Abstract: Uranium is recovered from a carbonate leach solution containing a dissolved uranium salt and a monovalent ion. The pH of the leach solution is adjusted to about 5 to about 7.5, and preferably to about 6 to about 7. Phosphate ion is then added to typical in-situ leach solutions in an amount from about 10 to about 30 mole % in excess of the amount needed to stoichiometrically react with the uranium in said solution. This results in the precipitation of a compound made up of the monovalent ion, uranium, and the phosphate ion, which is insoluble in the solution. The precipitate is then separated from the solution preferably by means of a centrifuge or a vortex clarifier. It can then be dissolved in acid, and the uranium extracted into an organic solvent such as DEHPA-TOPA in kerosene.

Abstract: Uranium and other base metals are leached from their ores with aqueous solutions containing bicarbonate ions that have been generated or reconstituted by converting other non-bicarbonate anions into bicarbonate ions. The conversion is most conveniently effected by contacting solutions containing SO.sub.4.sup.= and Cl.sup.- ions with a basic anion exchange resin so that the SO.sub.4.sup.= and Cl.sup.- ions are converted into or exchanged for HCO.sub.3.sup.- ions. CO.sub.2 may be dissolved in the solution so it is present during the exchange. The resin is preferably in bicarbonate form prior to contact and CO.sub.2 partial pressure is adjusted so that the resin is not fouled by depositing metal precipitates. In-situ uranium mining is conducted by circulating such solutions through the ore deposit. Oxidizing agents are included in the injected lixiviant.

Abstract: Increasing the strip coefficient of a strip solution stream in a reductive stripping uranium recovery process involving wet process phosphoric acid feed is accomplished by adding effective amounts of H.sub.2 SO.sub.4 to the strip solution stream before it exits the reductive stripping means.

Abstract: In a uranium recovery process, barren, organic solvent, which contains entrained, reduced iron containing phosphoric acid, exiting the stripper is contacted with oxidized raffinate acid, before extraction, in an amount effective to remove about 70% to 98% of the entrained acid.

Abstract: Entrainment of contaminated water in the organic phase and poor phase disengagement is prevented in the second cycle scrubber, in a two cycle uranium recovery process, by washing the organic solvent stream containing entrained H.sub.3 PO.sub.4 from the second cycle extractor, with a dilute aqueous sulfuric or nitric acid solution in an acid scrubber, prior to passing the solvent stream into the second cycle stripper.

Abstract: An acidic solution containing metal values and also containing humic acid impurities, is purified by: (1) forming a gas in the acid solution, (2) adding high molecular weight anionic organic flocculant in a volume ratio of gas containing acid solution: flocculant of 1,000:0.002 to 0.10, to form a bottom, purified, aqueous acidic phase containing metal values and a top flocculated humic acid impurities froth phase and (3) removing the humic acid froth phase.

Abstract: A process to maintain a high iron +2/iron +3 concentration in the reductive stripping system for recovering uranium from wet-process phosphoric acid, comprises contacting partially oxidized strip acid, containing substantial amounts of iron +3, which had just previously stripped uranium values from an organic solvent, with elemental iron to reduce iron +3 to iron +2 and provide a Fe.sup.+2 /Fe.sup.o conversion ratio of from about 1.5 to 3.0.

Abstract: Disclosed is a method of removing uranium from a slurry containing a water-insoluble molybdate and a Group II element, usually calcium. The pH of the slurry is raised with an alkali metal hydroxide and maintained at a level sufficiently high to solubilize insoluble molybdenum compounds. This will generally require a pH of at least 10 which can be achieved by adding sodium hydroxide. Carbonate ion is then added to the slurry in an amount at least sufficient to precipitate the cation of the insoluble molybdenum compound. And finally, the solid portion of a slurry, which now contains calcium carbonate and sodium uranate, is separated from the liquid portion, which contains a soluble molybdenum compound.

Abstract: Precipitation of Fe.sub.3 HN.sub.4 H.sub.8 (PO.sub.4).sub.6 is prevented in the second cycle extractor, in a two cycle uranium recovery process, by washing ammonia laden organic solvent stream, from the second cycle stripper, with first cycle raffinate iron stream containing phosphoric acid, prior to passing the solvent stream into the second cycle extractor.

