Abstract: An apparatus including a heating element and a sublimation vessel disposed adjacent the heating element such that the heating element heats a portion thereof. A collection vessel is removably disposed within the sublimation vessel and is open on an end thereof. A crucible is configured to sealingly position a solid mixture against the collection vessel.

Abstract: Provided is a method for improving the recovery rate of valuable metals such as cobalt when drying the battery waste of lithium ion batteries and the like. A second alloy excellent in terms of iron-cobalt separation performance and containing a small amount of iron is obtained by performing: a pre-oxidation step (ST20) for roasting and pre-oxidizing battery waste containing aluminum and iron; a melting step (ST21) for obtaining a molten product by melting the battery waste after the pre-oxidation step; a first slag separation step (ST22) for separating and recovering first slag containing aluminum oxide from the molten product; a second oxidation step (ST23) for oxidizing a molten first alloy after the first slag separation step; and a second slag separation step (ST24) for separating and recovering a second slag containing iron from a second alloy after the second oxidation step (ST23).

Abstract: A method of purification for recycling of gallium-69 isotopes includes processes of proton irradiation and dissolution for a silver alloy plating target with gallium-69. After the proton irradiation and dissolution, a high concentration elution liquid of gallium-69 and germanium-68 is obtained by washing through an ion-exchange resin to filter out gallium-69 solution, followed by neutralizing precipitation, drying, and sintering treatments to obtain a gallium oxide. The gallium oxide can be dissolved to produce a solid target, and the washing processes can be repeated. The solid target after use can be placed in recycling again. This method is not only implemented to reduce the cost of production and comply with recycling notion nowadays, but also enhance efficiency in the practical application of radioisotopes.

Abstract: A method for separating an amount of osmium from a mixture containing the osmium and at least one other additional metal is provided. In particular, method for forming and trapping OsO4 to separate the osmium from a mixture containing the osmium and at least one other additional metal is provided.

Abstract: A polymer composite with superior granulometric properties for the extraction of active and non-active cesium from high level acidic radioactive nuclear waste and/or other inorganic wastes/solutions that is particularly useful to nuclear industry. The void volumes of the said polymer composite is varied to obtain the desired Cs ion exchange kinetics wherein the composite material is radiation resistant and stable in highly acidic and alkaline medium.

Abstract: A method with which americium may be selectively recovered from a nitric aqueous phase containing americium, curium and fission products including lanthanides and yttrium, but which is free of uranium, plutonium and neptunium or which only contains these three last elements in trace amounts. The method is applicable for treatment and recycling of irradiated nuclear fuels, in particular for removing americium from raffinates stemming from methods for extracting and purifying uranium and plutonium such as the PUREX and COEX™ methods.

Abstract: The invention relates to novel compounds which are useful as ligands of actinides, to the synthesis of these compounds and to their uses. These compounds fit the general formula (I) hereafter: wherein R1 and R2, either identical or different, represent H, a linear or branched, saturated or unsaturated C1-C12 hydrocarbon group, a phenyl, benzyl, diphenyl or tolyl group; R3 represents H, a linear or branched, saturated or unsaturated C1-C12 hydrocarbon group, a phenyl, tolyl or linear or branched C1-C12 alkoxy group; while R4 represents H, a linear or branched, saturated or unsaturated C1-C12 hydrocarbon group, a phenyl or tolyl group. Field of applications: the processing of used nuclear fuels via a hydrometallurgical route.

Abstract: The present invention provides a method for fabricating an indium(In)-111 radioactive isotope. A target of cadmium(Cd)-112 is processed through steps of dissolving with heat, absorbing, washing, desorbing and drying for obtaining the In-111 radioactive isotope. Thus, chemical separation is coordinated with the target for fabricating the In-111 radioactive isotope with high efficiency and low cost for production procedure.

Abstract: The present invention describes a method for purification of 225Ac from irradiated 226Ra-targets provided on support, comprising a leaching treatment of the 226Ra-targets for leaching essentially the entirety of 225Ac and 226Ra with nitric or hydrochloric acid, followed by a first extraction chromatography for separating 225Ac from 226Ra and other Ra-isotopes and a second extraction chromatography for separating 225Ac from 210Po and 210Pb. The finally purified 225Ac can be used to prepare compositions useful for pharmaceutical purposes.

