Abstract: In a process for producing uranium and/or at least one rare earth element selected from the group consisting of cerium, dysprosium, erbium, europium, gadolinium, holmium, lanthanum, lutetium, neodymium, praseodymium, promethium, samarium, scandium, terbium, thulium, ytterbium and yttrium out of an ore, the ore is mixed with sulphuric acid with a concentration of at least 95 wt.-% to a mixture, wherein the mixture is granulated to pellets. The pellets are fed into at least one fluidized bed fluidized by a fluidizing gas for a thermal treatment at temperatures between 200 and 1000° C. The at least one fluidized bed is developed such that it at least partly surrounds a gas supply tube for a gas or a gas mixture fed into the reactor and the gas or gas mixture is used as a heat transfer medium.

Abstract: A substrate treatment apparatus according to an embodiment includes a treatment part, a cyclic path, a heater, and a first injector. The treatment part is supplied with an etchant containing phosphoric acid and a silica deposition suppressor, and brings a substrate having a silicon nitride film on a surface thereof into contact with the etchant to remove the silicon nitride film from the substrate. The cyclic path circulates the etchant in the treatment part. The heater heats the etchant. The first injector is provided on the cyclic path, and injects the silica deposition suppressor into the etchant.

Abstract: A heap of a material to be leached to recover a valuable metal from the material includes an electromagnetic heating system to generate heat in situ in the heap.

Abstract: A process for controlling the dissolution of a metal in an acid bath is described. The metal may comprise aluminum and the acid bath may contain a metal catalyst that causes the metal to dissolve. In order to control the rate of dissolution, the metal can be contacted with a cathodic member. In one embodiment, the process can be completely stopped even without removing the metal from the acid bath. The cathodic member provides anodic protection to the metal. In one embodiment, the cathodic member is made from a nickel-chromium-based alloy.

Abstract: The present invention provides systems and methods capable of improving the efficiency and effectiveness of leaching operations. In one embodiment, the present invention may utilize a coiled tubing directional drilling system capable of treating interior portions of the heap/formation. In one embodiment, the present invention may utilize a system and method capable of capturing real time temperature and resistivity data pertaining to pregnant solution characteristics in the heap/formation. In one embodiment, the present invention may utilize one or more wire line deployed X-Ray Fluorescence (XRF) spectrometers capable of quantitatively measuring concentrations of desired metals in the heap/formation during leaching operations. In one embodiment, the present invention utilizes multiple passes of elemental capture spectroscopy logs acquired at regular time intervals to monitor metal concentrations during leaching operations.

Abstract: The present invention relates to a treatment method of spent uranium catalyst, and more specifically, to a method which can considerably reduce the volume of the spent uranium catalyst to be disposed of and simultaneously minimize secondary wastes that can be generated during the process of treating the spent uranium catalyst.

Abstract: A mineral processing facility is provided that includes a cogen plant to provide electrical energy and waste heat to the facility and an electrochemical acid generation plant to generate, from a salt, a mineral acid for use in recovering valuable metals.

Abstract: Various embodiments provide a process roasting a metal bearing material under oxidizing conditions to produce an oxidized metal bearing material, roasting the oxidized metal bearing material under reducing conditions to produce a roasted metal bearing material, leaching the roasted metal hearing material in a basic medium to yield a pregnant leach solution, conditioning the pregnant leach solution to thrill a preprocessed metal bearing material; and leaching the preprocessed metal bearing material in acid medium.

Abstract: The invention relates to a process which makes it possible to separate together all the actinide(III), (IV), (V) and (VI) entities present in a highly acidic aqueous phase from fission products, in particular lanthanides, also present in this phase by using a solvating extractant in a salting-out medium. Applications: reprocessing of irradiated nuclear fuels, in particular for recovering plutonium, neptunium, americium, curium and possibly uranium, present in the form of traces, in a pooled but selective fashion with regard to lanthanides, from a solution for the dissolution of an irradiated nuclear fuel, downstream of a cycle for the extraction of uranium.

Abstract: The invention relates to a process for collectively separating all the actinides (III), (IV), (V) and (VI) present in a strongly acidic aqueous phase, from the fission products, and in particular from the lanthanides, which are also present in this phase, using two extractants that operate in unconnected chemical fields. Applications: reprocessing of irradiated nuclear fuels, especially to recover plutonium, neptunium, americium, curium and, possibly, uranium present in trace amounts, in a grouped manner but selectively with respect to the lanthanides, from a solution for dissolution of an irradiated nuclear fuel, downstream of a uranium extraction cycle.

