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Occurrence
Extraction
(in Australia):
The ore is crushed and ground to liberate the mineral particles. It is then leached with sulfuric acid:
UO3 + 2H+ ====> UO22+ + H2OUO22+ + 3SO42- ====> UO2(SO4)34-
The UO2 is oxidised to UO3.
With some ores, carbonate leaching is used to form a soluble uranyl tricarbonate ion: UO2(CO3)34-. This can then be precipitated with an alkali, eg as sodium or magnesium diuranate. Alkaline leaching is not undertaken in Australia at present.
Two methods have been used for concentration and purification of uranium: ion exchange and solvent extraction. Early operations in Australia used ammonium type resins in polystyrene beads for ion exchange, but solvent extraction is now in general use.
In solvent extraction, tertiary amines* are used in a kerosene diluent, and the phases move countercurrently.
2R3N + H2SO4 ====> (R3NH)2SO42 (R3NH)2SO4 + UO2(SO4)34- ====> (R3NH)4UO2(SO4)3 + 2SO42-
* "R" is an alkyl (hydrocarbon) grouping, with single covalent bond.
The loaded solvents may then be treated to remove impurities. First, cations are removed at pH 1.5 using sulfuric acid and then anions are dealt with using gaseous ammonia.
The solvents are then stripped in a countercurrent process using ammonium sulfate solution.
(R3NH)4UO2(SO4)3 + 2(NH4)2SO4 ====> 4R3N + (NH4)4UO2(SO4)3 + 2H2SO4
Precipitation of ammonium diuranate is achieved by adding gaseous ammonia to neutralise the solution (though in earlier operations caustic soda and magnesia were used).
2NH3 + 2UO2(SO4)34- ====> (NH4)2U2O7 + 4SO42-
The diuranate is then dewatered and roasted to yield U3O8 product, which is the form in which uranium is marketed and exported.
In situ leaching of sandy unconsolidated orebodies is an important mining method in the USA, Kazakhstan and Australia.
Typically the uranium is recovered by circulating weakly acidified groundwater with oxygen added through injection and recovery wells. In the plant, the pregnant liquor then deposits uranium on a plastic ion exchange resin. The loaded resin is washed with sulfuric acid, hydrogen peroxide is added, and uranyl peroxide precipitated. With low temperature drying this becomes U3O8.
Alternatively, an oxygenated carbonate leach may be employed, depending on the chemistry of the orebody. In that case a soluble uranyl tricarbonate ion: UO2(CO3)34- is formed and then precipitated with an alkali, eg as sodium or magnesium diuranate.
Refining and Conversion to UF6 prior to Enrichment
(in Europe and North America)
The mixed uranium oxide concentrate U3O8 received by the refinery is dissolved in nitric acid. The resulting solution of uranium 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 (heated strongly) to produce pure UO3.
Most nuclear reactors require uranium to be enriched from its natural isotopic composition of 0.7% U-235 (most of the rest being U-238) to 3.5-4% U-235. The uranium therefore needs to be in a gaseous form and the most convenient way of achieving this is to convert the uranium oxides to uranium hexafluoride.
After purification, the uranium oxide UO3 is reduced in a kiln by hydrogen to UO2.
UO3 + H2 ====> UO2 + H2O ............ delta H = -109 kJ/mole
This reduced oxide is then reacted with gaseous hydrogen fluoride in another kiln to form uranium tetrafluoride, UF4, though in some places this is made with aqueous HF by a wet process.
UO2 + 4HF ====> UF4 + 2H2O ............. delta H = -176 kJ/mole
The tetrafluoride is then fed into a fluidised bed reactor with gaseous fluorine to produce uranium hexafluoride, UF6. Hexafluoride is condensed and stored.
UF4 + F2 ====> UF6
Enrichment is a physical process using either centrifuges or fractionation by diffusion of the gaseous uranium hexafluoride. See Enrichment paper.
The Reactor Fuel
After enrichment, the hexafluoride is turned into UO2, which is made into pellets and these are assembled into fuel rods for the reactor.
In the reactor the nuclear fission chain reaction produces neutrons which cause further fission in U-235 atoms. The fission of a U-235 atom typically releases about 200 MeV*, or 3.2 x 10-11 Joule, (contrasting with 4 ev or 6.5 x 10-19 J per molecule of carbon dioxide released in the combustion of carbon). The fission reaction produces fission products such as Ba, Kr, Sr, Cs, I, and Xe with atomic masses distributed around 95 and 135. See also Physics paper in this series.
Some of the U-238 in the reactor core becomes plutonium-239 and Pu-240. The Pu-239 is fissile in the same way as U-235. If fuel is left in the reactor for a typical three years, about two thirds of the Pu-239 is fissioned with the U-235, leaving a high proportion of Pu-240.
* these are total available energy release figures, consisting of kinetic energy values (Ek) of the fission fragments plus neutron, gamma and delayed energy releases which add about 30 MeV.
In a typical coal or nuclear power station making steam to turn turbines, the thermal efficiency is usually around 33%. That is to say, the energy released by combustion or fission results in about one third as much energy being produced as electricity.
Reprocessing of Spent Fuel
Used fuel may be reprocessed to recover the uranium and plutonium in it for recycle. All commercial reprocessing plants use the PUREX process (see Figure), which has been proved over some fifty years. However, a number of variations of it are now being developed.
The used fuel (which is highly radioactive) is chopped up and dissolved in hot concentrated nitric acid. The first stage separates the uranium and plutonium in the aqueous nitric acid stream from the fission products and minor actinides by a countercurrent solvent extraction process, using tributyl phosphate dissolved in kerosene or dodecane. In a pulsed column uranium and plutonium enter the organic phase while the fission products and other elements remain in the aqueous raffinate. (Future processes will likely then separate the minor actinides from the fission products.)
In a second pulsed column uranium is separated from plutonium by reduction with excess U4+ added to the aqueous stream. Plutonium is then transferred to the aqueous phase while the mixture of U4+ and U6+ remains in the organic phase. It is then stripped from the organic solvent with dilute nitric acid.
The plutonium nitrate is concentrated by evaporation then subject to an oxalate precipitation process followed by calcination to produce PuO2 in powder form. The uranium nitrate is concentrated by evaporation and calcined to produce UO3 in powder form. It is then converted to UO2 product by reduction in hydrogen.
Recycled plutonium oxide is mixed with depleted uranium to make mixed-oxide fuel (MOX). The particular blend (to achieve the desired fissile proportion) of the two oxides are mechanically milled to form a solid solution of U-PuO2. This is then made into fuel pellets which are sintered and assembled into fuel rods.
In the future, reprocessing will move to variants of PUREX which leave the plutonium mixed with other actinides and/or with some uranium. The French COEX process leaves some uranium with the plutonium and possibly also neptunium, the US UREX+ process leaves plutonium with other transuranics as fuel for a fast neutron reactor. Japanese reprocessing is tied in with French developments.
Wastes
The high-level liquid wastes from the first extraction cycles are concentrated by evaporation, stored for some years in shielded and cooled tanks, then calcined to produce some 35 kg of solids which are incorporated in borosilicate glass. The glass contains about 11% radioactive oxides. It is poured into heavy stainless steel flasks, a lid is welded on and in UK and France they are stored pending deep geological disposal.
See also information paper on Processing of Nuclear Wastes.