Abstract: A method for producing copper fine particles by heating and reducing an oxide, hydroxide, or salt of copper included in a solution of ethylene glycol, diethylene glycol, or triethylene glycol, the method comprising controlling a total halogen content of the solution to be less than 20 ppm by mass relative to copper and adding a water-soluble polymer as a dispersant such as polyethyleneimine and a noble metal compound or noble metal colloid for nucleation to the solution. This method makes it possible to provide copper fine particles for use in a wiring material, which are very fine as small as 50 nm or less in average particle size and high dispersibility, extremely low undesirable halogen content, and can be sintered at a low temperature.

Abstract: The present invention provides a method for fabricating an indium(In)-111 radioactive isotope. A target of cadmium(Cd)-112 is processed through steps of dissolving with heat, absorbing, washing, desorbing and drying for obtaining the In-111 radioactive isotope. Thus, chemical separation is coordinated with the target for fabricating the In-111 radioactive isotope with high efficiency and low cost for production procedure.

Abstract: An alteration of the traditional zinc/zinc-amalgam reduction procedure which eliminates both the hazardous mercury and dangerous hydrogen gas generation. In order to avoid the presence of water and hydrated protons in the working solution, which can oxidize Eu2+ and cause hydrogen gas production, a process utilizing methanol as the process solvent is described. While methanol presents some flammability hazard in a radiological hot cell, it can be better managed and is less of a flammability hazard than hydrogen gas generation.

Abstract: The present invention relates to a method for the generation of 223Ra of pharmaceutically tolerable purity comprising: i) preparing a generator mixture comprising 227Ac, 227Th and 223Ra in a first aqueous solution comprising a first mineral acid; ii) loading said generator mixture onto a DGA separation medium (e.g. resin); iii) eluting said 223Ra from said DGA separation medium using a second mineral acid in a second aqueous solution to give an eluted 223Ra solution; and iv) stripping the DGA separation medium of said 227Ac and 227Th by flowing a third mineral acid in a third aqueous solution through the DGA separation medium in a reversed direction; The invention further relates to high purity radium-223 formed or formable by such a method as well as pharmaceutical compositions comprising such radium-223 of pharmaceutical purity.

Abstract: The present relates to a method and a device for producing a radionuclide in which an absorption column containing the radionuclide is eluted by means of an eluent in a first flow direction and subsequently in a second, opposite flow direction.

Abstract: A multiple generator elution system for selectively eluting from a plurality of parent-daughter generators according to an elution schedule it calculates taking into account supply data, demand data, and available activity in each of the generators.

Abstract: The invention relates to novel compounds which are useful as ligands of actinides, to the synthesis of these compounds and to their uses. These compounds fit the general formula (I) hereafter: wherein R1 and R2, either identical or different, represent H, a linear or branched, saturated or unsaturated C1-C12 hydrocarbon group, a phenyl, benzyl, diphenyl or tolyl group; R3 represents H, a linear or branched, saturated or unsaturated C1-C12 hydrocarbon group, a phenyl, tolyl or linear or branched C1-C12 alkoxy group; while R4 represents H, a linear or branched, saturated or unsaturated C1-C12 hydrocarbon group, a phenyl or tolyl group. Field of applications: the processing of used nuclear fuels via a hydrometallurgical route.

Abstract: The invention relates to a process for producing 228Th from a natural thorium salt, which comprises in succession: a) the separation of the radium from the other radioelements present in this salt, by at least one coprecipitation of the radium by barium sulphate, this coprecipitation comprising: i) the addition of sulphuric acid and a barium salt to an aqueous solution of said natural thorium salt in order to form a barium-radium sulphate coprecipitate and ii) the separation of the coprecipitate from the medium in which it has formed; b) the extraction of the thorium 228 coming from the decay of radium 228 from the coprecipitate thus separated; and, optionally c) the purification and concentration of the 228Th thus extracted. Applications: manufacture of radiopharmaceutical products useful in nuclear medicine, in particular in radioimmunotherapy for the treatment of cancers and AIDS.

Abstract: A method using diglycolamide for increasing the separation factor between americium and curium and/or between lanthanides during an extraction operation. The operation comprising putting an acid aqueous phase, in which are found the americium, curium and/or lanthanides, in contact with an organic phase non-miscible with water, containing at least one extractant in an organic diluent. The aqueous and organic phases are then separated, and the diglycolamide is added to the aqueous phase.