Abstract: A method inhibits the formation of zirconium molybdate precipitate in an aqueous solution containing the element molybdenum and the element zirconium by adding a chemical element selected from plutonium, tellurium, antimony and mixtures thereof with the aqueous solution. The method can be used for reprocessing used fuels with the element molybdenum and the element zirconium.

Abstract: Disclosed herein is a uranium ion exchange adsorption method using ultrasound. The method includes placing a slurry obtained by mixing uranium ions, sulfuric acid and an ion exchange resin into a reaction bath, and stirring the slurry in the reaction bath while simultaneously applying ultrasound to the reaction bath to allow the uranium ions to be adsorbed to the ion exchange resin through ion exchange adsorption. The method has an improved ion exchange adsorption rate of the uranium ions.

Abstract: A highly efficient uranium leaching method using ultrasound is disclosed. The uranium leaching method includes preparing black slate powder containing uranium by pulverizing black slate containing uranium, placing the black slate powder and water in a reaction bath, and performing uranium leaching by adding and mixing sulfuric acid and an oxidant with the black slate powder and water to prepare a mixture in the reaction bath while applying ultrasound to the reaction bath. In this method, uranium leaching efficiency can be maximized by adding sulfuric acid to the uranium ore while applying ultrasound thereto.

Abstract: A method inhibits the formation of zirconium molybdate precipitate in an aqueous solution containing the element molybdenum and the element zirconium by adding a chemical element selected from plutonium, tellurium, antimony and mixtures thereof with the aqueous solution. The method can be used for reprocessing used fuels with the element molybdenum and the element zirconium.

Abstract: A method of extracting Te and bismuth oxide and recovering byproduct comprises: leaching raw materials with a Te content of ?1.8% by utilizing a leaching system containing H2SO4, Cl?, Br?, NH4+ and NaClO3, reducing leach solution with SO2 gas by precipitation method after separating impurities from it, washing with concentrated hydrochloric acid to obtain tellurium precipitation (18), purifying to obtain Te with a purity of higher than 99.99%. The filtrate produced is used for extracting Bi2O3 with a purity of higher than 99.99% when Bi content in the raw material is ?2%. Acidic waste solution produced during the process could be returned to the leaching step for recycle.

Abstract: A highly efficient uranium leaching method using ultrasound is disclosed. The uranium leaching method includes preparing black slate powder containing uranium by pulverizing black slate containing uranium, placing the black slate powder and water in a reaction bath, and performing uranium leaching by adding and mixing sulfuric acid and an oxidant with the black slate powder and water to prepare a mixture in the reaction bath while applying ultrasound to the reaction bath. In this method, uranium leaching efficiency can be maximized by adding sulfuric acid to the uranium ore while applying ultrasound thereto.

Abstract: A process for recovering rare earth elements from a composite ore containing rare earth elements that includes a monazite group mineral and an apatite mineral, includes pre-leaching the composite ore with an acid so as to substantially dissolve the apatite mineral into the leach liquor and precipitating rare earth elements from the pre-leach liquor. The residue of the pre-leaching step is subjected to an acid bake treatment, followed by a water leach, to produce a water leach liquor rich in rare earth elements. Impurities including thorium and iron are separated from the water leach liquor by introducing a neutralizing additive to the water leach liquor rich in rare earth elements, and rare earth elements are precipitated from the post-neutralization liquor.

Abstract: A mineral processing facility is provided that includes a cogen plant to provide electrical energy and waste heat to the facility and an electrochemical acid generation plant to generate, from a salt, a mineral acid for use in recovering valuable metals.

Abstract: The invention relates to a process for reprocessing a spent nuclear fuel and for preparing a mixed uranium-plutonium oxide, which process comprises: a) the separation of the uranium and plutonium from the fission products, the americium and the curium that are present in an aqueous nitric solution resulting from the dissolution of the fuel in nitric acid, this step including at least one operation of coextracting the uranium and plutonium from said solution by a solvent phase; b) the partition of the coextracted uranium and plutonium to a first aqueous phase containing plutonium and uranium, and a second aqueous phase containing uranium but no plutonium; c) the purification of the plutonium and uranium that are present in the first aqueous phase; and d) a step of coconverting the plutonium and uranium to a mixed uranium/plutonium oxide. Applications: reprocessing of nuclear fuels based on uranium oxide or on mixed uranium-plutonium oxide.

