- Most of the 470 commercial nuclear power reactors operating or under construction in the world today require uranium 'enriched' in the U-235 isotope for their fuel.
- Two commercial processes are employed for this enrichment. Another process based on laser excitation is under development in Australia and USA.
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.)
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 and Russia.
After initial refining, which may involve the production of uranyl nitrate, uranium trioxide is reduced in a kiln by hydrogen to uranium dioxide (UO2). This is then reacted in another kiln with hydrogen fluoride (HF) to form uranium tetrafluoride (UF4). The tetrafluoride is then fed into a fluidised bed reactor with gaseous fluorine to produce UF6. Removal of impurities takes place at each step.
An alternative wet process involves making the UF4 from UO2 by a wet process, using aqueous HF.
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.
A number of enrichment processes have been demonstrated 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.
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.
||capacity in 2002
x 1000 kg SWU/yr
|Germany-Netherlands-UK - Urenco
source: OECD NEA (2003), Nuclear Energy Data; Nuclear Engineering International (2003), World Nuclear Handbook, USEC, JNFL.
The capacity of enrichment plants is measured in terms of 'separative work units' or SWU. The SWU is a complex unit which is a function of the amount of uranium processed and the degree to which it is enriched (ie the extent of increase in the concentration of the U-235 isotope relative to the remainder) and the level of depletion of the remainder. The unit is strictly: Kilogram Separative Work Unit, and it measures the quantity of separative work performed to enrich a given amount of uranium a certain amount. It is thus indicative of energy used in enrichment when feed and product quantities are expressed in kilograms. The unit 'tonnes SWU' is also used.
For instance, to produce one kilogram of uranium enriched to 3% U-235 requires 3.8 SWU if the plant is operated at a tails assay 0.25%, or 5.0 SWU if the tails assay is 0.15% (thereby requiring only 5.1 kg instead of 6.0 kg of natural U feed).
About 100-120,000 SWU is required to enrich the annual fuel loading for a typical 1000 MWe light water reactor. Enrichment costs are substantially related to electrical energy used. The gaseous diffusion process consumes about 2500 kWh (9000 MJ) per SWU, while modern gas centrifuge plants require only about 50 kWh (180 MJ) per SWU.
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 trend in enrichment technology is to retire obsolete diffusion plants:
|HEU ex weapons
The electromagnetic isotope separation (EMIS) process 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.
Two aerodynamic processes were brought to demonstration stage. 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. They depend on a high-speed gas stream bearing the UF 6 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.
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.
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. 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 40% 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 seems likely to replace them.
The large Tricastin enrichment plant in France (beyond cooling towers)
The four nuclear reactors in the foreground provide over 3000 MWe power for it.
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 UF 6 product is withdrawn from one end of the cascade and the depleted UF 6 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.
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 has been deployed at a commercial level in Russia and in Europe by Urenco, an industrial group formed by British, German and Dutch companies. 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 also has a small centrifuge plant imported from Russia at Lanzhou, which is expected to reach 0.5 million SWU/yr about 2005. Another small plant at Hanzhong is operating and was to reach 0.5 million SWU/yr by 2003. 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 had not been commissioned as of early 2006.
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 is expected to start commercial operation in 2009 and ramp up to full capacity of 7.5 million SWU/yr in 2018. The $1.5 billion US plant in New Mexico will use the same 6th generation Urenco technology and first production is expected in 2008, with full capacity of 3 million SWU/yr being reached in 2013.
USEC is building its American Centrifuge Plant in Piketon, Ohio, on the same Portsmouth site where the DOE's experimental plant operated in the 1980s, involving 1300 centrifuges as the culmination of a very major R&D program. The Lead Cascade demonstration plant, is due to start operation in mid 2007. For the main centrifuge plant initial annual capacity of 3.5 million SWU from 2011 is envisaged, costing $1.7 billion, 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 bank of centrifuges at a Urenco plant
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 one to two metres long and 15-20 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. These concentration changes are enhanced by inducing the gas to circulate axially within the cylinder.
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 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. Any process using UF 6 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 process. It provides for GE to construct in the USA an engineering-scale test loop (3 years) then a pilot plant or lead cascade. A full commercial plant would then follow. 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 as "game-changing technology" with a "very high likelihood" of success. The SILEX process is now at prototype stage with Silex Systems near Sydney. Applications to silicon and zirconium are also being developed.
Enrichment of reprocessed uranium
In some countries spent 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 spent nuclear fuel 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 neutron capture in the reactor. 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 the plant. 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.
Both diffusion and centrifuge processes can be used for re-enrichment, though contamination issues prevent commercial application of the former. A laser process would theoretically be ideal as it would ignore all but the desired U-235, but this remains to be demonstrated with reprocessed feed.
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 is stored long-term as UF6 or preferably, after deconversion, as U3O8, allowing HF to be recycled. Ownership title is normally transferred to the enricher as part of the commercial deal. At present the only deconversion plant is in France, but others are planned. It is sometimes considered as a waste, but usually as a long-term strategic resource which can be used in a future generation of fast neutron reactors.
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 corrosive material (HF) when exposed to moisture, therefore any leakage is undesirable. Hence:
- in almost all areas of a centrifuge plant the pressure of the UF 6 gas is maintained below atmospheric pressure and thus any leakage could only result in an inward flow;
- double containment is provided for those few areas where higher pressures are required,
- effluent and venting gases are collected and appropriately treated.
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.