Chemistry of Uranium ore processing

Crushed ore is mixed with hot water to a 58% solids slurry. The solids slurry is then processed through 
a series of tanks, where sulfuric acid, sodium chlorate, and steam are used to extract the uranium from 
 the solids slurry. The average leaching efficiency for this process is 98.5%. The uranium-bearing solution
 is then decanted and directed to a solvent extraction (SX) process for further purification. In this 
extraction step, the dissolved uranium is transferred from the feed solution into the organic solvent. 
Next a stripping step recovers the uranium into a sodium chloride aqueous phase after which the
 barren solvent is recycled. The average efficiency of the SX circuit is 99.9%. The high-grade 
“pregnant” strip solution from SX goes to the next stage where magnesia slurry is added to 
precipitate magnesium diuranate. The yellow cake precipitate is then thickened, dried, re-crushed
 and packed into industry standard 220 litre steel drums for shipment to customers.

Uranium meeting nuclear-grade specifications is usually obtained from yellow cake through a
 tributyl phosphate solvent-extraction process. First, the yellow cake is dissolved in nitric acid to
 prepare a feed solution. Uranium is then selectively extracted from this acid feed by tributyl 
phosphate diluted with kerosene or some other suitable hydrocarbon mixture. Finally, uranium is
 stripped from the tributyl phosphate extract into acidified water to yield a highly purified uranyl nitrate,
 UO2(NO3)2.

Summary: The acid leaching process  comprises the following reactions:

Oxidation and dissolution of U(IV): UO2(s) + Cl2(aq) → (UO2)2+(aq) + 2Cl–(aq)

Dissolution of U(VI): UO3(s) + 2H+(aq) → (UO2)2+(aq) + H2O(l)

Neutralization: 6H+(aq) + UO2(CO3)34-(aq) → (UO2)2+(aq) + 3CO2(g) + 3H2O(l)

Precipitation: (UO2)2+(aq) + H2O2(l) → UO4·2H2O(s) + 2H+(aq) + 2H2O(l)

Reduction: (UO2)2+(aq) + C6H8O6(aq) + 2H+(aq) → U4+(aq) + C6H6O6(aq) + 2H2O(l)

Precipitation: U4+(aq) + 4HF(aq) + 2.5H2O(l) → UF4·2.5H2O(s) + 4H+(aq)

Uranium Minerals
Primary uranium minerals
Name	Chemical Formula
uraninite or pitchblende	UO2
coffinite	U(SiO4)1–x(OH)4x
brannerite	UTi2O6
davidite	(REE)(Y,U)(Ti,Fe3+)20O38
thucholite	Uranium-bearing pyrobitumen
Secondary uranium minerals
Name	Chemical Formula
autunite	Ca(UO2)2(PO4)2 x 8-12 H2O
carnotite	K2(UO2)2(VO4)2 x 1–3 H2O
gummite	gum like amorphous mixture of various uranium minerals
saleeite	Mg(UO2)2(PO4)2 x 10 H2O
torbernite	Cu(UO2)2(PO4)2 x 12 H2O
tyuyamunite	Ca(UO2)2(VO4)2 x 5-8 H2O
uranocircite	Ba(UO2)2(PO4)2 x 8-10 H2O
uranophane	Ca(UO2)2(HSiO4)2 x 5 H2O
zeunerite	Cu(UO2)2(AsO4)2 x 8-10 H2O

Uranium Processing :

The crushed and ground ore, or the underground ore in the case of ISL mining, is leached with sulfuric acid:

UO3 + 2H+ ====> UO22+ + H2O

UO22+ + 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.

The uranium in solution is recovered in a resin/polymer ion exchange (IX) or liquid ion exchange (solvent extraction – SX) system. The pregnant liquor from acid ISL or heap leaching is treated similarly.

Further treatment for IX involves stripping the uranium from the resin/polymer either with a strong acid or chloride solution or with a nitrate solution in a semi-continuous cycle. The pregnant solution produced by the stripping cycle is then precipitated by the addition of ammonia, hydrogen peroxide, caustic soda or caustic magnesia. Solvent extraction is a continuous loading/stripping cycle involving the use of an organic liquid to carry the extractant which removes the uranium from solution.

