process would theoretically be ideal for enriching RepU as it would ignore all but the desired U-235, but this remains to be demonstrated with reprocessed feed.
Tails from enriching reprocessed uranium remain the property of the enricher. Some recycled uranium has been enriched by Tenex at Seversk for Areva, under a 1991 ten-year contract covering about 500 tonnes UF6. French media reports in 2009 alleging that wastes from French nuclear power plants were stored at Seversk evidently refer to tails from this.
Enrichment of depleted uranium tails
Early enrichment activities often left depleted uranium tails with about 0.30% U-235, and there were tens of thousands of tonnes of these sitting around as the property of the enrichment companies. With the wind-down of military enrichment, particularly in Russia, there was a lot of spare capacity unused. Consequently, since the mid 1990s some of the highest-assay tails have been sent to Russia by Areva and Urenco for re-enrichment by Tenex. These arrangements however cease in 2010, though Tenex may continue to re-enrich Russian tails. Tenex now owns all the tails from that secondary re-enrichment, and they are said to comprise only about 0.10% U-235.
After enrichment
The enriched UF6 is converted to UO2 and made into fuel pellets - ultimately a sintered ceramic, which are encased in metal tubes to form fuel rods, typically up to four metres long. A number of fuel rods make up a fuel assembly, which is ready to be loaded into the nuclear reactor.
Depleted uranium and deconversion
Depleted uranium (DU) is stored long-term as UF6 or preferably, after deconversion, as U3O8, allowing HF to be recycled. . To early 2007, about one quarter of the 1.2 million tonnes of DU had been deconverted.
The main deconversion plant is run by Areva NC at Tricastin, France, though the technology has been sold to Russia and a plant is now operational at Zelenogorsk in Siberia. Another plant is planned for New Mexico in the USA.
These use essentially a dry process, with no liquid effluent. It is the same as that used for the enriched portion, albeit at a scale of 20,000 tonnes per year in the one plant.
The UF6 is first vapourised in autoclaves with steam, then the uranyl fluoride is reacted with hydrogen at 700°C to yield an HF by-product for sale and U3O8 powder which is packed into 10-tonne containers for storage.
3UO2F2 + 2H2O + H2 ===> U3O8 + 6HF
Ownership title is normally transferred to the enricher as part of the commercial deal. It is sometimes considered as a waste, though only for legal or regulatory reasons and in the USA, but usually it is understood as a long-term strategic resource which can be used in a future generation of fast neutron reactors. Any much more efficient enrichment process would also make it into an immediately usable resource to supply more U-235. Enrichment companies with ownership of large amounts of depleted uranium are quite clear that their stocks are a significant asset.
Environmental Issuses
With the minor exception of reprocessed uranium, enrichment involves only natural, long-lived radioactive materials; there is no formation of fission products or irradiation of materials, as in a reactor. Feed, product, and depleted material are all in the form of UF6, though the depleted uranium may be stored long-term as the more stable U3O8.
Uranium is only weakly radioactive, and its chemical toxicity - especially as UF6 - is more significant than its radiological toxicity. The protective measures required for an enrichment plant are therefore similar to those taken by other chemical industries concerned with the production of fluorinated chemicals.
Uranium hexafluoride forms a very corrosive material (HF - hydrofluoric acid) when exposed to moisture, therefore any leakage is undesirable. Hence:
Sources:Heriot, I.D. (1988). Uranium Enrichment by Centrifuge, Report EUR 11486, Commission of the European Communities, Brussels.Kehoe, R.B. (2002). The Enriching Troika, a History of Urenco to the Year 2000. Urenco, Marlow UK.Wilson, P.D. (ed)(1996). The Nuclear Fuel Cycle - from ore to wastes. Oxford University Press, Oxford UK.IAEA 2007, Management of Reprocessed Uranium - current status and future prospects, Tecdoc 1529.