Abstract: A method and apparatus are provided for producing uranyl peroxide (UO.sub.4.2H.sub.2 O) from uranium-bearing metal pieces. The uranium-bearing metal pieces are dissolved in a first aqueous solution containing nitric and fluoboric acids to provide a second aqueous solution which includes uranyl ions (UO.sub.2.sup.+2) and nitric and fluoboric acids. Hydrogen peroxide is added to the second aqueous solution to precipitate uranyl peroxide out of that solution and provide a third aqueous solution which contains nitric and fluoboric acids. The uranyl peroxide is then separated from the third aqueous solution.

Abstract: A method and apparatus are provided for producing green salt (UF.sub.4) from uranium-bearing metal pieces. The uranium-bearing metal pieces are dissolved in a first aqueous solution containing hydrochloric and fluoboric acids to provide a second aqueous solution which includes uranium (U.sup.+4), chlorine ions (Cl.sup.-) and hydrochloric and fluoboric acids. Hydrofluoric acid is added to the second aqueous solution to precipitate green salt out of that solution and provide a third aqueous solution which contains hydrochloric acid. The green salt is then separated from the third aqueous solution.

Abstract: A treating process for the separation of coated nuclear fuel particles from a graphitic matrix in which the nuclear fuel particles are embedded, which is employed in an installation for the conditioning of graphitic fuel elements of high-temperature nuclear reactors. The graphite which encompasses the nuclear fuel particles is conducted away through the action of a brush which isolates the nuclear fuel particles together with their coatings or the nuclear particles themselves, and wherein the nuclear fuel particles which are contained in the brushed product are then separated from the comminuted graphite. The nuclear fuel particles are worked out of the graphitic matrix through the action of the bristles of the brush which conduct the graphite away, but in which the coated nuclear fuel particles when they are contained within hard coatings or the nuclear fuel particles themselves, remain preserved.

Abstract: A process for dissolving PuO.sub.2, NpO.sub.2, or fuel containing PuO.sub.2 and/or NpO.sub.2 in acid, particularly nitric acid by electrolytically continuously maintaining an oxidizing regenerable reagent. The regenerable reagent may be added to the mixture of oxide(s) and nitric acid. Alternately, it may already be present as a fission product or other nuclear reaction product in irradiated nuclear reactor fuel, in which case it need not be added to the nitric acid.

Abstract: Ferric nitrate leaching at 75.degree. C. is found to remove up to 97% of the uranium and 93% of the radium from ores occurring in the Elliot Lake area of Canada, after an initial flotation-separation of the sulfide minerals from the ore. In processes of the invention an aqueous acidic ferric nitrate solution of relatively low concentration, e.g. 0.01M to 0.1M is used, giving tailings which are effectively sulfide-free (less than 0.45 wt %) and with radium levels approaching a desired maximum value of 24 pCi/g. Radium may be removed from the leachate by adsorption and uranium by solvent extraction. 80-87% of the ferric nitrate may be recirculated for further leaching. Because of the low reagent concentrations and the recycle of the ferric nitrate, it is possible to keep nitrate ion levels in the effluent below the prescribed level of 10 ppm.

Abstract: The invention relates to a method for recovering rare metals from the combustion residues of various coals, in particular brown coals. The recovery is performed by digestion, more particularly by aqueous and/or dilute alkaline and/or dilute acidic digestion, where two or three of these steps can be combined in any desired order or they may be carried out separately. During digestion the concentration of the solution is monitored and the subsequent digestion steps are terminated at a desired concentration. The solid and liquid phase are then separated and the rare metals are isolated from the liquid phase while the solid phase, optionally after neutralization and/or washing can be utilized for example as a source of energy.

Abstract: The uranium, yttrium, thorium and rare earth values contained in a phosphate rock are essentially totally recovered therefrom by, during the wet-process preparation of phosphoric acid from said phosphate rock, acidulating same in the presence of aluminum and/or iron, and optionally silica, and thence recovering said values from the phosphoric acid thus produced.

