Chemistry of Uranium ore processing

Primary uranium minerals, include uraninite and pitchblende (the latter a variety of uraninite). The uranium in these two ores occurs in the form of uranium dioxide, which—owing to oxidation—can vary in exact chemical composition from UO₂ to UO_2.67. Other uranium ores of economic importance are autunite, a hydrated calcium uranyl phosphate; tobernite, a hydrated copper uranyl phosphate; coffinite, a black hydrated uranium silicate; and carnotite, a yellow hydrated potassium uranyl vanadate.

Uranium ores occur in deposits that are both near-surface and very deep (e.g., 300 to 1,200 metres, ). The deep ores sometimes occur in seams as thick as 30 metres. 

Roasted uranium ores are leached of their uranium values by both acidic and alkaline aqueous solutions. For the successful operation of all leaching systems, uranium must either be initially present in the more stable hexavalent state or be oxidized to that state in the leaching process.

Acid leaching is commonly performed by agitating an ore-leach mixture for 4 to as long as 48 hours at ambient temperature. Except in special circumstances, sulfuric acid is the leachant used; it is supplied in amounts sufficient to obtain a final leach liquor at about pH 1.5. Sulfuric acid leaching circuits commonly employ either manganese dioxide or chlorate ion to oxidize the tetravalent uranium ion (U⁴⁺) to the hexavalent uranyl ion (UO₂²⁺). Typically, about 5 kilograms (11 pounds) of manganese dioxide or 1.5 kilograms of sodium chlorate per ton suffice to oxidize tetravalent uranium. In any case, the oxidized uranium reacts with the sulfuric acid to form a uranyl sulfate complex anion, [UO₂(SO₄)₃]^4-.

Uranium ores that contain significant amounts of basic minerals such as calcite or dolomite are leached with 0.5 to 1 molar sodium carbonate solutions. Although a variety of reagents has been studied and tested, oxygen is the uranium oxidant of choice. Typically, candidate ores are leached in air at atmospheric pressure and at 75° to 80° C (167° to 175° F) for periods that vary with the particular ore. The alkaline leachant reacts with uranium to form a readily soluble uranyl carbonate complex ion, [UO₂(CO₃)₃]^4-.

The complex ions [UO₂(CO₃)₃]^4- and [UO₂(SO₄)₃]^4- can be sorbed from their respective leach solutions by ion-exchange resins. These special resins—characterized by their sorption and elution kinetics, particle size, stability, and hydraulic properties—can be used in a variety of processing equipment—e.g., fixed-bed, moving-bed, basket resin-in-pulp, and continuous resin-in-pulp. Conventionally, sodium and ammonium chloride or nitrate solutions are then used to elute the sorbed uranium from the exchange resins.

Uranium can also be removed from acidic ore leach-liquors through solvent extraction. In industrial methods, alkyl phosphoric acids—e.g., di(2-ethylhexyl) phosphoric acid—and secondary and tertiary alkyl amines are the usual solvents. As a general rule, solvent extraction is preferred over ion-exchange methods for acidic leachates containing more than one gram of uranium per litre. Solvent extraction is not useful for recovery of uranium from carbonate leach liquors, however.

Prior to final purification, uranium present in acidic solutions produced by the ion-exchange or solvent-extraction processes described above, as well as uranium dissolved in carbonate ore leach solutions, is typically precipitated as a polyuranate. From acidic solutions, uranium is precipitated by addition of neutralizers such as sodium hydroxide, magnesia, or (most commonly) aqueous ammonia. Uranium is usually precipitated as ammonium diuranate, (NH₄)₂U₂O₇. From alkaline solutions, uranium is most often precipitated by addition of sodium hydroxide, producing an insoluble sodium diuranate, Na₂U₂O₇. It can also be precipitated by acidification (to remove carbon dioxide) and then neutralization (to remove the uranium) or by reduction to less soluble tetravalent uranium. In all cases, the final uranium precipitate, commonly referred to as yellow cake, is dried. In some cases—e.g., with ammonium diuranate—the yellow cake is ignited, driving off the ammonia and oxidizing the uranium to produce uranium trioxide (UO₃) or the more complex triuranium octoxide (U₃O₈). In all cases, the final product is shipped to a central uranium-purification facility.

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.

