This page contains information based on the Encyclopedia Britannica's Uranium Processing trail and the processes used by British Nuclear Fuels PLC for the conversion of uranium ore concentrates to uranium metal and uranium hexafluoride¹.

I decided to focus on the acid leaching and solvent extraction pathway, and thus the information here will focus on these methods.

Also keep in mind the objective of this document is to adapt real processes into a realistic yet implementable minecraft mod. Simplification, creative liberty, selective use of reality will be employed and this should not be considered an accurate reference to the actual uranium refinement processes.

• Add DEHPA page
• Add TBP page

• Game Implementation Plan
• Scientific Background
  • Mining and Concentrating
  • Roasting
  • Leaching
  • Treatment of Leachates
  • Precipitation of Yellowcake
  • Refining of Yellowcake
  • Conversion and isotopic enrichment
  • Reactor-grade uranium
  • Depleted uranium
  • Oxide Fuels

• 2 ore levels:
  • Near surface: TopSolidRangeConfig from level 48 to 60.
  • Deep: DepthAverageConfig with base line at level 15, spread 5.

Basic process outline: Ore Crushing -> Initial Refinement (leaching) -> Yellowcake Precipitation -> Purification-> Uranium Dioxide

• Process begins by pulverizing ore (-> Crushed Uraninite, TExp pulverizer, impl. Grinder)
• Initial Refinement:
  • Roasting
  • Leaching with H₂SO₄ - agitate 4h to 48h at ambient temperature, pH 1.5.
    5 kilograms of manganese dioxide or 1.5 kilograms of sodium chlorate per ton of ore used to force hexavalent oxidation state.
  • Clarification of leachate, separate clear pregnant leaching solution.
  • Solvent extraction with DEHPA (di(2-ethylhexyl)phosphoric acid).
  • Precipitation with aqueous ammonia (NH₃).
• Yellowcake
• Basic Purification:
  • The yellowcake is ignited, driving off the ammonia and oxidizing the uranium to UO₃ or U₃O₈.
    This method provides a less pure mix of uranium oxides.
• Industrial/Advanced Purification:
  • Nitric acid (HNO₃)
  • Solvent extraction, 20% v/v tributyl phosphate in odorless kerosene.
  • Evaporation: pure uranyl nitrate @ 11% w/v U -> 110% w/v U molten concentrate.
  • Thermal decomposition of the concentrate (300-350°C) -> Uranium Trioxide, NO_x, oxygen and steam.
    This process is easily automatable and provides high purity uranium oxides.
  • Nitrous fumes can be used to recover nitric acid.
• Reduction with hydrogen at 480°C to UO₂
• Next: Fluoride Production TBD

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.

Uranium ores typically contain only a small amount of uranium-bearing minerals, and these are not amenable to smelting by direct pyrometallurgical methods; instead, hydrometallurgical procedures must be used to extract and purify the uranium values. Physical concentration would greatly reduce the load on hydrometallurgical processing circuits, but none of the conventional beneficiation methods typically employed in mineral processing — e.g., gravity, flotation, electrostatics, and even hand sorting — are generally applicable to uranium ores. With few exceptions, concentration methods result in excessive loss of uranium to tailings.

The hydrometallurgical processing of uranium ores is frequently preceded by a high-temperature calcination step. Roasting dehydrates the clay content of many ores, removes carbonaceous materials, oxidizes sulfur compounds to innocuous sulfates, and oxidizes any other reductants that may interfere in subsequent leaching operations.

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 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-.

Prior to further processing, solutions resulting from leaching must be clarified. Large-scale separation of clays and other ore slimes is accomplished through the use of effective flocculants, including polyacrylamides, guar gum, and animal glue.

Uranium can 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.

Prior to final purification, uranium present in acidic solutions produced by the solvent-extraction processes described above 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.

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).

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 (described below, see Conversion to plutonium). 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 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. (In an often-used modification of the Ames process, calcium metal is substituted for magnesium.) Because the vapour pressure of 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, at which point an exothermic (heat-producing) 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 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 (see below Chemical compounds: Oxide fuels). 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.

Uranium metal intended for use in production reactors to produce plutonium-239 is rolled into round billets typically 23 centimetres (9 inches) in diameter and 51 centimetres long. Metallic uranium fuel elements for power reactors are prepared by hot extrusion of the uranium into tubing made of Zircaloy, a corrosion-resistant alloy of zirconium and tin. Uranium fuel elements can also be clad in alloys of magnesium and aluminum.

Uranium reacts with a large variety of other metals to form intermetallic compounds, solid solutions, or (in a few instances) true alloys. Many of these systems have been exploited to prepare reactor fuels that possess increased resistance to in-reactor corrosion and radiation damage as well as greater mechanical strength than pure uranium metal. Examples include low-enrichment uranium-molybdenum and uranium-aluminum alloys.

Even though uranium and plutonium are completely miscible, the plutonium-uranium system is not suitable for nuclear applications. As described above, uranium exists in three crystal structures between ambient temperature and its melting point of 1,132° C (2,070° F). Plutonium metal undergoes five phase transformations below its melting temperature of 640° C (1,183° F). Transformation from one phase to the next occurs no matter what the concentration of either element, thereby preventing use of plutonium-uranium alloys over a major temperature range. However, the addition of zirconium—for example, in a fuel containing 20 percent plutonium, 10 percent zirconium, and 70 percent uranium—can yield a metallic system completely adaptable to reactor use.

Uranium fuel elements can be sheathed in a metallic blanket containing, for example, 10 percent zirconium and 90 percent uranium depleted of its uranium-235 content. The depleted uranium, consisting almost completely of uranium-238, captures neutrons that are emitted in the fission of the fuel elements, thus producing (or “breeding”) plutonium-239 simultaneous with the generation of nuclear power.

Certain alloys of depleted uranium are also used in armour for tanks and other military vehicles. Because of its very high density, uranium metal is well suited for this purpose as well as for armour-piercing projectiles.

Pellets made of low-enrichment UO₂ are universally employed as fuel in commercial light-water reactors that produce electrical energy. The pellets are made by blending appropriate quantities of enriched and natural or depleted UO₂ powders, mechanically compacting them, adding an organic binder, pressing into pellets, heating to burn off the binder, and finally sintering at high temperature to 95 percent theoretical density. Fuel pins are fabricated by loading the pellets into a Zircaloy tube.

Similar procedures are employed to fabricate mixed uranium-plutonium dioxide (MOX) pellets for use in fast-neutron breeder reactors. Unirradiated MOX fuel typically contains 20 to 35 percent plutonium dioxide.

1. Naylor, A., Ellis, J.F. and Watson, R.H. (1986), Chemical aspects of nuclear fuel fabrication processes. J. Chem. Technol. Biotechnol., 36: 162-168. https://doi.org/10.1002/jctb.280360403