Abstract: A nonaqueous adhesive skin patch at least comprises a backing film and an adhesive layer placed thereabove, the adhesive layer containing a local anesthetic, wherein when the total mass of an adhesive is set to 100% by mass, the adhesive layer includes the following in the ratio of: 20 to 40% by mass of an elastomer; 20 to 35% by mass of a softener; 20 to 40% by mass of a resin; 3.0 to 7.0% by mass of the local anesthetic; and 0.3 to 4.0% by mass of organic acid, and wherein the adhesive layer has a mass per unit area of 100 to 200 g/m2.

Abstract: An object of the present invention is to provide a polyhedral polysiloxane composition that has high heat resistance and high light resistance, is excellent in gas-barrier properties and thermal shock resistance, and exhibits good handleability when used to encapsulate an optical semiconductor device. The polyhedral polysiloxane composition of the present invention is characterized by including a modified polyhedral polysiloxane which is obtained by hydrosilylation of an alkenyl group-containing polyhedral polysiloxane compound (a) and a hydrosilyl group-containing compound (b) and has a structure derived from an organic silicon compound (a?) having one alkenyl group per molecule.

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: Method for coprecipitation (or simultaneous precipitation) of at least one actinide in oxidation state (IV) with at least one actinide in oxidation state (III), wherein: a solution i.e. mixture of actinide(s) in oxidation state (IV) and actinide(s) in oxidation state (III) is prepared by adding to it a singly charged cation whose presence makes it possible to stabilize the aforementioned oxidation states in the mixture, or a singly charged cation which does not act to stabilize the aforementioned oxidation states in the mixture; a solution containing oxalate ions is mixed with the said mixture of actinides in order to carry out coprecipitation, i.e. simultaneous precipitation, of the said actinides in oxidation states (IV) and (III) and a fraction of the singly charged cation. According to another embodiment, a solution i.e.

Abstract: The invention relates to a method for separating uranium(VI) from one or more actinides selected from actinides(IV) and actinides(VI) other than uranium(VI), characterized in that it comprises the following steps: a) bringing an organic phase, which is immiscible with water and contains the said uranium and the said actinide or actinides, in contact with an aqueous acidic solution containing at least one lacunary heteropolyanion and, if the said actinide or at least one of the said actinides is an actinide(VI), a reducing agent capable of selectively reducing this actinide(VI); and b) separating the said organic phase from the said aqueous solution. Applications: reprocessing irradiated nuclear fuels, processing rare-earth, thorium and/or uranium ores.

Abstract: The present invention generally relates to the preparation of mixed actinide oxides, such as mixed oxides of uranium and plutonium (U, Pu) O2, by simultaneously coprecipitation and then calcinations.

Abstract: Disclosed herein are siloxane-based resins prepared by hydrolyzing and polycondensing cyclic and/or cage-shape siloxane compounds, optionally with at least one silane compound, in an organic solvent in the presence of a catalyst and water. Also, disclosed herein are methods for forming insulating film between interconnect layers in semiconductor devices by using the siloxane-based resins thus prepared as low dielectric insulating materials.

Abstract: A process for removing dissolved uranium from water is provided. The process basically comprises (a) mixing phosphoric acid or particulate bone ash with the water, (b) mixing calcium hydroxide with the mixture produced in step (a) to thereby form calcium hydroxy phosphate or calcium hydroxy apatite which reacts with and complexes at least a portion of the uranium in the water to form a precipitate thereof, and (c) separating the precipitate from the water.

Abstract: Process for the production of nuclear fuel pellets based on mixed uranium and plutonium oxide having a specific plutonium content from a charge of UO.sub.2 and PuO.sub.2 powders by lubrification, pelletizing and sintering, in which a solid, sulphur, organic additive of the zwitterion type is incorporated into the mixture during the co-milling stage for the powders.

Abstract: A method of reducing the concentration of neptunium and plutonium from alkaline radwastes containing plutonium and neptunium values along with other transuranic values produced during the course of plutonium production. The OH.sup.- concentration of the alkaline radwaste is adjusted to between about 0.1M and about 4M. [UO.sub.2 (O.sub.2).sub.3 ].sup.4- ion is added to the radwastes in the presence of catalytic amounts of Cu.sup.+2, Co.sup.+2 or Fe.sup.+2 with heating to a temperature in excess of about 60.degree. C. or 85.degree. C., depending on the catalyst, to coprecipitate plutonium and neptunium from the radwaste. Thereafter, the coprecipitate is separated from the alkaline radwaste.

Abstract: There is disclosed a waste treatment process for alkaline waste liquid in which substances to be removed exist in ionic and colloidal states. In the process, powdered tannin is added to the waste liquid to produce solid substance of the tannin and to capture the substances to be removed on the solid substance. Then, the waste liquid is subjected to filtration to thereby separate the solid substance including the substances to be removed.