Abstract: Methods of separating actinides from lanthanides are disclosed. A regio-specific/stereo-specific dithiophosphinic acid having organic moieties is provided in an organic solvent that is then contacted with an acidic medium containing an actinide and a lanthanide. The method can extend to separating actinides from one another. Actinides are extracted as a complex with the dithiophosphinic acid. Separation compositions include an aqueous phase, an organic phase, dithiophosphinic acid, and at least one actinide. The compositions may include additional actinides and/or lanthanides. A method of producing a dithiophosphinic acid comprising at least two organic moieties selected from aromatics and alkyls, each moiety having at least one functional group is also disclosed. A source of sulfur is reacted with a halophosphine. An ammonium salt of the dithiophosphinic acid product is precipitated out of the reaction mixture. The precipitated salt is dissolved in ether.

Abstract: Methods and systems for removing copper minerals from a molybdenite concentrate. One embodiment provides leaching copper from the molybdenite concentrate with a leaching solution comprising ferric chloride, removing molybdenite from the leaching solution, introducing an acid into the leaching solution and introducing O2, O3, or a combination of both, into the leaching solution. A method for regenerating ferric chloride in a leaching solution is also provided. One embodiment provides adding a leaching solution comprising Fe(II) ions, Fe(III) ions, or a combination of both, to a mixture of mineral sulfides, introducing an acid into the leaching solution, and introducing O2, O3, or a combination of both, into the leaching solution.

Abstract: An apparatus for the removal of uranium from a body of material is provided. The apparatus has at least one ultrasonic extractor, having a bottom and a top. The at least one ultrasonic extractor is configured to accept solids at the bottom and acid at the top, and has a mixing screw and at least one source of ultrasonic energy. The mixing screw is configured to transport the solids in a direction countercurrent to the acid in the at least one ultrasonic extractor; and the source of ultrasonic energy is configured to impart ultrasonic energy into the solids and the acid, as the solids and the acid traverse the at least one ultrasonic extractor countercurrently.

Abstract: A method is disclosed for separating trivalent americium from trivalent curium, coming from an aqueous solution containing at least these cations, wherein, at an acid concentration of 0.01 mol/l-0.3 mogl/l, the aqueous solution is brought into contact with an organic solvent containing a bis(aryl)dithiophosphinic acid having the formula (4) where R1=phenyl or naphthyl R2=phenyl or naphthyl, and radicals of R1 and R2 substituted by at least one methyl, ethyl, propyl, isopropyl-, cyano, nitro, or halo substituent, and containing a synergist having the formula (5) where X and/or Y and/or Z is R or RO, wherein R is branched or unbranched alkyl.

Abstract: The invention relates to a process for the chemical beneficiation of raw material containing tantalum-niobium such as wastes, scoria, concentrates and ores.

Abstract: A method of recovering daughter isotopes from a radioisotope mixture. The method comprises providing a radioisotope mixture solution comprising at least one parent isotope. The at least one parent isotope is extracted into an organic phase, which comprises an extractant and a solvent. The organic phase is substantially continuously contacted with an aqueous phase to extract at least one daughter isotope into the aqueous phase. The aqueous phase is separated from the organic phase, such as by using an annular centrifugal contactor. The at least one daughter isotope is purified from the aqueous phase, such as by ion exchange chromatography or extraction chromatography. The at least one daughter isotope may include actinium-225, radium-225, bismuth-213, or mixtures thereof. A liquid-liquid extraction system for recovering at least one daughter isotope from a source material is also disclosed.

Abstract: The invention relates to a process for the chemical beneficiation of raw material containing tantalum-niobium such as wastes, scoria, concentrates and ores.

Abstract: A process for production of titanium concentrate with low contents of radionuclide elements from anatase mechanical concentrates. The process involves calcination in air and reduction with hydrogen or any other reducing gas, both in fluidized bed reactor or rotary kiln, low-intensity magnetic separation of the reduced product, high-intensity magnetic separation of the non-magnetic fraction resulting from the low-intensity magnetic separation, hydrochloric acid leaching of the product of high-intensity magnetic separation, filtering and dewatering of the leached product, high temperature oxidation of the dewatered material, cooling of the oxidized ore, hydrochloric acid leaching of the oxidation product in the presence of sodium fluoride, filtration and drying of the product of the second leaching and high intensity magnetic separation, the non-magnetic fraction of this final magnetic separation becoming the end product.