Abstract: A method with which americium may be selectively recovered from a nitric aqueous phase containing americium, curium and fission products including lanthanides and yttrium, but which is free of uranium, plutonium and neptunium or which only contains these three last elements in trace amounts. The method is applicable for treatment and recycling of irradiated nuclear fuels, in particular for removing americium from raffinates stemming from methods for extracting and purifying uranium and plutonium such as the PUREX and COEX™ methods.

Abstract: A polymer composite with superior granulometric properties for the extraction of active and non-active cesimn from high level acidic radioactive nuclear waste and/or other inorganic wastes/solutions that is particularly useful to nuclear industry. The void volumes of the said polymer composite is varied to obtain the desired Cs ion exchange kinetics wherein the composite material is radiation resistant and stable in highly acidic and alkaline medium.

Abstract: The invention generally relates to the extraction of rare earth elements and heavy metals from geothermal fluids used in geothermal electrical production. The invention provides systems and methods for extracting these elements from hydrothermal products by the application of one or more forces that affect different components of a condensate differently.

Abstract: To use 99mTc as a raw material for a radioactive medicine, a very small amount of 99mTc in the high concentration Mo(99Mo) solution is purified and recovered with high yield without contamination of 99Mo. 99mTc with high purity is recovered by forming a high concentration Mo(99Mo) solution which contains radionuclides 99Mo which is the parent nuclide of 99mTc used for the radioactive medicine and the raw material for its labeled compound, forming a high concentration Mo(99Mo) solution which contains radionuclides 99Mo and 99mTc by generating 99mTc to a radioactive-equilibrium state, getting 99mTc in the high concentration Mo(99Mo) solution adsorbed to activated carbon selectively by feeding the solution to an adsorption column which has activated carbon, and undergoing desorption and purification treatment of 99mTc with a desorbent from the activated carbon to which 99mTc is adsorbed.

Abstract: A Gallium-68 (Ga-68) radioisotope generator includes a generating column and a citrate eluent. The generating column is at least partially filled with an ion-exchange resin with glucamine groups to absorb germanium-68 (Ge-68) and Ga-68 radioisotopes. The citrate eluent is added into the generating column to desorb the Ga-68 radioisotope and form an eluent containing the Ga-68 radioisotope in the form of Ga-68 citrate. A method for generating Ga-68 radioisotope is also disclosed.

Abstract: The invention provides a method of chemical recovery of no-carrier-added radioactive tin (NCA radiotin) from intermetallide TiSb irradiated with accelerated charged particles. An irradiated sample of TiSb can be dissolved in acidic solutions. Antimony can be removed from the solution by extraction with dibutyl ether. Titanium in the form of peroxide can be separated from tin using chromatography on strong anion-exchange resin. In another embodiment NCA radiotin can be separated from iodide solution containing titanium by extraction with benzene, toluene or chloroform. NCA radiotin can be finally purified from the remaining antimony and other impurities using chromatography on silica gel. NCA tin-117m can be obtained from this process. NCA tin-117m can be used for labeling organic compounds and biological objects to be applied in medicine for imaging and therapy of various diseases.

Abstract: A method for separating a lanthanide from a mixture containing at least one other lanthanide is provided. In particular, an HPLC and liquid separation method using a chromatographic column for separating a lanthanide from a mixture containing at least one other lanthanide is provided.

Abstract: One embodiment of the present invention includes a process for production and recovery of no-carrier-added radioactive tin (NCA radiotin). An antimony target can be irradiated with a beam of accelerated particles forming NCA radiotin, followed by separation of the NCA radiotin from the irradiated target. The target is metallic Sb in a hermetically sealed shell. The shell can be graphite, molybdenum, or stainless steel. The irradiated target can be removed from the shell by chemical or mechanical means, and dissolved in an acidic solution. Sb can be removed from the dissolved irradiated target by extraction. NCA radiotin can be separated from the remaining Sb and other impurities using chromatography on silica gel sorbent. NCA tin-117m can be obtained from this process. NCA tin-117m can be used for labeling organic compounds and biological objects to be applied in medicine for imaging and therapy of various diseases.