Abstract: Sodium silicate (waterglass) is added to a waste process stream containing metal to form a waterglass sludge containing the metal contaminants. The waterglass sludge matrixed with the metal is removed from the stream. Thereafter, a caustic is added to the separated waterglass sludge which dissolves the sodium silicate, leaving a metal solid in suspension. The sodium silicate solution is filtered off and recycled for use in the waterglass precipitation process. The metal solid remains and concentrated acid is added to form a metallic acid solution which is substantially free of silicates. This solution may then be treated by solvent extraction or other means to recover the metal. The process provides substantial metal recovery from the process waste stream and eliminates the need for burial of the waste, thereby eliminating burial costs.

Abstract: Stable alkaline aqueous solutions of sodium tetraphenylborate.

Abstract: This invention can treat a uranium-containing solution of high or low concentration and/or the waste generated from uranium conversion processes, etc. It is characterized by the use of acorns, nuts of oak tree, which can be easily obtained in our botanical system. By coprecipitating the uranium or other heavy metal elements with the extract of acorns, this invention can not only recover the uranium of other heavy metal elements but also can reduce them before being discharged to the environmental radioactivity level.

Abstract: Improvement to the separation and recovery of solids from liqiuds containing them, and using a ionic flotation process, implementing an installation characterized in that it comprises at least a vat (1) for pretreating the solution to be treated for the insolubilization of the solid to be recovered and the formation of a suspension maintained under stirring conditions (stirrer 2), at least a flotation cell (5) supplied with said suspension, at least a device (4) provided between said vat (1) and said flotation cell (5) adapted to generate within the cell a rising non-turbulent stream intended to convey the desired solid towards the surface of said suspension while maintaining at said surface a calm area for the accumulation of said solid and at least one means for the recovery of said solid (6-7).

Abstract: A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper (II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the orginal organic compounds, is subsequently blended with high level radioactive sludge and transferred to a virtrification facility for processing into borosilicate glass for long-term storage.

Abstract: A method of treating a heavy metal and/or a radioactive metal-containing natural water or liquid such as a radioactive metal-containing wastewater stream, a potable water supply containing naturally-occuring radioactive elements, an oil containing one or more radioactive metals, or other nuclear metal-bearing liquid by contacting the radioactive heavy metal-containing liquid with a water-insoluble carboxylated cellulose-transition metal oxide mixture to separate the heavy metals from the liquid. The heavy metal and radioactive heavy metals precipitate from the liquid onto the cellulose material to form a radioactive metal-laden solid material. The radioactive metal-laden solid then is air-dried, calcined and/or admixed with a leach-resistant matrix, such as grout or asphalt, for suitable disposal.

Abstract: The present invention relates to a method and an apparatus for producing a waste package of radioactive waste containing particles of radioactive waste material of low modulus of elasticity, particles of radioactive waste material of high modulus of elasticity, and a solidifying agent in which the particles of radioactive waste material of low modulus of elasticity and the particles of radioactive waste material of high modulus of elasticity are fixed in an almost uniformly dispersed state. According to this invention, the radioactive waste generated from nuclear power plants can be greatly reduced in volume and also a waste package of radioactive waste with high strength and excellent water resistance can be obtained.

Abstract: A process for separately recovering uranium and hydrofluoric acid from a waste liquor containing uranium and fluorine comprises a neutralizing precipitation step wherein a magnesium compound is added to the waste liquor containing uranium and fluorine to form a precipitate and the thus formed precipitate is then separated; a distillation step wherein an aqueous solution of sulfuric acid is added to the precipitate separated in said neutralizing precipitation step to dissolve the precipitate and the thus formed solution is then distilled to recover hydrofluoric acid as a distillate; and a uranium recovery step wherein uranium is recovered from a residue produced by said distillation step.

Abstract: The invention relates to a process for decontaminating and adjusting the pH of uraniferous solutions to render them compatible with the natural environment into which they may be discharged. This process is characterized in that the solutions to be treated having a natural pH from about 2.5 to about 6.5 and containing from about 1 to about 100 mg/l of uranium, are supplemented with an aluminum salt, such as sodium aluminate, in a sufficient amount for the final pH to be from about 5.5 to about 8.5 and for there to be precipitation, coagulation and adsorption of about 90% of the uranium initially contained in the solution and for the uranium content remaining in the final solution obtained to be equal to or less than about 1.8 mg/l.

Abstract: A process for separating radium-226 from an aqueous liquid by adding a precipitant to the liquid. The liquid, including the precipitant, is passed upwardly through a particulate bed to assist co-precipitation and removal of the radium-226. The effect is to produce a fluidized bed of the particulate bed. The process is efficient, rapid and undemanding of space, all advantages not found in the prior art.

