Abstract: A method of leaching a material from a cavity having a closed or a partially closed top. The method consists of subjecting the liquid to a number of pressure cycles, each pressure cycle involving a decrease in pressure to cause boiling of the liquid, followed by a rise in pressure to inhibit the boiling. The method may include the step of heating the liquid to a temperature near to its boiling point.

Abstract: The invention concerns the recovery of non-ferrous metals, such as gold, uranium or the like from iron oxide containing calcines which have the non-ferrous metal present in solid solution and/or encapsulated within the iron oxide. The calcine is reacted, while stirring vigorously, with sulphuric acid or another strong inorganic acid to cause the iron to form the ferric salt. The material obtained is mixed with water and the liquid and solid phases are separated from each other. The non-ferrous metal is then obtained from at least one of these phases by leaching, or the like.

Abstract: Certain metals, and especially uranium, are recovered from an organic solvent by stripping with an aqueous solution containing polymeric phosphates under conditions in which the polymeric phosphate does not quickly hydrolize and metal preferentially dissolves in the aqueous solution. The metal may then be recovered from the aqueous solution by hydrolizing the polymeric phosphates and either precipitating the metal or reextracting the metal at higher concentrations with an organic solvent containing extractants.

Abstract: A process for the recovery of uranium values from uranium-bearing material by sulphuric acid leaching when the material also contains siliceous matter which is at least partly soluble in sulphuric acid. The process comprises leaching the material with sulphuric acid in an aqueous slurry under oxidizing conditions at a temperature within the range of from about 65.degree. C. to about 200.degree. C. and at elevated pressure, with sulphuric acid being present in a quantity only slightly in excess of the amount required to dissolve the uranium and any other element whose dissolution is desired, to produce an insoluble residue containing siliceous matter and a leach solution containing dissolved values of uranium and any other desired element.

Abstract: There is disclosed a process for extracting uranium values from a uranium-containing carbonaceous ore such as lignite. This process includes the step of treating the ore with an aqueous leach solution containing a water-soluble ferric salt.

Abstract: There is provided a method for simultaneously carrying out pulverization of ores and their leaching or extraction within a vertical abrasion mill.

Abstract: An improvement in the process for the recovery of uranium from a phosphoric acid solution derived from the acidification of uraniferous phosphate ores including the steps of reducing the uranium in the phosphoric acid solution to the uranous form and contacting the resulting solution with a liquid-liquid solvent extraction agent which is a mixture of mono- and di-(alkyl-phenyl) esters of orthophosphoric acid dissolved in an inert organic diluent, to effect recovery of the uranium with the liquid-liquid solvent extraction agent, the improvement being maintaining the proper ratio of the mono and di-components of the liquid-liquid solvent extraction mixture which comprises removing dissolved and entrained uranium process organics from the raffinate resulting from said liquid-liquid solvent extraction step and recycling the removed organics to said liquid-liquid solvent extraction step.

Abstract: Improvement in the process for recovering uranium from wet-process phosphoric acid solution derived from the acidulation of uraniferous phosphate ores by the use of two ion exchange liquid-liquid solvent extraction circuits in which in the first circuit (a) the uranium is reduced to the uranous form; (b) the uranous uranium is recovered by liquid-liquid solvent extraction using a mixture of mono- and di-(alkyl-phenyl) esters of orthophosphoric acid as the ion exchange agent; and (c) the uranium oxidatively stripped from the agent with phosphoric acid containing an oxidizing agent to convert uranous to uranyl ions, and in the second circuit (d) recovering the uranyl uranium from the strip solution by liquid-liquid solvent extraction using di(2-ethylhexyl)phosphoric acid in the presence of trioctylphosphine oxide as a synergist; (e) scrubbing the uranium loaded agent with water; (f) stripping the loaded agent with ammonium carbonate, and (g) calcining the formed ammonium uranyl carbonate to uranium oxide, the i