Abstract: Controlling the uranium extraction coefficient in uranium extraction processes involving wet process phosphoric acid feed is accomplished by monitoring the oxidation potential of the raffinate acid stream exiting the extractor, and maintaining the oxidation potential of the raffinate at a value above 350 mV.

Abstract: A process to restore solvent after the stripping step in recovering uranium from phosphoric acid, comprises passing stripped, water-immiscible organic extraction solvent, containing contaminants and having a lowered uranium extraction coefficient value, through a water scrub, to scrub contaminants from the organic solvent and raise its uranium extraction coefficient. The scrubbed solvent is then passed back to the extractor and the contaminated water is passed into at least one purification means and then back to the scrubber.

Abstract: A method of recovering liquid from an emulsion formed in a liquid-liquid contacting process, where the emulsion comprises at least one of the liquids entrapped by water-permeable membranes, comprises: admixing water and the emulsion in a volume ratio of water:emulsion of between 0.5 to 4.5:1 and raising the temperature of the admixture to between about 45.degree. C. and 99.degree. C., to cause water permeation of the membranes, causing them to disintegrate and release liquid, forming a three-phase composition consisting essentially of the two liquids and the solids.

Abstract: A neutron activation probe comprises a pulsed neutron source in series with a plurality of delayed neutron detectors for measuring radioactivity in a well borehole. The neutron source emits neutrons which produce fission in uranium in the ore body and the delayed neutron detectors measure the delayed neutrons produced from such fission while the probe is moved through the borehole. The signal from each neutron detector is delayed in time so that the activity measured by each neutron detector from a particular ore zone may be added thereby increasing the accuracy of the reading corresponding to that ore zone.

Abstract: A method is described for reducing the volume of radioactive waste produced during the solution mining of uranium and for recovering uranium from it. The recovery leach, which contains uranium in solution and is supersaturated with calcium carbonate, is treated with bicarbonate and made basic which precipitates calcium carbonate and some of the uranium. The precipitated calcium carbonate is dissolved with acid and the uranium in the solution is removed by extraction or precipitation. The remaining solution is contacted with sulfate ions and barium or strontium ions to precipitate BaSO.sub.4.RaSO.sub.4 or SrSO.sub.4.RaSO.sub.4, the principal radioactive constituent in the solid waste product.

Abstract: A method for optimal placement and orientation of a well field for solution mining comprises first determining the direction and magnitude of the major and minor axes of transmissivity of the ore body. Then, determining the location of the injection and recovery wells based on the magnitude and orientation of the major and minor axes of transmissivity.

Abstract: An emulsion, formed at the interface of an acidic solution and a solvent, is removed by passing at least one continuous belt through the emulsion by floating action, where portions of the emulsion attach to the surface of the belt material.

Abstract: A method for leaching mineral values from an underground ore body in situ comprises introducing an oxygen-bearing gas into a pressurized saturation-disengagement chamber located above ground of the ore body where the oxygen-bearing gas becomes dissolved in a leach solution. The leach solution with dissolved oxygen therein is then pumped into the well and through the ore body where the leach solution oxidizes the mineral values in the ore body permitting their solubilization and transport by the leach solution to a recovery well where the pregnant leach solution is pumped above ground. The pregnant leach solution is then treated to remove the mineral values therefrom. Insuring that the oxygen-bearing gas is completely dissolved in the leach solution above ground in the saturation-disengagement chamber prevents gas bubbles from being pumped into the ore formation, thereby preventing loss of injectivity.

Abstract: Uranium ore deposits which contain certain proportions of other metals and elemental components, such as are present in redox roll front ore deposits, are selectively leached in situ by passing therethrough relatively dilute aqueous leach solutions comprising essentially from about 0.5 to 5 grams per liter of ammonium bicarbonate and from about 0.1 to 3 grams per liter of peroxide, preferably introduced as aqueous H.sub.2 O.sub.2, and sufficient ammonia to bring the solution to a pH of from about 7.4 to 9, and preferably from 7.5 to 8.5, thereafter withdrawing from the ore deposit the aqueous leach solution enriched in uranium which it preferentially extracts along with a generally lower proportion of other metals and elements as compared to their respective ratios in the ore deposit, and contacting the enriched leach solution with a strong base anion exchange material to strip the uranium from the leach solution.