Abstract: An apparatus and method to remove uranium from a body of material wherein the method includes the steps of depositing the body of solid material in an ultrasonic extractor and depositing an amount of acid in the ultrasonic extractor. The method also provides for the steps of heating a jacket of the ultrasonic extractor, transporting the body of solid material in the ultrasonic extractor and the amount of acid such that the body of solid material and the acid contact each other inside the heated ultrasonic extractor while the ultrasonic extractor provides ultrasonic energy to both the body of solid material and the amount of acid, wherein the amount of acid strips uranium from the body of solid material. The method further provides for collecting the amount of acid and the body of solid material in the ultrasonic extractor in different positions, transporting the amount of acid with the stripped uranium to an extraction mixer settler, and settling uranium product from the extraction mixer settler.

Abstract: The present invention relates generally to a process for recovering copper and other metal values from metal-containing materials using controlled, super-fine grinding and medium temperature pressure leaching. Processes embodying aspects of the present invention may be beneficial for recovering a variety of metals such as copper, gold, silver, nickel, cobalt, molybdenum, rhenium, zinc, uranium, and platinum group metals, from metal-bearing materials, and find particular utility in connection with the extraction of copper from copper sulfide ores and concentrates.

Abstract: A method for recovering and recycling catalyst coated fuel cell membranes includes dissolving the used membranes in water and solvent, heating the dissolved membranes under pressure and separating the components. Active membranes are produced from the recycled materials.

Abstract: The invention relates to a process for the preparation of a product based on a phosphate of at least one element M(IV), for example of thorium and/or of actinide(IV)(s). This process comprises the following stages: a) mixing a solution of thorium(IV) and/or of at least one actinide(IV) with a phosphoric acid solution in amounts such that the molar ratio PO 4 M ? ? ( IV ) ?is from 1.4 to 2, b) heating the mixture of the solutions in a closed container at a temperature of 50 to 250° C. in order to precipitate a product comprising a phosphate of at least one element M chosen from thorium(IV) and actinide(IV)s having a P/M molar ratio of 1.5, and c) separating the precipitated product from the solution. The precipitate can be converted to phosphate/diphosphate of thorium and of actinide(s). The process also applies to the separation of uranyl ions from other cations.

Abstract: A process for reducing a level of radionuclei in a phosphogypsum material having a radioactive contaminant is provided, wherein the process includes the steps of: (a) mixing the phosphogypsum material with an acid decontamination solution to form a mixture; (b) separating the mixture into a liquid stream having a large portion of the radioactive contamination and a solid stream having a residual component of the acid decontamination solution; and (c) removing the residual component of the acid decontamination solution from the solid stream, thus providing a gypsum material that meets or exceeds federal standards for level of radioactive nuclei and providing a process that is environmentally friendly and capable of avoiding the generation of phosphogypsum stacks, as well as eliminating existing or abandoned phosphogypsum stacks, further providing for recovery of the radioactive nuclei if desired.

Abstract: An object of the present invention is to provide a structure of a metal collecting material for use in seawater or river water or industrial wastewater to give the most excellent adsorption efficiency. Another object of the present invention is to provide a process for efficiently separating and recovering valuable metals or noxious metals from a metal collecting material having collected said valuable metals from seawater or noxious metals from river water or wastewater with the least possible waste generation. A metal collector according to an aspect of the present invention is formed by stacking at least partially alternate layers of a fibrous metal collecting material with a spacer for introducing a liquid to be treated into said collecting material, wherein the area of spacer side faces represents 25-75% of the total area of collecting material side faces and spacer side faces in the side faces of the stack.

Abstract: A nuclear fuel dissolution apparatus which comprises a perforated sloping ramp contained within a process chamber for containing solvent for the fuel, a pulsation member which in use creates pulses in solvent in the process chamber, the perforations being designed to direct pulses of solvent along and up the ramp, and a discharge point for fuel hulls disposed at an upper region of the ramp is characterised in that the ramp is made out of flat blades and the perforations of the ramp comprise inclined slits formed between the blades.

Abstract: A method for removing a wide variety of radioactive contaminants from a contaminated surface sufficient for the surface to be classified as a low-level waste or as free release. Contaminated surfaces may be classified as Class C, Class B, Class A, or high-level radioactive wastes prior to treatment. An aqueous solution having a wetting agent and a complex substituted keto-amine is provided. The aqueous solution is left on the surface for a time sufficient to remove the radioactive contaminants into the aqueous solution which is then removed. Depending on the type and condition of the surface, an acid may be added to the aqueous solution to aid in the contaminant removal process. However, typical metals surfaces may often be treated effectively without the use of concentrated acids or with dilute concentrations of such acids. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a variety of previously contaminated surfaces.