Uranyl nitrate is produced by the ore-processing operations described above as well as by solvent extraction from irradiated nuclear reactor fuel . In either case, it is an excellent starting material for conversion to uranium metal or for eventual enrichment of the uranium-235 content. Both of these routes conventionally begin with calcining the nitrate to UO₃ and then reducing the trioxide with hydrogen to uranium dioxide (UO₂). Subsequent treatment of powdered UO₂ with gaseous hydrogen fluoride (HF) at 550° C (1,025° F) produces uranium tetrafluoride (UF₄) and water vapour, as in the following reaction:

UO2 + 4HF---> UF4 + 2H2O

This hydrofluorination process is usually performed in a fluidized-bed reactor

Conversion to uranium metal is accomplished through the Ames process, in which UF₄ is reduced with magnesium (Mg) at temperatures exceeding 1,300° C (2,375° F). (In an often-used modification of the Ames process, calcium metal is substituted for magnesium.) Because the vapour pressureof magnesium metal is very high at 1,300° C, the reduction reaction is performed in a refractory-lined, sealed container, or “bomb.” Bombs charged with granular UF₄ and finely divided Mg (the latter in excess) are heated to 500° to 700° C (930° to 1,300° F), at which point an exothermic  reaction occurs. The heat of reaction is sufficient to liquefy the conversion contents of the bomb, which are essentially metallic uranium and a slag of magnesium fluoride (MgF₂):

UF4 +2Mg--> U + 2MgF2

When the bomb is cooled to ambient temperature, the massive uranium metal obtained is, despite its hydrogen content, the best-quality uranium metal available commercially and is well suited for rolling into fuel shapes for nuclear reactors.

Uranium tetrafluoride can also be fluorinated at 350° C (660° F) with fluorine gas to volatile uranium hexafluoride (UF₆), which is fractionally distilled to produce high-purity feedstock for isotopic enrichment. Any of several methods—gaseous diffusion, gas centrifugation, liquid thermal diffusion—can be employed to separate and concentrate the fissile uranium-235 isotope into several grades, from low-enrichment (2 to 3 percent uranium-235) to fully enriched (97 to 99 percent uranium-235). Low-enrichment uranium is typically used as fuel for light-water nuclear reactors.

After enrichment, UF₆ is reacted in the gaseous state with water vapour to yield hydrated uranyl fluoride (UO₂F₂ · H₂O). Hydrogen reduction of the uranyl fluoride produces powdered UO₂, which can be used to prepare ceramic nuclear reactor fuel. In addition, UO₂ obtained from enriched UF₆ or from UF₆ that has been depleted of its uranium-235 content can be hydrofluorinated to yield UF₄, and the tetrafluoride can then be converted to uranium metal in the Ames process described above.

India's plans for thorium cycle

With huge resources of easily-accessible thorium and relatively little uranium, India has made utilization of thorium for large-scale energy production a major goal in its nuclear power programme, utilising a three-stage concept.

1. Pressurised heavy water reactors (PHWRs) and light water reactors fuelled by natural uranium producing plutonium that is separated for use in fuels in its fast reactors and indigenous advanced heavy water reactors.
2. Fast breeder reactors (FBRs) will use plutonium-based fuel to extend their plutonium inventory. The blanket around the core will have uranium as well as thorium, so that further plutonium (particularly Pu-239) is produced as well as U-233.
3. Advanced heavy water reactors (AHWRs) will burn thorium-plutonium fuels in such a manner that breeds U-233 which can eventually be used as a self-sustaining fissile driver for a fleet of breeding AHWRs. 

In all of these stages, used fuel needs to be reprocessed to recover fissile materials for recycling.

India is focusing and prioritizing the construction and commissioning of its fleet of 500 MWe sodium-cooled fast reactors in which it will breed the required plutonium which is the key to unlocking the energy potential of thorium in its advanced heavy water reactors. This will take another 15-20 years, and so it will still be some time before India is using thorium energy to any extent. The 500 MWe prototype FBR under construction in Kalpakkam was expected to start up in 2014, but 2018 is now the target date.

Fast breader reactor [India]

After 15 years of research, development and construction, India’s first Prototype Fast Breeder Reactor (PFBR), the second of its kinds in the world is nearing completion in Kalpakkam. This plant has the potential of becoming the greatest source of renewable energy for the country.

Nuclear scientists of India have been working on a huge nuclear power plant on the shores of the Bay of Bengal, in Kalpakkam, Tamil Nadu. Unlike other nuclear power plants in the country, the one in Kalpakkam is a fast breeder nuclear reactor.

This reactor differs from conventional nuclear power plants as it can produce 70 per cent more energy and is safer than its traditional counterparts.

The Kalpakkam PFBR is using uranium-238 not thorium, to breed new fissile material, in a sodium-cooled fast reactor design. The power island of this project is being engineered by Bharat Heavy Electricals Limited, largest power equipment utility of India.