Abstract: Production of UO.sub.2 nuclear fuel pellets containing PuO.sub.2 in set amounts, which pellets are soluble in nitric acid and suitable for reprocessing without the aid of a special additive. Uranium oxide powder is mixed with 15 to 50% by weight of the mixture of plutonium oxide powder; the mixture milled and pressed into pellets. The pellets are sintered, comminuted, pressed to form pellets and the pellets comminuted to free-flowing granules. The free-flowing granules are mixed with UO.sub.2 granules to obtain a desired UO.sub.2 /PuO.sub.2 ratio, and the mixture pressed into pellets and sintered to form UO.sub.2 nuclear fuel pellets containing PuO.sub.2 soluble in nitric acid.

Abstract: Solid carbonaceous fossil fuels such as coal, lignite and peat are treated with an aqueous medium containing a novel catalyst to remove undesirable constituents and produce valuable products. The catalyst is prepared by steps including admixing a water soluble alkali metal silicate with an aqueous medium containing carefully controlled amounts of dissolved water soluble substances which are sources of calcium ion and magnesium ion, reacting the same to produce an aqueous colloidal suspension of the reaction product, admixing a micelle-forming surfactant with the aqueous medium, and agitating the aqueous medium containing the colloidal particles and surfactant to form catalyst-containing micelles. In one variant, particles of a fossil fuel containing metal values are treated with an aqueous medium in the presence of the catalyst, and thereafter the treated particles are separated from the aqueous medium and extracted with an aqueous leach solution in which the metal values are soluble.

Abstract: A process for selectively removing tetravalent and/or hexavalent uranium ions from a solution by using a chelate-forming cation exchanger. The solution may be prepared by contacting uranium bearing ore with either an acid or a dilute carbonate or bicarbonate solution. To facilitate removal of the uranium ions, the solution may have a pH value greater than 1.0 and a redox potential of from -50 to +250 mV.

Abstract: Process for recovering compounds of enriched nuclear fuel from scrap materials is disclosed. The process yields an acid solution with the nuclear fuel dissolved therein. Except for spent filter media, the scrap materials are calcined to produce an oxidized material, which is reduced to a particulate material and leached in an acid bath that contains some recycled acid to yield an acid solution of the nuclear fuel material. The scrap materials comprised of spent filter media are mulched and contacted with the acid solution. The insoluble materials are separated from the acid solution, and at least a portion of the acid solution is recycled to the leaching step, while the remainder of the acid solution is collected for subsequent treatment to recover the dissolved nuclear fuel material. Apparatus for performing the foregoing process is also disclosed.

Abstract: In a process according to the present invention uranium is extracted into solution from its ore by leaching with an aqueous solution containing peroxomonosulphuric acid, the peroxoacid oxidizing the uranium through to its hexavalent state. Preferably the leaching is carried out at a temperature in the range of 50.degree. to 100.degree. C. The leach liquor can initially contain additional amounts of sulphuric acid or merely that present by virtue of the method of making the peroxomonosulphuric acid. In a preferred method of operation, the peroxoacid is introduced progressively into the leach liquor during the course of the leaching so as to maintain an electrochemical potential in the range of 450 to 650 mV.By use of the process, uranium is cleanly extracted into solution.

Abstract: A process for the continuous leaching of ores and an apparatus for practicing this continuous leaching process are disclosed. According to this process, a plurality of unit layers composed of a pulverized uranium ore or other ore are continuously laminated while a minimum necessary amount of an acid, alkali or organic solvent (hereinafter referred to as "solvent") is uniformly sprinkled on the flat surfaces of these unit layers. In the state where the concentration of the solvent mixed into the ore is maintained at a high level, the heat generated by exothermic reaction caused by contact among the solvent, ore and water is effectively stored and used for thermally curing the ore. According to this process, the speed of extraction of the intended metal component can be increased, the leaching time shortened and the filtration characteristics improved, whereby a highly concentrated pregnant liquor can be recovered at a high efficiency.

