Abstract: A method of dissolving impure uranium tetrafluoride in a hot state in a nitric acid solution in the presence of an aluminum compound. For the purpose of obtaining a uranyl nitrate solution which can easily be separated from the solid phase formed during treatment, the dissolving is carried out in two stages at an appropriate temperature. The first stage comprises introducing quantities of nitric acid and of the aluminum compound which are insufficient to dissolve the impure uranium tetrafluoride completely, and keeping the resultant suspension agitated for a period of at least 0.5 hour. The second stage comprises introducing quantities of nitric acid and of the aluminum compound which are at least sufficient to dissolve the uranium not dissolved in the first stage, while keeping the suspension agitated.

Abstract: Mineral values, particularly uranium, are recovered from subsurface earth formations containing the same, as well as cations which form precipitates with sulfuric acid leach solutions, by injecting into at least one injection well a preliminary acidic solution capable of forming soluble materials with at least one of the precipitate-forming cations, such as calcium and iron cations, passing the preliminary acidic solution through the subsurface formation for a time sufficient to form such soluble materials, withdrawing the preliminary acidic solution containing the solubilized materials from at least one producing well, thereafter, injecting a sulfuric acid leach solution into said injection well, contacting the subsurface formation with the leach solution for a time sufficient to extract significant amounts of mineral values from the formation and produce a pregnant leach solution containing the thus solubilized mineral values and withdrawing the pregnant leach solution from the production well.

Abstract: A process for the recovery of uranium from spent shale by treating the spent shale with a compound selected from the group consisting of methanol, a mixture of methanol and water, and a mixture of methanol and sodium methoxide at between about 240.degree. and 450.degree. C. and atmospheric pressure or higher. The treated spent shale is then leached with an aqueous acid solution such as hydrochloric acid, sulfuric acid, or nitric acid to remove 90% of the uranium from the spent shale and the leached uranium is recovered from the acid solution in a manner known per se.

Abstract: Disclosed is a process for the selective separation of uranium from metals accompanying it in a uranium-containing ore, comprising the steps of preparing a uranium-containing solution; adding to the solution (i) hydrochloric acid in an amount sufficient to form complex anions of the type(UO.sub.2 Cl.sub.n).sup.n-2wherein n is 3 or 4, or (ii) sulfuric acid in an amount sufficient to form complex anions of the typeUO.sub.2 (SO.sub.4).sub.m.sup.2-2mwherein m is 2 or 3; adding to the solution a cationic surface active agent which forms a relating insoluble precipitate with the complex anion; subjecting the solution containing said precipitate to a gas flotation step; separating the foam fraction from the liquid fraction; and recovering uranium from the foam fraction. Also disclosed is an apparatus for carrying out the process.

Abstract: The present invention relates to a process for recovery of uranium from a carbonate lixiviant additionally containing other contaminants such as molybdenum and silica, by adjusting the pH of the lixiviant to a value of at least 4 and treating the lixiviant to obtain a carbonate concentration of at least 50 ppm by weight. Subsequently the lixiviant is treated with an aqueous solution containing a sufficient amount of ferric iron to precipitate the contaminants without substantial precipitation with the uranium values. The precipitate is separated from the lixiviant. The treated lixiviant is passed through an ion exchange to retain uranium values.

Abstract: A process is described for improving yields and leaching rates of mineral values in highly reducing uranium-bearing formations, while minimizing deleterious environmental impact, by injecting an oxidant such as gaseous air or O.sub.2 into the formation prior to leaching. The preoxidation may be enhanced by the presence of CO.sub.2 gas in the pre-leaching oxidant. The process is particularly suitable for systems employing a CO.sub.2 /O.sub.2 lixiviant. The presence of sulfate ion further improves the leaching rate of such a system.

Abstract: A process for recovering one or more non-radioactive transition metal compounds from an ore containing one or more compounds of said transition metal or metals and further containing at least one complex of a member selected from the group consisting of uranium, thorium, radium, titanium, and rare earth metals, which comprises decomposing said ore in crushed condition by means of an acid so that a portion of the ore is brought into solution in a liquid phase and another portion of the ore remains in a solid phase, said compound or compounds of the transition metal or metals to be recovered passing into only the liquid or into only the solid phase, the uranium in the crushed ore being treated so as to cause substantially all of said uranium to be present in an oxidation state in which it cannot, during the decomposition step, pass into the phase containing the transition metal compound or compounds.