The surplus plutonium (or uranium-233 for thorium reactors) from each fast reactor can be used to set up more such reactors and grow the nuclear capacity in tune with India's needs for power. The PFBR is part of the three-stage nuclear power program.

India has the capability to use thorium cycle based processes to extract nuclear fuel. This is of special significance to the Indian nuclear power generation strategy as India has one of the world's largest reserves of thorium, which could provide power for more than 10,000 years,^[3][4] and perhaps as long as 60,000 years.

The design of this reactor was started in the 1980s, as a prototype for a 600 MW FBR. Construction of the first two FBR are planned at Kalpakkam, after a year of successful operation of the PFBR. Other four FBR are planned to follow beyond 2030, at sites to be defined.

As of July 2017 the reactor is in final preparation to go critical.    

The reactor will be a pool-type reactor with 1,750 tonnes of sodium as coolant. Designed to generate 500 MWe of electrical power, with an operational life of 40 years, it will burn a mixed uranium-plutonium MOX fuel, a mixture of PuO2 and UO2. A fuel burnup of 100 GWd/t is expected.

The Advanced Fuel Fabrication Facility (AFFF), under the direction of BARC, Tarapur, is responsible for the fuel rods manufacturing. AFFF comes under " Nuclear Recycle Board" of Bhabha Atomic Research Center. AFFF has been responsible for fuel rod manufacturing of various types in the past.  

The fact that PFBR will be cooled by liquid sodium creates additional safety requirements to isolate the coolant from the environment, since sodium explodes if it comes into contact with water and burns when in contact with air. Another hazard associated with the use of sodium as a coolant is the absorption of neutrons to generate the radioactive isotope 24Na.

There are two independent shutdown systems installed, designed to shut the reactor down effectively within a second. The reactor also has decay heat removal systems consisting of four independent circuits of 8MWt capacity each.

Thorium processing

The major commercial source of thorium is monazite, an anhydrous rare earth phosphate with the chemical formula (Ce,La,Nd,Th)PO₄. Typically, 3 to 5 percent of the metal content of monazite is thorium (in the form of thorium dioxide, ThO₂). Much of the world’s current demand for thorium metal and its compounds is satisfied by mining placers along India’s Malabar Coast, 

The dredged monazite is admixed with a variety of other minerals, including silica, magnetite, ilmenite, zircon, and garnet. Concentration is accomplished by washing out lighter minerals in shaking tables and passing the resulting monazite fraction through a series of electromagnetic separators, which separate monazite from other minerals by virtue of their different magnetic permeabilities.  

Acidic and alkaline digestion

Although monazite is very stable chemically, it is susceptible to attack by both strong mineral acids (e.g., sulfuric acid, H₂SO₄) and alkalies (e.g., sodium hydroxide, NaOH). In the acid treatment, finely ground monazite sand is digested at 155 to 230 °C (310 to 445 °F) with highly concentrated (93 percent) H₂SO₄. This converts both the phosphate and the metal content of the monazite to water-soluble species. The resulting solution is contacted with aqueous ammonia, first precipitating hydrated thorium phosphate as a gelatinous mass and then metathesizing the thorium phosphate to thorium hydroxide. Finally, the crude thorium hydroxide is dissolved in nitric acid to produce a thorium nitrate-containing feed solution suitable for final purification by solvent extraction.

In alkaline digestion, finely ground monazite sand is carefully treated with a concentrated NaOH solution at 138 °C (280 °F) to produce a solid hydroxide product. Any one of several mineral acids is then used to dissolve this solid residue. For example, treatment with hydrochloric acid yields a solution of thorium and rare earth chlorides. Conventionally, thorium is partially separated from the rare earths by addition of NaOH to the acidic chloride solution. The crude thorium hydroxide precipitate is then dissolved in nitric acid for final purification by solvent extraction.  

Solvent extraction 

For the purification of thorium from residual rare earths and other contaminants present in nitric acid feed solutions, the crude thorium nitrate concentrate is usually contacted with a solution of tributyl phosphate diluted by a suitable hydrocarbon. The resulting organic extract, containing the thorium (and any uranium that may be present), is then contacted countercurrently with a small volume of nitric acid solution in order to remove contaminating rare earths and other metallic impurities to acceptable levels. Finally, the scrubbed tributyl phosphate solution is contacted with a dilute nitric acid solution; this removes, or strips, thorium from the organic solvent into the aqueous solution while retaining uranium (if present) in the organic phase. Thermal concentration of the purified thorium nitrate solution yields a product suitable for the fabrication of gas mantles. The nitrate can also be calcined to ThO₂, which is incorporated into ceramic fuel elements for nuclear reactors or is converted to thorium metal. 