Abstract: An object of the present invention is to provide a method for preventing concentration of a radioactive substance in a generated extraction residue in a method of producing tantalum, niobium, or a similar substance including collecting and refining a raw material containing the substances through a fluoridation process by use of a hydrofluoric acid-containing solution. The object can be attained by employing an ingredient-regulated raw material prepared from an ore or a concentrate and, as an additive, a substance insoluble to hydrofluoric acid or a mixed acid containing hydrofluoric acid as an essential component; or by increasing the amount of the extraction residue through addition of the insoluble substance to a solvent during the fluoridation process, to thereby reduce the relative radioactive substance content to an arbitrary value.

Abstract: The present invention relates to a process to dissolve plutonium or a plutonium alloy, by placing it in contact with an aqueous dissolution mixture, wherein said dissolution mixture comprises nitric acid, a carboxylic acid with complexing properties with respect to plutonium, and a compound comprising at least one —NH2 radical such as urea. The invention also relates to a process to convert plutonium or a plutonium alloy into plutonium oxide and to manufacture nuclear fuel from said oxide. The invention particularly applies to the dismantling of plutonium contained in nuclear weapons with a view to its use in civilian nuclear reactors, particularly in the form of MOX fuel.

Abstract: A method for removing radioactive contaminants from a given surface. An aqueous solution having a wetting agent and a complex substituted keto-amine is provided. The solution is left on the surface for a time sufficient to remove the radioactive contaminants into the aqueous solution which is then removed. Depending on the type and condition of the surface, a concentrated acid may be added to the aqueous solution to aid in the contaminant removal process. In such a case, a pH of less than 3.0, and preferably less than 1.5 is maintained. If a concentrated acid is used, the acidic solution containing radioactive contaminants is preferably neutralized by an alkaline material to a pH of between 5.5 and 9.0. Removal of thorium contamination from railcars is one useful application of the invention. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a previously contaminated surface.

Abstract: A method for removing radioactive contaminants from a given surface. An acidic solution having a pH of less than 3.0, and preferably less than 1.5, of a complex substituted keto-amine, and a mixture of a saturated and unsaturated lower alcohol (e.g. isopropyl alcohol and propargyl alcohol), and concentrated acid is applied to a contaminated surface. The solution is left on said surface for a time sufficient to remove the radioactive contaminants into the acidic solution and then removed. The acidic solution containing the radioactive contaminants is preferably neuralized by an alkaline material to a pH of between 5.5 and 9.0. Removal of thorium contamination from railcars is one useful application of the invention. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a previously contaminated surface.

Abstract: Proposed is an improvement in a multi-stage solvent extraction process of a metal value such as rare earth elements from an acidic feed solution containing the metal value in the form of a water-soluble salt by bringing the aqueous solution into contact with a water-immiscible organic extractant solution containing an extractant compound which releases an acid when combined with the metal ions.

Abstract: A method for recovering nitrate ions as nitric acid from nuclear industry effluents by thermally decomposing the nitrate ions in solution, and recovering the NOx vapors generated by the heat treatment in an aqueous medium. The resulting nitric acid may be recycled in the nuclear industry.

Abstract: Radioactively contaminated material is cleaned by contacting the material with a decontaminating liquid comprising an aqueous solution of nitric acid containing an NOx generating agent. The NOx generating agent may be a nitrite, for example, sodium nitrite, or a ferrous metal. The material to be cleaned may comprise a plastics material contaminated with uranium or other actinides. Cleaning is effected by placing the material in a rotatable, apertured vessel in which the material is subjected to a leaching cycle by contact with the decontaminating liquid and then a washing cycle in which the material is contacted with a washing liquid.

Abstract: A process for the decontamination of radioactively contaminated metal which comprises contacting the metal with a decontamination reagent solution containing an organic acid and an oxidising agent, allowing said solution to react with the contaminated metal at a pH of up to 4.5, treating the resulting solution to cause substantially complete precipitation of dissolved metal together with radionuclides and separating precipitated material, containing radioactive contaminants, from said solution.