Abstract: The present invention relates to insoluble compositions, which are capable of removing metal (e.g. selectively) from solution (e.g. Fe.sup.3+ from a liquid nutrient medium so as to lower the Fe.sup.3+ content to less than 0.1 .mu.M); the insoluble compositions comprise: a suitable insoluble carrier and organic co-ordinating sites covalently fixed to the surface of said carrier, said co-ordinating sites being capable of chelating Fe.sup.3+, Th.sup.4+ and/or UO.sub.2.sup.2+.

Abstract: A sealing composition suitable for sealing alumina packages for integrated circuits at a temperature below about 450.degree. C. for a short time of about 10 minutes. The composition is a mixture of 50-80 wt % vitreous solder glass powder of PbO-B.sub.2 O.sub.3 system having a deformation point of 350.degree. C. or less, 1-35 wt % zinc material powder and 1-35 wt % zircon powder. The zircon powder is of a synthetic zircon artificially prepared to eliminate radioactive impurities such as uranium and/or thorium.

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: The process for recovering metal values from mineral ores containing small amounts of metal, which comprises specific curing and conditioning steps followed by formation of a slurry of critical portions, which is processed by a countercurrent filter means to efficiently remove dissolved metal salts. The process of the invention permits rapid and efficient recovery of metal values from ores, such as uranium ores, and permits the economical use of very low grade ores.

Abstract: An otherwise unusable western sandstone uranium ore is blended with uranium-containing phosphate rock. The mixture is then digested with sulfuric acid to produce a wet process phosphoric acid solution containing higher levels of uranium. Recovery of the uranium values from this solution can be effected by various known processes such as solvent extraction.

Abstract: Method for dissolving hard-to-dissolve thorium and/or plutonium oxides, especially dioxides such as ThO.sub.2, PuO.sub.2 or (U/Pu)O.sub.2 mixed oxides by heating the oxides in a hermetically sealed vessel in fluoride-free nitric acid. The use of a gas atmosphere containing oxygen in the sealed vessel is advantageous.

Abstract: Reprocessing a nuclear reactor fuel rod which contains nuclear fuel in a cladding tube, by removing the cladding tube from the nuclear fuel. The cladding tube is uniformly heated in hermetically sealed condition together with the nuclear fuel contained thereon to permanently expand the diameter of the cladding tube without the formation of cracks in the cladding tube, increasing the distance between the nuclear fuel and the cladding tube. Subsequently the expanded cladding tube is opened at one end. The nuclear fuel is removed from the opened cladding tube and is processed further, separate from the cladding tube.

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 method of reversibly immobilizing sulfate ash at least about 20% of which is sulfates of transuranic elements. The ash is mixed with a metal which can be aluminum, cerium, samarium, europium, or a mixture thereof, in amounts sufficient to form an alloy with the transuranic elements, plus an additional amount to reduce the transuranic element sulfates to elemental form. Also added to the ash is a fluxing agent in an amount sufficient to lower the percentage of the transuranic element sulfates to about 1% to about 10%. The mixture of the ash, metal, and fluxing agent is heated to a temperature sufficient to melt the fluxing agent and the metal. The mixture is then cooled and the alloy is separated from the remainder of the mixture.

Abstract: A treating process is provided for the separation of coated nuclear fuel particles from a graphitic matrix in which the nuclear fuel particles are embedded, and an installation provides for the conditioning of graphitic fuel elements of high-temperature nuclear reactors. The graphite which encompasses the nuclear fuel particles is conducted away through the action of a brush which isolates the nuclear fuel particles together with their coatings or the nuclear particles themselves. The nuclear fuel particles which are contained in the brushed product are then separated from the comminuted graphite.