Reduction to the metal

Powdered ThO₂ can be fluorinated with gaseous hydrogen fluoride (HF), yielding thorium tetrafluoride (ThF₄). The metal is obtained by the Spedding process, in which powdered ThF₄ is mixed with finely divided calcium (Ca) and a zinc halide (either zinc chloride or zinc fluoride) and placed in a sealed, refractory-lined “bomb.” Upon heating to approximately 650 °C (1,200 °F), an exothermic reaction ensues that reduces the thorium and zinc to metal and produces a slag of calcium chloride or calcium fluoride:

After solidification, the zinc-thorium alloy product is heated above the boiling point of zinc (907 °C, or 1,665 °F) but below the melting temperature of thorium. This evaporates the zinc and leaves a highly purified thorium sponge, which is melted and cast into ingots.

Uranium ore processing

Uranium ore is mined in several ways: by open pit, underground or by leaching uranium from low-grade ores  Uranium ore typically contains 0.1 to 0.25 percent of actual uranium oxides. So extensive measures must be employed to extract the metal from its ore. Uranium ore is crushed and rendered into a fine powder and then leached with either an acid or alkali. The leachate is then subjected to one of several sequences of precipitation, solvent extraction, and ion exchange. The resulting mixture, called yellowcake, contains at least 75 percent uranium oxides. Yellowcake is then generally further refined using nitric acid to create a solution of uranyl nitrate. Additional solvent extraction procedures finish the process.

Commercial-grade uranium can be produced through the reduction of uranium halides with alkali or alkaline earth metals. Uranium metal can also be made through electrolysis of KUF5 or UF4, dissolved in a molten calcium chloride (CaCl2) and sodium chloride (NaCl). Very pure uranium can be produced through the thermal decomposition of uranium halides on a hot filament. 

Cascades of gas centrifuges are used to enrich uranium ore to concentrate its fissionable isotopes. Enrichment of uranium ore through isotope separation to concentrate the fissionable uranium-235 is needed for use in nuclear power plants and nuclear weapons. A majority of neutrons released by a fissioning atom of uranium-235 must impact other uranium-235 atoms to sustain the nuclear chain reaction needed for these applications. The concentration and amount of uranium-235 needed to achieve this is called a 'critical mass.'

To be considered 'enriched' the uranium-235 fraction has to be increased to significantly greater than its concentration in naturally-occurring uranium. Enriched uranium typically has a uranium-235 concentration of between 3 and 5 percent.

The gas centrifuge process, where gaseous uranium hexafluoride (UF6) is separated by weight using high-speed centrifuges, has become the cheapest and leading enrichment process (lighter UF6 concentrates in the center of the centrifuge).   

The most common forms of uranium oxide are triuranium octaoxide (U3O8) and the aforementioned UO2.  Both oxide forms are solids that have low solubility in water and are relatively stable over a wide range of environmental conditions

Triuranium octaoxide is (depending on conditions) the most stable compound of uranium and is the form most commonly found in nature. Uranium dioxide is the form in which uranium is most commonly used as a nuclear reactor fuel. At ambient temperatures, UO2 will gradually convert to U3O8. Because of their stability, uranium oxides are generally considered the preferred chemical form for storage or disposal

At room temperatures, UF6 has a high vapor pressure, making it useful in the gaseous diffusion process to separate highly valuable uranium-235 from the far more common uranium-238 isotope. This compound can be prepared from uranium dioxide and uranium hydride by the following process:

UO2 + 4HF + heat (500 °C) → UF4 + 2H2O

UF4 + F2 + heat (350°) → UF6

The resulting UF6 white solid is highly reactive (by fluorination), easily sublimes (emitting a nearly perfect gas vapor), and is the most volatile compound of uranium known to exist.   

The main use of uranium in the civilian sector is to fuel commercial nuclear power plants; by the time it is completely fissioned, one kilogram of uranium can theoretically produce about 20 trillion joules of energy (20 × 1012 joules); as much electricity as 1500 metric ton of coal.

The south Basin Uranium deposits 

In the south of the Basin, the Tummalapalle belt with low-grade strata-bound carbonate uranium mineralisation is 160 km long, and appears increasingly prospective – AMD reports 37,000 tU in 15 km of it and over 100,000 tU overall, extending down dip to 1000 metres. Some secondary mineralisation is reported in the Srisailam sub-basin.