Abstract: An improved method and apparatus for effecting a substantial reduction in retention time of oxidation processes by the autoclave oxidation of ion species dissolved in aqueous solutions. In one embodiment, the invention provides for oxidation of dissolved reduced ion species including ferrous iron (Fe.sup.+2), uranium (U.sup.+4) and vanadium (V.sup.+3) from wet process phosphoric acids, or for conditioning of aqueous solutions for other uses, such as corrosion control, by raising the valence state of dissolved ion species. In accordance with the present invention, pure oxygen, present in quantity of at least 94% by volume, is utilized as an oxidizing agent with specified process parameters: temperature, pressure and mixing conditions in an autoclave reaction vessel.

Abstract: A process for recovering one or more nonradioactive transition metal compounds from an ore containing one or more compounds of said transition metal or metals and further containing at least one complex of a member selected from the group consisting of uranium, thorium, radium, titanium, and rare earth metals, which comprises decomposing said ore in crushed condition by means of an acid so that a portion of the ore is brought into solution in a liquid phase and another portion of the ore remains in a solid phase, said compound or compounds of the transition metal or metals to be recovered passing into only the liquid or into only the solid phase, the uranium in the crushed ore being treated so as to cause substantially all of said uranium to be present in an oxidation state in which it cannot, during the decomposition step, pass into the phase containing the transition metal compound or compounds.

Abstract: An economically advantageous method of recovering uranium from a wet process phosphoric acid solution through the steps of making hemihydrate gypsum contact with the acid solution thereby transferring uranium from the acid solution into the gypsum, dispersing the U-containing gypsum separated from the acid solution in water to convert the gypsum to dihydrate accompanied by the transfer of uranium into water, separating the obtained U-containing aqueous solution from the dihydrate gypsum, and adding precipitant such as an inorganic base to the aqueous solution to form a precipitate comprising an insoluble uranium compound. The contact of hemihydrate gypsum with the phosphoric acid solution is preferably preceded by reduction of hexavalent uranium in the acid solution to tetravalent uranium, and can be achieved either by adding hemihydrate gypsum to the acid solution or by converting dihydrate gypsum to hemihydrate within the acid solution preferably preceded by the addition of sulfuric acid.

Abstract: A method of improving the acid leaching of uranium ores which comprises adding to such ores while they are being leached between 0.01-2 lb. per ton of a composition comprising:a. aminotris methylene phosphonic acid, andb. a water-soluble copolymer of vinyl sulphonic acid and acrylic acid combined in a weight ratio of from 2:1 to 1:2 and having a molecular weight within the range of from 500 to 100,000,with the weight ratio of a:b being within the range of 1:2 to 2:1.

Abstract: In preparing wet process phosphoric acid by decomposing a phosphate rock containing uranium with sulfuric acid and phosphoric acid on condition that hemihydrate gypsum is formed in an acid solution either at the stage of decomposing the phosphate rock or subsequently, uranium contained in the phosphate rock can almost entirely be retained in the obtained phosphoric acid solution by forming the hemihydrate gypsum in the presence of an oxidizing agent, such as a soluble chlorate, hydrogen peroxide or oxygen gas, in the acid solution in a quantity sufficient to render the entire uranium dissolved in the acid solution hexavalent because hemihydrate gypsum adsorbs almost exclusively tetravalent ions of uranium. The uranium is then recovered.

Abstract: A wet process for producing phosphoric acid and phosphohemihydrate by acidulation of phosphate rock. A strong phosphoric acid is obtained with concomitant production of an improved calcium sulfate that is extremely low in radioactivity by operation in Region II for shorter time so as to produce small sized hemihydrate particles, then hydration of a substantial proportion but less than all of the hemihydrate to coarse gypsum substantially free of radioactivity in a dilute acidic aqueous slurry at a temperature and P.sub.2 O.sub.5 concentration selected so as to be within Region I of FIG. 1 and having a substantial weight proportion of coarser gypsum particles that are substantially free of radioactivity.