Abstract: A process for oxidation of ferrous ions in solution, and more particularly a process for improved base metal and/or uranium leaching from ores, concentrates or tailings using ferric ion as an oxidizing agent. A reaction vessel (10) holds a ferrous ion-containing solution, for example, a copper sulphide leach slurry or concentrate. An agitator (12) may be provided to promote leaching of the base metal into solution. Some of the ferrous ion-containing solution is drawn off from the reaction vessel (10) and pumped through an in-line mixer (14) via a feed pump (16). Oxygen is injected into the reactor (14) to facilitate oxidation of the ferrous sulphate to form ferric sulphate. The ferric ion-containing solution is then recirculated back to the reaction vessel (10) where the ferric ions are reused in the dissolution of copper sulphide into soluble copper sulphate.

Abstract: Trivalent actinides can be separated from trivalent lanthanides in aqueous solutions at H.sup.+ concentrations of 2 mol/l to 0.001 mol/liter by extraction with bis(aryl)dithiophosphinic acid and a synergist such as TBP, TOPO and TBPO in an organic solvent. A high separation efficiency results. The method is applicable to high-level liquid waste from reactors and nuclear-material processing plants.

Abstract: The present invention relates to methods for digesting diphosphonic acid substituted cation exchange resins that have become loaded with actinides, rare earth metals, or heavy metals, in a way that allows for downstream chromatographic analysis of the adsorbed species without damage to or inadequate elution from the downstream chromatographic resins. The methods of the present invention involve contacting the loaded diphosphonic acid resin with concentrated oxidizing acid in a closed vessel, and irradiating this mixture with microwave radiation. This efficiently increases the temperature of the mixture to a level suitable for digestion of the resin without the use of dehydrating acids that can damage downstream analytical resins. In order to ensure more complete digestion, the irradiated mixture can be mixed with hydrogen peroxide or other oxidant, and reirradiated with microwave radiation.

Abstract: A process for facilitating the removal of impurities e.g. radionuclides, such as uranium and thorium, and/or one or more of their radionuclide daughters, from titaniferous material includes contacting the titaniferous material with one or more reagents at an elevated temperature selected to enhance the accessibility of at least one of the radionuclide daughters in the titaniferous material. The reagent(s) may be a glass forming reagent and is selected to form a phase at the elevated temperature which disperses onto the surfaces of the titaniferous material and incorporates the radionuclides and one or more radionuclide daughter. The titaniferous material may be, e.g., ilmenite, reduced ilmenite, altered ilmenite or synthetic rutile.

Abstract: The process for converting feed materials of high mineral content containing primary metal values and fluorine values to the primary metal or useful compounds thereof and to fluorine values or useful compounds thereof, wherein the feed materials constitutes a difficultly soluble matrix, the process having the steps of contacting the feed materials in a reactor with a humidified, gaseous system at from about 200.degree. C. to about 1600.degree. C., the contacting being carried out such as to convert the primary metal values to oxide residues at commercially acceptable rates and to evolve gaseous fluoride from the feed, digesting said oxide residues in an acidic digest medium and separating the primary metal values from the resulting digest liquor and from other components of the residues.

Abstract: Magnesium fluoride slag contaminated with metallic uranium or uranium compounds is treated by digestion with potassium hydroxide to soluble potassium fluoride and insoluble magnesium hydroxide. The solid and liquid phases are then separated and the solids phase is dissolved in acid to form a solution of metal salts. The liquid phase is treated with lime to precipitate calcium fluoride and convert the potassium back to potassium hydroxide for recycle to the digestion reaction. The metal salts are separated to remove the uranium salt from the magnesium salt. The result is an efficient removal of uranium contamination from the magnesium, the recovery of magnesium in a manner that permits efficient reuse or safe disposal, and the production of calcium fluoride useful for a variety of purposes.

Abstract: A process for facilitating removal of one or more impurities from titaniferous material containing the impurities in a form which is highly soluble in acid. The material is in turn leached with a sulphuric acid solution and with a hydrochloric acid solution in either order. To enhance the susceptibility of the impurities to removal, the hydrochloric acid leach is augmented by one or more of the following: (i) addition of an effective amount of an added chloride salt; (ii) pre-treatment of the titaniferous material with a solution of an effective amount of a carbonate salt, preferably an alkali metal or alkaline earth metal carbonate; and/or (iii) pre-treatment of the titaniferous material with a solution of an effective amount of a hydroxide, preferably an alkali metal or alkaline earth metal hydroxide.