Abstract: The present invention relates to an improved method for the solution mining of a mineral from a subterranean formation. More specifically, the invention relates to an improved method which enhances significantly the recovery of the mineral from a subterranean formation via solution mining by reducing the oxidant consuming reactions which compete with the mineral for the oxidant injected therein.

Abstract: Irradiated nuclear fuel is dissolved in nitric acid under a carbon dioxide atmosphere which is treated to remove iodine, krypton and xenon and recycled.

Abstract: The oxidation of uranium is catalyzed with water-soluble copper or zinc chelates. An aqueous solution is prepared containing carbonate or bicarbonate ion, an oxidizer such as hydrogen peroxide or oxygen, and a water-soluble copper or zinc chelate which has at least two amino, carboxylic acid, or thiocarboxylic acid donor groups. The solution is pumped into an underground deposit of uranium. The water-soluble +4 state uranium is oxidized to the water-soluble +6 state which forms a complex with the carbonate ion. The solution is pumped to the surface and is recycled after precipitation of the uranium.

Abstract: A process for recovering compounds of enriched nuclear fuel from scrap materials is disclosed. The process yields an acid solution with the nuclear fuel dissolved therein. Except for spent filter media, the scrap materials are calcined to produce an oxidized material, which is reduced to a particulate material and leached in an acid bath that contains some recycled acid to yield an acid solution of the nuclear fuel material. The scrap materials comprised of spent filter media are mulched and contacted with the acid solution. The insoluble materials are separated from the acid solution, and at least a portion of the acid solution is recycled to the leaching step, while the remainder of the acid solution is collected for subsequent treatment to recover the dissolved nuclear fuel material. Apparatus for performing the foregoing process is also disclosed.

Abstract: A process for the extraction of uranium from ores which is an improvement in the known "dry attack" treatment of uranium ores to an economical yet high rate of extraction. The process concerns solubilization of the uranium contained in the ores and consists of carrying out digestion of the mixtures of ore and reagent, with the possible addition of oxidizing agents. The process is carried out on an improved trough type of conveyor belt at predetermined conditioning of temperatures and humidity.

Abstract: The addition of a normally insoluble fluoride to a reaction being a chemical treatment process provides a surprisingly improved rate is dissolution in many cases. Specifically, where the fluoride is added to a reaction using hydrochloric acid or ferric chloride, the yield is more rapid than has previously been shown in the art. The invention is most applicable where the stability constant of the fluoride or fluoride complex of the element to be recovered is higher than that of the cation with which the fluoride is originally associated.

Abstract: PuO.sub.2 -containing solids, particularly residues from incomplete HNO.sub.3 dissolution of irradiated nuclear fuels, are dissolved in aqueous HI. The resulting solution is evaporated to dryness and the solids are dissolved in HNO.sub.3 for further chemical reprocessing. Alternatively, the HI solution containing dissolved Pu values, can be contacted with a cation exchange resin causing the Pu values to load the resin. The Pu values are selectively eluted from the resin with more concentrated HI.

Abstract: A method of extracting uranium from its ores, by mixing ground ore with sulphuric acid to give a moist ore containing a sulphuric acid concentration of less than 4N, without forming a continuous liquid phase, and curing the moist ore at 50.degree. to 100.degree. C while passing an oxidizing gas through it.

Abstract: The invention is a method for working-up shale while recovering metals therefrom and the sulphur content thereof. The method comprises crushing the shale and leaching and roasting the shale. The sulphur content is oxidized and converted into sulphuric acid, concentrated sulphur dioxide or sulphur. The leaching operation is effected with sulphuric acid. The leached metals are recovered in a manner known per se and the residual solution is passed to the roasting stage.

Abstract: A method is disclosed for the manufacture of phosphoric acid directly from phosphate rock wherein the crushed phosphate rock is mixed with dilute phosphoric acid to form a slurry and the slurry is then heated to produce calcium monophosphate. Thereafter, oxalic acid is added to the slurry to precipitate the calcium therein as calcium oxalate which is separated. The liquid resulting therefrom which contains the phosphoric acid from the rock is then treated conventionally to recover the phosphoric acid therein.