Abstract: A process for reducing radioactive contamination of phosphogypsum. Phosphogypsum containing radioactive material is calcined to form hemihydrate crystals carrying the radioactive contaminants, and a portion of the crystals is converted to substantially radiation-free gypsum crystals which are readily separated from the hemihydrate crystal relics containing substantially all of the radioactive contamination.

Abstract: The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

Abstract: A direct precipitation method based on the insolubility of uranyl phosphate in carbonate containing solutions at pH 6-6.5 and its insolubility at pH 8 is described. The method eliminates the use of ion exchange columns for removing uranium values from uranium leachates and can be readily applied to an integrated field operation.

Abstract: A process for the recovery of uranium values from uranium-containing material which also contains iron, arsenic and siliceous matter, includes leaching the uranium-containing material in aqueous sulphuric acid solution under conditions to provide dissolved iron present in the resultant leach solution as predominantly ferrous iron rather than ferric iron and/or to provide a sulphuric acid concentration in the leach solution sufficiently high to substantially prevent the precipitation of arsenates. Uranium values are recovered from the leach solution by solvent extraction agent which has little affinity for arsenic.

Abstract: A process for recovering uranium values from a sulphate solution containing dissolved uranium and molybdenum and with a pH not exceeding about 5.5, includes reacting the solution with ammonia at a pH in the range of from about 8 to about 10, without the solution existing for any significant time at a pH of around 7, with resultant precipitation of uranium values relatively uncontaminated by molybdenum. The uranium containing precipitate is separated from the remaining solution while maintaining the pH of the remaining solution within the same range.

Abstract: A process for solubilizing uranium and other values in an ore containing the same together with aluminum and phosphorus and effecting regeneration of a mineral acid employed to solubilize the ore. The regeneration is effected by heating the spent acid solution resulting from solubilization of the ore in a reaction zone to a temperature above 100.degree. C. while maintaining at least the autogenic pressure of the heated solution within the reaction zone. The treatment causes aluminum phosphate to precipitate from the solution while simultaneously causing regeneration of at least a portion of the mineral acid in the solution that was consumed to originally solubilize the ore. The dissolved uranium and other elements of value can be recovered from the solution, either before or after acid regeneration, by any known technique. The regenerated mineral acid then can be used to solubilize fresh ore.

Abstract: A process for treating water containing chemicals comprising ammonium nitrate, fluorides, hydrogen peroxide and uranium compounds, wherein valuable and/or hazardous components of the chemicals-containing water are separated therefrom and concentrated for recovery. The process includes an ion exchange operation in addition to manipulations of chemical components and pH conditions of the water by means of the application of reagents and adsorbents.

Abstract: The level of radioactivity in gypsum produced from phosphate rock can be reduced by adding to the slurry of acid and phosphate rock in the conventional process for producing phosphoric acid a combination of concentrated nitric acid or hydrochloric acid and an ammonium salt, especially ammonium nitrate or ammonium phosphate, before the precipitation of the calcium sulfate.

Abstract: A process is described for restoring and maintaining the total ion exchange capacity of the resin used for uranium recovery in in situ uranium leaching. The used resin of lowered exchange capacity is treated with a solution containing Na.sub.2 CO.sub.3 or NaHCO.sub.3, or admixtures thereof. The process preferably is used in conjunction with acid elution of the uranium from the resin.

Abstract: A process is described for improving yield and leaching rates of mineral values in uranium-bearing formations associated with high brine aquifers by using high pressure CO.sub.2 /oxidant in the leaching solution. The high pressure CO.sub.2 overcomes the inhibiting effect of NaCl on the rate of leaching. Uranium is recovered at the well head by ion exchange at a pH of about 4.5 to about 5.0 either under pressure to keep the CO.sub.2 in solution or with provision for CO.sub.2 recovery and recycle.