Abstract: The present invention relates to the in situ recovery of mineral values, particularly uranium, from subterranean formations that contain sulfur in organic or inorganic forms where the sulfur causes premature deletion of an oxidant and reduction in permeability of the formation. The adverse effects due to the presence of the sulfur in the formation are substantially reduced in accordance with the present invention by treating the formations with an aqueous solution of iron-complexing agent and then oxident to preferentially oxidize and solubilize at least a portion of the sulfur in the formation. The present process may be applied either as a pre-treatment process to be followed by a leaching process or simultaneously with the leaching process.

Abstract: A process is described for the recovery of uranium from ore wherein the ore is leached, the resulting leachate is passed over an anion exchange resin, and the resin is then eluted with an elution solution containing an anion capable of replacing the uranyl anion captured on the resin, followed by treatment with a sulfite-containing solution so that polythionates captured on the ion exchange resin are reacted to form essentially thiosulfate and trithionate. Preferably, the sulfite solution also contains other anions.

Abstract: A process of leaching ores, ore concentrates, intermediate metallurgical products and the like, which contain acid-soluble constituents, by a treatment in an acid aqueous solution under pressure in a tubular reactor. The slurry contains suitably 300 to 1100 g starting material per liter of water and a velocity of flow of 0.5 to 4.0 m/sec. is suitably maintained throughout the flow path in the tubular reactor of the slurry and of the leached suspension.

Abstract: Rare earth, uranium and thorium values are separated and recovered from aqueous solutions of the chlorides thereof, said aqueous solutions having an acidity of less than 2 N and a concentration in such values, expressed as the oxides thereof, of at least 50 g/l, by liquid-liquid extracting said aqueous solutions with an organic extractant comprising at least one neutral organophosphorus compound.

Abstract: Ferric chloride leaching at temperatures in the range 47.degree.-74.degree. C. is found to remove up to 97% of the uranium from ores occurring in the Elliot Lake area of Canada, but radium removal was found to be poor due to the formation of sulphates from the sulphides present in the ore. In processes of the invention the sulphides are initially removed by flotation, when aqueous acidic ferric chloride of relatively low concentration, e.g. 0.1 M can extract as much as 92% of the radium, giving tailings which are effectively sulphide-free and with radium levels approaching a desired maximum of 24 pCi/g. Radium may be removed by adsorption on manganese dioxide and uranium may be removed by liquid extraction with D2EHPA (DAPEX process). The ferric chloride may be recirculated for further leaching, with reduction before the uranium extraction and reoxidation afterwards. Because of the recycle, it is possible to keep chloride ion levels in the effluent below the prescribed level in Ontario, Canada of 750 ppm.

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 metallic values by putting the values into solution and separating undesired mineral matter from the solution using countercurrent flotation is provided. The process involves leaching metallic values from host rock, conditioning the resultant ore pulp with the required reagents to achieve selective flotation of mineral matter in the metallic values solution, introducing the conditioned ore pulp into flotation cells, along with counterflow of solution from an immediately subsequent flotation step, wherein simultaneous washing and flotation is achieved, and the mineral matter is removed leaving a solution of the metallic values. The resultant mineral matter froth product is subjected to subsequent stages of flotation and simultaneous washing with counterflow of solution removed from each subsequent stage of flotation, water and/or barren solution being used for washing in the final flotation stage.

Abstract: The present invention relates to a process for the recovery of uranium and other minerals from uranium ore where at least part of the uranium is present as refractory uranium-mineral complexes, comprising subjecting the uranium ore to mild oxidative carbonate leach fluid to dissolve and remove readily soluble uranium minerals, subsequently subjecting the uranium ore to an oxidative chemically severe leaching system to dissolve and remove the refractory uranium-mineral complexes, and separating and recovering the uranium and other mineral species in the leachate fluids. The process may be applied to in-situ uranium leaching operations or to surface leaching operations.

Abstract: Dissolving hard-to-dissolve nuclear fuels such as ThO.sub.2 and PuO.sub.2 in a nitric acid fission material solution which contains nitric acid and a Pu-fluoride complex to provide free fluorine ions in a small catalytically active amount due to small dissociation of fluoride ions from the Pu-fluoride complex. This permits dissolving hard-to-dissolve nuclear fuels in normal metallic dissolving vessels and also eliminates need for using polytetrafluoroethylene.