Typically, in solvent extraction, tertiary amines* are used in a kerosene diluent, and the phases move countercurrently.

2R3N + H2SO4 ====> (R3NH)2SO4

2 (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.

Enrichment

The vast majority of all nuclear power reactors require 'enriched' uranium fuel in which the proportion of the uranium-235 isotope has been raised from the natural level of 0.7% to about 3.5% to 5%.  The enrichment process needs to have the uranium in gaseous form, so on the way from the mine it goes through a conversion plant which turns the uranium oxide into uranium hexafluoride.

The enrichment plant concentrates the useful uranium-235, leaving about 85% of the uranium by separating gaseous uranium hexafluoride into two streams: One stream is enriched to the required level of uranium-235 and then passes to the next stage of the fuel cycle. The other stream is depleted in uranium-235 and is called 'tails' or depleted uranium. It is mostly uranium-238 and has little immediate use. 

Today's enrichment plants use the centrifuge process, with thousands of rapidly-spinning vertical tubes. Research is being conducted into laser enrichment, which appears to be a promising new technology.

A small number of reactors, notably the Canadian CANDU reactors, do not require uranium to be enriched.

Fuel fabrication

About 27 tonnes of fresh fuel is required each year by a 1000 MWe nuclear reactor. In contrast, a coal power station requires more than two and a half million tonnes of coal to produce as much electricity. (1)Enriched UF6 is transported to a fuel fabrication plant where it is converted to uranium dioxide powder. This powder is then pressed to form small fuel pellets, which are then heated to make a hard ceramic material. The pellets are then inserted into thin tubes to form fuel rods. These fuel rods are then grouped together to form fuel assemblies, which are several meters long. 

The number of fuel rods used to make each fuel assembly depends on the type of reactor. A pressurized water reactor may use between 121-193 fuel assemblies, each consisting of between 179-264 fuel rods. A boiling water reactor has between 91-96 fuel rods per assembly, with between 350-800 fuel assemblies per reactor.

Three stage Nuclear Program [ India ]

In the first stage, Heavy water reactors using enriched uranium derived from India’s limited uranium reserve, would be constructed and begin operating. The use of heavy water reactors meant that India did not need to to develop expensive and power demanding uranium enrichment facilities.

During the second stage, India was to construct Fast Breeder Reactors, which burned plutonium reprocessed from the spent fuel of the heavy water reactors as well as their depleted uranium. India needed to develop breeder technology quickly, because it had limited uranium resources. Breeders allowed India’s uranium supply to be used much more efficiently.

During the third stage thorium was to be bred, and U-233 would fuel Indian power reactors.

 Natural Uranium contains two isotopes, U-235 (0.7%) and U-238 (99.3%). Out of these two isotopes, only U-235 is useful for nuclear reactors, as it is fissile. U-238 is not fissile, similar to Th-232. However, this U-238 gets converted into Plutonium (Pu-239) during its stay inside the Uranium reactors by absorbing one neutron. This Pu-239 can then be extracted and used as fuel in Fast Breeder reactors. Therefore to sustain the Fast Breeder Reactors, enough Plutonium from Uranium based reactors is necessary. The only way it can be done is to have enough operational Uranium based reactors. This is why India is importing Uranium to sustain Uranium based reactors.

As mentioned above, Pu-239 will be used in Fast Breeder Reactors as fuel, but a blanket, or coating of Th-232 will be placed over Pu-239 (the fuel). This Th-232, during its stay inside the Fast Breeder Reactor, will get converted into Uranium-233 by absorbing one neutron. Uranium-233 is fissile but is not naturally occurring.

This Uranium-233  can then be extracted from the spent-fuel, and used as fuel in another type of reactors. Now, if you place a blanket of Th-232 over this Uranium-233 fuel, that blanket will again get converted into Uranium-233 during its stay inside the reactor by absorbing one neutron, and we will have a process where fuel can be re-generated inside the reactor! Though Uranium-233 is the fuel in these reactors, they are also termed as Thorium based reactors.

Thus in order to reach the Thorium based energy generation, building enough Plutonium stock for fast breeder reactors is necessary, which can only be done by having enough Uranium based electricity generation.