Abstract: Uranium ores, concentrates, calcines or tailings are processed to remove radium and thorium as well as uranium. Selected ores, concentrates or tailings, or if more appropriate, chlorination calcines thereof, are leached by selected aqueous chlorine-containing media (preferably in two stages) until uranium, radium and thorium are substantially all dissolved, with the insoluble residual solids being suitable for disposal. The leach solution is treated to recover sequentially uranium, usually thorium, and radium by selected techniques. The radium recovered can be disposed of in any environmentally-acceptable manner. The amount of iron in the residual leach liquor should be controlled to avoid iron build-up, with the barren leach solution being suitable for recycle.

Abstract: Uranium in wet-process phosphoric acid in the tetravalent state is extracted with a mixture of mono- and di-(alkylphenyl) esters of orthophosphoric acid containing a phenol modifier such as nonylphenol or octylphenol.

Abstract: Methods for control of excessive corrosion in wet process phosphoric circuits are effected by oxidation of reduced ion species in the acid with maintenance of EMF value of the acid above about 190 millivolts (S.C.E. reference) through the digestion circuit. Supplementary monitoring is by a ceric red-ox titration procedure. In one embodied form, the present invention is utilized to control corrosion in a phosphoric acid plant digestion system (and subsequent processing equipment) operating by a dihydrate (gypsum) process even when such a process employs a calcined western U.S. phosphate ore feed from a relatively highly carbonaceous ore. The invention is also applicable for treating wet process phosphoric acid in phosphoric acid plants having a digester system operating by a hemihydrate process.

Abstract: A process for removing arsenic from an aqueous solution containing arsenic in the form of a soluble arsenate includes adjusting the pH of the solution if necessary to at least about 10, adding a soluble barium salt to precipitate arsenic as barium arsenate, and removing the precipitated barium arsenate from the solution.

Abstract: A process is described for improving yields and leaching rates of mineral values in highly reducing uranium-bearing formations, while minimizing deleterious environmental impact, by injecting an oxidant such as gaseous air or O.sub.2 into the formation prior to leaching. The preoxidation may be enhanced by the presence of CO.sub.2 gas in the pre-leaching oxidant. The process is particularly suitable for systems employing a CO.sub.2 /O.sub.2 lixiviant. The presence of sulfate ion further improves the leaching rate of such a system.

Abstract: The present invention relates to in-situ leaching of uranium, particularly employing an acidic leach liquor containing an oxidant, and especially in respect of ores containing significant amounts of transition metals that act as catalysts for peroxidant decomposition. When hydrogen peroxide is used as oxidant under such conditions it decomposes leading to the formation of gas bubbles and exacerbation of ore-blinding, and a reduction in the efficiency of extraction of uranium.The present invention employs peroxymonosulphuric acid as oxidant and thereby ameliorates the problems aforesaid. Preferably, additionally, sulphuric acid is present in the leach liquor and in many preferred embodiments the peroxymonosulphuric acid concentration is from 0.001 to 0.03 moles/liter and the sulphuric acid from 0.025 to 0.075 moles/liter.

Abstract: A method of accelerating and increasing the solubility in nitric acid of plutonium dioxide and uranium/plutonium oxide fuels and insoluble residues from such nitric acid dissolution, includes the steps of arranging metal loadings to be at least 200 g (Pu) or (U+Pu) per liter of dissolving solution so that the plutonium in solution acts as an autocatalyst.

Abstract: A method for controlling calcium, e.g. calcite, build-up in the leach solution of a uranium and/or related values recovery operation wherein the leach solution is flowed through a value bearing ore to dissolve the desired values. A soluble fluoride, e.g. sodium fluoride, is added to the leach solution after it has passed through the ore to thereby precipitate calcium fluoride from the leach solution and lower the calcium content of the leach solution. The soluble fluoride may be added to the leach solution before the leach solution passes through the process equipment which is used to remove the values from the leach solution or the soluble fluoride may be added after the leach solution passes through the process equipment. If added before, it is preferable to also add a carbonate/bicarbonate solution along with the soluble fluoride to prevent coprecipitation of uranyl/desired value fluoride or to redissolve coprecipitated fluoride back into the leach solution.