In August 2007 the government approved a new US$ 270 million underground mine and mill at Tummalapalle near Pulivendula in Kadapa district of Andhra Pradesh, 300 km south of Hyderabad. Its resources have been revised upwards by AMD to 71,690 tU (March 2014) and its cost to Rs 19 billion ($430 million), and to the end of 2012 expenditure was Rs 11 billion ($202 million). First commercial production was in June 2012, using an innovative pressurised alkaline leaching process (this being the first time alkaline leaching is used in India). Production is expected to reach 220 tU/yr as sodium diuranate, and in 2013 mill capacity was being doubled at a cost of Rs 8 billion ($147 million). An expansion of or from the Tummalapalle project is the Kanampalle U project, with 38,000 tU reserves. Further southern mineralisation near Tummalapalle are Motuntulapalle, Muthanapalle, and Rachakuntapalle 

In Karnataka, to the west of north Cuddapah Basin, UCIL is planning a small underground uranium mine in the Bhima basin at Gogi in Gulbarga area from 2014, after undertaking a feasibility study, and getting central government approval in mid-2011, state approval in November 2011 and explicit state support in June 2012. A portable mill is planned for Diggi or Saidpur nearby, using conventional alkaline leaching. Total cost is about $135 million. Resources are 4250 tU at 0.1% (seen as relatively high-grade) including 2600 tU reserves, sufficient for 15 years mine life, at 127 tU/yr, from fracture/fault-controlled uranium mineralisation. UCIL plans also to utilise the uranium deposits in the Bhima belt from Sedam in Gulbarga to Muddebihal in Bijapur.

India's Uranium mines and mills

State, district	Mine	Mill	Operating from	tU per year
Jharkhand	Jaduguda	Jaduguda	1967 (mine) 1968 (mill)	200 total from mill
	Bhatin	Jaduguda	1967
	Narwapahar	Jaduguda	1995
	Bagjata	Jaduguda	2008
Jharkhand, East Singhbum dist.	Turamdih	Turamdih	2003 (u/g mine) 2008 (mill)	190 total from mill
	Banduhurang	Turamdih	2007 (open pit)
	Mohuldih	Turamdih	2012
Andhra Pradesh, Kadapa/YSR district	Tummalapalle	Tummalapalle	2012 2015 (mill)	220 increasing to 330
Andhra Pradesh, Kadapa/YSR district	Tummalapalle	Kanampalle?	2017?
Telengana, Nalgonda district	Lambapur-Peddagattu	Seripally/Mallapuram	2024? (open pit + 3 u/g)	130
Karnataka, Yadgir (Gulbarga) district	Gogi	Diggi/Saidapur	2020? (underground)	130
Meghalaya, West Khasi Hills district	Kylleng-Pyndeng-Sohiong-Mawthabah (KPM), (Domiasiat), Wakhyn	Mawthabah	2022? (open pit)	340

Acid Leaching

Acid leaching has the advantage of being more effective with difficult ores, requiring lower temperatures and leaching times compared to alkaline solutions. It also requires less pretreatment than alkaline leaching, most notably because the particle size from the grinding process does not need to be as small.  Acid leaching is sometimes also referred to as heap leaching because the leaching process can be performed on large "heaps" of uranium ore that have been collected from mines. The chemistry of the leaching process revolves around oxidation of the uranium compounds, which is typically achieved using manganese dioxide (MnO2), sodium chlorate (NaClO3), and Fe(II) salts.   Sulfuric acid is typically used due to the solubility of uranyl sulfate complexes.  The reaction is typically performed at slightly elevated temperatures (~60C) and can often release H2, H2S, and CO2 gases during the process.  The uranium, which typically begins in the tetravalent state, goes through a series of reactions, eventually leading to the formation of the desired complex, [UO2(SO4)3]4-. While the solubility of this complex makes sulfuric acid a desirable leaching agent, nitric and hydrochloric acid can also be used, but are typically not due to their higher cost and corrosiveness.

Alkaline Leaching

While both acidic and alkaline leaching agents are used, alkaline leaching has some significant advantages. Alkaline solutions tend to be more selective to uranium minerals, which means the solution will contain fewer impurities. Consequently, the uranium oxide (commonly called "yellow cake") can be directly precipitated without purification. Furthermore, the solutions are less corrosive and can be recycled without the annoyance of increasing impurity concentrations.  The alkaline leaching process relies on the formation of highly soluble uranyl tricarbonate, UO2(CO3)34-. As in the case of acid leaching, oxidizers are used to maintain the presence of the hexavalent U6+ cation. This can be achieved by simply introducing oxygen into the solution by bubbling air into the solution.  The leaching agents used are sodium bicarbonate and sodium carbonate. This prevents formation of uranyl hydroxide compounds. Due to the slower reactivity of the alkaline solutions, increased pressures and temperatures are sometimes used to accelerate the process.