Abstract: The invention is a process for the removal of radium from acidic aqueous solutions. In one aspect, the invention is a process for removing radium from an inorganic-acid solution. The process comprises contacting the solution with coal fly ash to effect adsorption of the radium on the ash. The radium-containing ash then is separated from the solution. The process is simple, comparatively inexpensive, and efficient. High radium-distribution coefficients are obtained even at room temperature. Coal fly ash is an inexpensive, acid-resistant, high-surface-area material which is available in large quantities throughout the United States. The invention is applicable, for example, to the recovery of .sup.226 Ra from nitric acid solutions which have been used to leach radium from uranium-mill tailings.

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: Uranium can be extracted from its ores at a pH of 2.5 to 5.5 using sulfuric acid, hydrogen peroxide, trace of iron and a sulfate. The extraction process is applicable to both tank leaching of conventionally mined ores and in situ leaching.

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: The phenolic chelate resin has a chelate-forming group wherein part or all of the hydrogen atoms in a primary and/or secondary alkylamino group introduced in a phenol nucleus are replaced by a methylenephosphonate group. Also disclosed is a process for producing such a chelate resin, and a method for recovering heavy metal ions with such a resin. The resin has particularly high selectivity for adsorbing uranium ions, as well as high heat, acid and alkali resistance as well as dimensional stability. The resin is very effective for recovering uranium from various uranium-containing solutions such as sea water, crude phosphoric acid fertilizer solutions, low-grade uranium ore, waste water from uranium refining, and uranium mine water.

Abstract: Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

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 process is described for minimizing accumulation of undesirable polythionates on an ion exchange resin used to recover uranium values from a leachate from an in situ mining operation by adding sulfite, either as a sulfite salt or as SO.sub.2, directly to the leachate or to the eluant.

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 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: A two-stage in situ uranium leaching process is described wherein H.sub.2 SO.sub.4 /oxidant is used as the principal leaching solution. Prior to leaching, the formation is treated with a dilute solution of hydrochloric acid, carbonic acid, or other acid (without oxidant) capable of removing the most accessible portions of CaCO.sub.3. The pretreatment improves the permeability of the formation and precludes plugging with CaSO.sub.4. The invention is particularly advantageous when applied to ores which cannot readily be leached with alkaline solutions.

Abstract: Metals such as iron, uranium, vanadium, molybdenum and rare earths are reduced to lower oxidation states in various acid media using silicon metal or an iron-silicon alloy. In particular, ferric iron and hexavalent uranium in wet-process phosphoric acid are reduced to the ferrous and tetravalent states, respectively, using silicon metal or an iron-silicon alloy to provide a feed acid which is suitable for extraction with an extractant which is selective for tetravalent uranium such as a mixture of mono- and di-(alkylphenyl) esters of orthophosphoric acid.

Abstract: A method for continuously extracting uranium from ores comprises the steps of forming a slurry of ore in a leaching solution; heating the slurry while pumping it through a tube reactor at high turbulences characterized by Reynolds numbers in excess of 50,000; supplying gaseous oxygen at high pressures of at least 30 bar into the tube reactor such that the uranium is substantially completely oxidized in a soluble form but impurities in the slurry are substantially kept from becoming soluble; recovering the uranium oxide solute which is substantially free of impurities.

Abstract: Uranium ore is crushed and formed into a slurry. The slurry is pumped into a leaching tank and classified by a sparge tube projecting water at teeter velocity at the bottom of the ore bed. Thereafter, a second ore bed is pumped by the slurry into the tank and the second bed classified. This is repeated until there are four or five ore beds in the tank.Thereafter, the ore is leached by passing liquid up from the bottom of the tank at velocities less than teeter velocities so that the bed remains classified.After leaching, the ore is formed into a slurry and the slurry pumped from the tanks to disposal.

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: 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.