Abstract: Phosphoric acid and gypsum are produced from phosphate rock by digesting same with sulfuric acid to produce a high concentration of phosphoric acid and calcium sulfate hemihydrate. The solubilized uranium in the rock is rendered recoverable by reducing same from the +6 valence state to the +4 valence state during the digestion step followed by oxidizing back to the +6 valence state in a subsequent step.

Abstract: Refractory uranium is crushed and impregnated with a concentrated aqueous solution of sulphuric acid to provide a mixture which retains a solid phase consistency. The impregnated mixture is disintegrated in an aqueous solution to form a thick paste or pulp. The pulp is digested for a duration not exceeding 10 hours and the uranium recovered by solid/liquid separation. The sulphuric acid content of the concentrated solution is selected for the residual acidity at the end of pulp digestion to be of from 10 g/l to 40 g/l.

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: A process is described for the recovery of uranium values from phosphoric acid utilizing an alkyl pyrophosphoric acid (APPA) primary extractant. After extracting the uranium from the phosphoric acid, the APPA extractant is deactivated by heating and the uranium values stripped into a phosphoric acid strip solution containing ferric ion as a salting agent. The uranium values may then be re-extracted directly from this stripping solution without adjustment of its concentration into a dialkyl phosphoric acid trialkyl phosphine oxide synergistic extractant from which a relatively pure yellow cake is precipitated. A new procedure for preparing the requisite APPA primary extractants is also disclosed.

Abstract: High purity ammonium uranyl tricarbonate is produced from crude plant yellow cake by treatment of the yellow cake with a mineral acid, neutralization of the acidic solution with a conversion of the uranium therein to ammonium uranyl tricarbonate, precipitation of the ammonium uranyl tricarbonate with a non-solvent, followed by separation. Optionally, the ammonium uranyl tricarbonate can be calcined to produce high purity uranium oxides.

Abstract: A method and apparatus for processing spent nuclear reactor fuel wherein plutonium is continuously contaminated with radioactive fission products and diluted with uranium. Plutonium of sufficient purity to fabricate nuclear weapons cannot be produced by the process or in the disclosed reprocessing plant. Diversion of plutonium is prevented by radiation hazards and ease of detection.

Abstract: Methods for control of excessive corrosion in wet process phosphoric circuits are effected by oxidation of reduced ion species in the acid with maintenance of EMF value of the acid above about 190 millivolts (S.C.E. reference) through the digestion circuit. Supplementary monitoring is by a ceric red-ox titration procedure. In one embodied form, the present invention is utilized to control corrosion in a phosphoric acid plant digestion system (and subsequent processing equipment) operating by a dihydrate (gypsum) process even when such a process employs a calcined western U.S. phosphate ore feed from a relatively highly carbonaceous ore. The invention is also applicable for treating wet process phosphoric acid in phosphoric acid plants having a digester system operating by a hemihydrate process.

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: 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: 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: 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: 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 rendering actinide values recoverable from sodium carbonate scrub waste solutions containing these and other values along with organic compounds resulting from the radiolytic and hydrolytic degradation of neutral organophosphorous extractants such as tri-n butyl phosphate (TBP) and dihexyl-N,N-diethyl carbamylmethylene phosphonate (DHDECAMP) which have been used in the reprocessing of irradiated nuclear reactor fuels. The scrub waste solution is preferably made acidic with mineral acid, to form a feed solution which is then contacted with a water-immiscible, highly polar organic extractant which selectively extracts the degradation products from the feed solution. The feed solution can then be processed to recover the actinides for storage or recycled back into the high-level waste process stream. The extractant is recycled after stripping the degradation products with a neutral sodium carbonate solution.

Abstract: A method for concentrating and recovering uranium material from an aqueous solution, comprises passing a feed solution containing uranium through at least one reverse osmosis membrane system to concentrate the uranium, and then flushing the concentrated uranium solution with water in a reverse osmosis membrane system to further concentrate the uranium.

Abstract: Molybdenum is separated from molybdenum-containing activated charcoal or char also containing small amounts of uranium obtained as a by-product in uranium leaching processes by stripping with an alkaline solution to provide a molybdenum-containing solution containing substantially less than 500 ppm U.sub.3 O.sub.8.

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: 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: 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.