Abstract: A process for denitrating aqueous, nitric acid and salt containing waste solutions in which actinides are present, with simultaneous separation of the actinides, comprises agitating the waste solution at room temperature with diethyl oxalate and heating the resulting suspension to at least 75.degree. C.

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: Method for precipitating and removing soluble metal compounds from solutions of phosphoric acid. The method is useful in the disposal of metal-containing phosphoric acid waste from electrolytic operations, including such acid solutions contaminated with uranium compounds.

Abstract: A process for treating a radioactive liquid waste is disclosed, in which a radioactive liquid waste containing uranium and .beta.-decay nuclides, daughter nuclides of uranium, is treated by combination of a flocculation method using water glass as pretreatment and a subsequent ion exchange method. An approximately total amount of the uranium and a part of the .beta.-decay nuclides, daughter nuclides of uranium, in the liquid waste are captured by an amorphous silica precipitate formed by addition of the water glass and a remaining part of the .beta.-decay nuclides, daughter nuclides of uranium, is captured thereafter by the ion exchange treatment. The thus captured radioactive materials are respectively eluted from the filtered out precipitate, a radioactive solid waste, and the ion exchanger by acid treatment to be recovered as an acidic solution. Thus, the radioactive materials in the liquid waste are recovered approximately completely, thereby making the radioactivities of a final drain remarkably reduced.

Abstract: High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

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 process for removing dissolved radioactive materials from aqueous solution by incorporating lime in a sufficient dosage to adjust the pH of the water to greater than 11.0 while simultaneously adding a material from the group consisting of hydrogen peroxide, oxygen, ferrous sulfate, ferric chloride and potassium permanganate.

Abstract: A process for recovering uranium and/or thorium from a liquid containing uranium and/or thorium is disclosed, which comprises capturing the uranium and/or thorium in the liquid by an amorphous silica precipitate formed by adding water glass to the liquid, making the captured uranium and/or thorium eluted from the precipitate by acid-treatment, recovering the eluted uranium and/or thorium as an acidic solution, and regenerating the precipitate to water glass by use of an alkali metal hydroxide solution. Thus, the uranium and/or thorium can be recovered in high yield and the amorphous silica precipitate, that is, a formed radioactive solid waste can be remarkably reduced.

Abstract: A process for recovering uranium and/or thorium from a liquid containing uranium and/or thorium is disclosed, which comprises making a precipitate composed mainly of amorphous silica formed by adding water glass to the liquid containing uranium and/or thorium, capturing the uranium and/or thorium in the liquid by the precipitate, treating the precipitate with acid to elute the captured uranium and/or thorium, adding ammonia or hydrogen peroxide to an acidic solution containing the eluted uranium and/or thorium to form a precipitate composed of ammonium salt or peroxide of the uranium and/or thorium, and filtering out the ammonium salt or peroxide precipitate. Thus, high purity solid uranium and/or thorium reutilizable for manufacture of nuclear fuel material can be recovered in high yield.

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: Bidentate organophosphorous compounds are purified of undesirable impurities by contacting a solution of the compounds with a mercuric nitrate solution to form an insoluble mercuric bidentate compound which precipitates while the impurities remain in solution. The precipitate is washed and then contacted with a mixture of an aqueous solution of a strong mercuric ion complexing agent and an organic solvent to complex the mercuric ion away from the bidentate compound which then dissolves in the solvent. The purified bidentate compounds are useful for extracting the actinide elements from aqueous acidic nuclear waste solutions.

Abstract: Process waste waters at a pH of about 7 contaminated with radioactive isotopes are decontaminated by (a) adjusting the pH to about 5.8, (b) adding CaO or Ca(OH).sub.2 to raise the pH to about 8.5, (c) agitating the mixture for at least 5 minutes to effect intimate contact and produce a suspension of solids containing radioactive contaminants, and (d) separating the suspension of solids from the water by centrifuging. Removal of radioactive uranium isotopes with an alpha emission is effected at a pH of about 10. The process provides a method for concentrating radioactive contaminants in water for subsequent ultimate storage and also purifies the contaminated water so it may be safe to discharge it into the sewer. The treatment may be carried out in a plurality of stages in series.

Abstract: High salt content, alkaline waste solutions containing plutonium and americium are contacted with a sodium titanate compound to effect removal of the plutonium and americium from the alkaline waste solution onto the sodium titanate and provide an effluent having a radiation level of less than 10 nCi per gram alpha emitters.

Abstract: UO.sub.2 for nuclear fuel is made from UF.sub.6. The method involves injecting UF.sub.6, with or without a nitrogen carrier, into a solution containing 1) an inert reaction medium, 2) water, 3) a Lewis base. The precipitate from the above reaction is then reduced in H.sub.2 at a temperature below 750.degree. C. to give ceramic grade UO.sub.2.