Abstract: The invention provides an apparatus for effectively removing ruthenium when ruthenium is removed from a solution containing platinum group metal by oxidation distillation. The invention provides an apparatus for selectively removing ruthenium from a solution containing ruthenium and other platinum group metal by adding an oxidizer to the solution to convert ruthenium into ruthenium tetroxide, wherein air is sucked into a reaction tank by reducing pressure within the reaction tank, and at least one outlet of the air is arranged in the apparatus such that the lowermost part thereof is located at the height of 5-20 m from the bottom of the reaction tank, whereby the solution within the reaction tank can be effectively stirred without ruthenium tetroxide, which has large specific gravity, being concentrated at the bottom of the reaction tank.

Abstract: The invention relates to a method for recovering gold from an acid digest of a gold-containing copper anode slime. The acid digest is selectively extracted with an alcohol having low miscibility with water. Gold is then recovered from the resulting alcoholic extract.

Abstract: A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution.

Abstract: A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution.

Abstract: The invention provides a method of chemical recovery of no-carrier-added radioactive tin (NCA radiotin) from intermetallide TiSb irradiated with accelerated charged particles. An irradiated sample of TiSb can be dissolved in acidic solutions. Antimony can be removed from the solution by extraction with dibutyl ether. Titanium in the form of peroxide can be separated from tin using chromatography on strong anion-exchange resin. In another embodiment NCA radiotin can be separated from iodide solution containing titanium by extraction with benzene, toluene or chloroform. NCA radiotin can be finally purified from the remaining antimony and other impurities using chromatography on silica gel. NCA tin-117m can be obtained from this process. NCA tin-117m can be used for labeling organic compounds and biological objects to be applied in medicine for imaging and therapy of various diseases.

Abstract: A method for separating and recovering trivalent and tetravalent actinoids in a simple and less costly manner without using an organophosphorus compound is provided. This method selectively separates and recovers the tetravalent actinoid plutonium Pu (IV) and the trivalent actinoids americium Am (III) and curium Cm (III) from trivalent lanthanoids Ln (III), etc. with the use of an extractant having a functional group with neutral multidentate ligand activity which is a hybrid donor type organic compound having both of donor atoms, i.e., an oxygen atom and a nitrogen atom.

Abstract: The present invention describes a method for purification of 225Ac from irradiated 226Ra-targets provided on a support, comprising a leaching treatment of the 226Ra-targets for leaching essentially the entirety of 225Ac and 226Ra with nitric or hydrochloric acid, followed by a first extraction chromatography for separating 225Ac from 226Ra and other Ra-isotopes and a second extraction chromatography for separating 225Ac from 210Po and 210Pb. The finally purified 225Ac can be used to prepare compositions useful for pharmaceutical purposes.

Abstract: A generator that allows for a non-fission based method of producing and recovering 99mTc from neutron-irradiated molybdenum. This generator system is based on the isolation of 99mTc, as the decay product from a source of 99Mo labelled molybdenum carbonyl Mo(CO)6 through a distillation process. The 99mTc obtained from this distillation is produced with high efficiency and purity in a solvent-free form, which can then be dissolved in water or other solvents to produce a solution at the required specific activity and concentration, as reasonably determined by the operator.

Abstract: This invention is provided for improvement of corrosion-resistant property of a crucible and for promotion of safety in a pyrochemical reprocessing method for the spent nuclear fuel. The spent nuclear fuel is dissolved in a molten salt placed in the crucible. In a pyrochemical reprocessing method, the nuclear fuel is deposited, and the crucible (2) is heated by induction heating. Cooling media (5, 6) are supplied to cool down, and a molten salt layer (7) is maintained by keeping balance between the heating and the cooling, and a solidified salt layer (8) is formed on inner wall surface of the crucible.

Abstract: Disclosed is a device for metallizing uranium oxide and recovering uranium, which reacts uranium oxide with a lithium metal to product uranium metal powder, and filters the resulting product using a porous filter to separate the uranium metal powder from lithium chloride molten liquid to recover the uranium metal powder. The device includes a heating furnace including at least one first heating unit, and a reactor includes a reaction vessel having a discharging valve hole located at the center of a bottom thereof and a conical bottom tapered to the discharging valve hole, a sealing lid for sealing the reaction vessel airtight, an argon gas inlet port for feeding argon gas into the reactor therethrough, and an argon gas outlet port for venting argon gas from the reactor therethrough. A valve assembly controls the discharging valve hole of the reaction vessel, and a plurality of agitators mix a mixture in the reactor.