Abstract: Uranium is recovered from refractory carbonaceous-sulfidic ores or concentrates by leaching with nitric acid, followed by sorption on an ion exchange resin. In addition, the nitric acid leach solution is regenerated and recycled.

Abstract: Minerals, for example, zirconium minerals, are subjected to a chemical process of comminution, in particular with the aid of strong acids. These acids dissolve the carbonatic and silicious cements which hold together the crystalline forms of different mineral species from one another, e.g., magnetic and non-magnetic species to render them amenable to further separation in a subsequent ore dressing stage, e.g., by high intensity magnetic separation. This process is operative as these cements dissolve in acid more readily than most of the other compounds of the ores. This separation was found to take place more effectively than by the usual methods of mechanical grinding and without the disadvantage of an unfavorable particle size distribution which results from grinding. It was found that the same process conditions also resulted in a particularly favorable removal of naturally occurring radioactive contaminants from the minerals, in particuar thorium and uranium.

Abstract: A technique is provided for leaching particulate ores for recovering metal values. Ore is crushed to less than about one inch with at least 25% being greater than 4 mesh. The crushed ore is mixed with a small amount of acid and the moistened mixture is kept in a curling pile for about two days for dehydration and induration of the ore. The induration is sufficient that the ore can sustain the weight of a thin layer of about one-half to one meter thick without appreciable compacting. The indurated ore is spread on a permeable substrate in a thin layer about one-half to one meter thick. A leaching solution is then percolated through this thin layer to remove the metal values. The initial effluent leach liquor has high metal values and low acid content and is passed to a solvent extraction system where the metal values are recovered. The solvent extraction raffinate, which is high in acid, is recirculated as a leach liquid.

Abstract: A method of regenerating fuel and/or breeder elements with thorium as brer material for nuclear reactors. The fuel and/or breeder material is dissolved in an acid such as nitric acid, sulfuric acid or the like and the thus obtained solution is subjected to a liquid-liquid reaction with an extraction substance such as tributyl phosphate dissolved in an organic solvent. The fuel material and breeder material are enriched in the organic phase and the fission products remain in the aqueous phase, whereafter the fuel material and/or breeder material are converted into oxides, carbides etc. Prior to the above mentioned liquid-liquid extraction, the acid solution containing the fuel material and/or breeder material is passed over a sorption column in such a way that the number of mols of adsorbed .sup.233 Pa per time unit equals the number of mols of .sup.233 Pa disintegrating per time unit in the sorption column to form .sup.233 U.

Abstract: The present invention provides an improved dissolution process for ZrO.sub.2 -UO.sub.2 -CaO-type pressurized water reactor fuels. The zirconium cladding is dissolved with hydrofluoric acid, immersing the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers in the resulting zirconium-dissolver-product in the dissolver vessel, and nitric acid is added to the dissolver vessel to facilitate dissolution of the uranium from the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers.

Abstract: A chemical digestion for treatment of low level combustible nuclear solid waste material is provided and comprises reacting the solid waste material with concentrated sulfuric acid at a temperature within the range of 230.degree.-300.degree.C and simultaneously and/or thereafter contacting the reacting mixture with concentrated nitric acid or nitrogen dioxide. In a special embodiment spent ion exchange resins are converted by this chemical digestion to noncombustible gases and a low volume noncombustible residue.

Abstract: After a first extraction of an aqueous nitric acid solution of spent nuclear fuel by a suitable organic solvent and a first washing of the resulting organic phase by an aqueous solution to remove tritiated water, the organic phase is submitted to a second washing with a volume of dilute aqueous solution of nitric acid free from tritium, said volume being substantially smaller than that of the organic phase, whereby the organic phase is rendered substantially tritium-free and contamination by tritium in the following stages of the process is avoided.