Abstract: A calixarene dimer of the general formula (I-G), comprising a first calixarene moiety I and a second calixarene moiety G, wherein: L is [—CH2—] or [—O—CH2—O—] and is the same or different between each aryl group; R5 is H, NO2, halogen, or C1-C10 aliphatic hydrocarbyl group, C6-C20 aryl group, C6-C20 hydrocarbylaryl group, any of which is optionally substituted by one or more halo or oxo groups or interrupted by one or more oxo or amide groups, and R5 is the same or different on each aryl group; R1 comprises a carboxy group which is or is not protonated or protected; two groups out of R2, R3 and R4 are H; the one group out of R2, R3 and R4 not being H comprises at least one atom of one or more of O and S, the said at least one atom being capable of causing the calixarene to be adsorbed onto the surface of the substrate; and the one group out of R2, R3 and R4 not being H being conjugated to the second calixarene moiety G.

Abstract: An object of the present invention is to provide a method for preventing concentration of a radioactive substance in a generated extraction residue in a method of producing tantalum, niobium, or a similar substance including collecting and refining a raw material containing the substances through a fluoridation process by use of a hydrofluoric acid-containing solution. The object can be attained by employing an ingredient-regulated raw material prepared from an ore or a concentrate and, as an additive, a substance insoluble to hydrofluoric acid or a mixed acid containing hydrofluoric acid as an essential component; or by increasing the amount of the extraction residue through addition of the insoluble substance to a solvent during the fluoridation process, to thereby reduce the relative radioactive substance content to an arbitrary value.

Abstract: An object of the present invention is to provide a method for preventing concentration of a radioactive substance in a generated extraction residue in a method of producing tantalum, niobium, or a similar substance including collecting and refining a raw material containing the substances through a fluoridation process by use of a hydrofluoric acid-containing solution. The object can be attained by employing an ingredient-regulated raw material prepared from an ore or a concentrate and, as an additive, a substance insoluble to hydrofluoric acid or a mixed acid containing hydrofluoric acid as an essential component; or by increasing the amount of the extraction residue through addition of the insoluble substance to a solvent during the fluoridation process, to thereby reduce the relative radioactive substance content to an arbitrary value.

Abstract: The present invention provides a novel process for the removal and recovery of radionuclides from waste waters and process streams. The process of the present invention utilizes a combination of a supported liquid membrane (SLM) and a strip dispersion to improve extraction of the target species while increasing membrane stability and reducing processing costs. Additionally, the invention provides a family of new extractants, alkyl phenylphosphonic acids, for the removal and recovery of radionuclides and/or metals, including the use of the new extractants in the process. The new extractant selectively removes radionuclides and metals from the feed stream to provide a concentrated strip solution of the target species.

Abstract: The present invention provides a novel process for the removal and recovery of radionuclides from waste waters and process streams. The process of the present invention utilizes a combination of a supported liquid membrane (SLM) and a strip dispersion to improve extraction of the target species while increasing membrane stability and reducing processing costs. Additionally, the invention provides a family of new extractants, alkyl phenylphosphonic acids, for the removal and recovery of radionuclides and/or metals, including the use of the new extractants in the process. The new extractant selectively removes radionuclides and metals from the feed stream to provide a concentrated strip solution of the target species.

Abstract: This invention refers to a method for producing Actinium-225, comprising the steps of preparing a target (1) containing Radium-226, of irradiating this target with protons in a cyclotron and of chemically separating Actinium from the irradiated target material thereafter. According to the invention the proton energy in the cyclotron is adjusted such that the energy incident on the Ra-226 is between 10 and 20 MeV, preferably between 9 14 and 17 MeV. By this means the yield of production of the desired isotope Ac-225 is enhanced with respect to other radioisotopes.

Abstract: A method is disclosed for recovering and separating precious and non-precious metals from waste streams, which removes, separates, and recovers such metals in a cost effective manner with more than 95% removal form waste streams and with minimal amounts of unprocessed solids and sludge remaining in the environment. Metals such as chromium, manganese, cobalt, nickel, copper, zinc, silver, gold, platinum, vanadium, sodium, potassium, beryllium, magnesium, calcium, barium, lead, aluminum, tin; and the lie are removed and recovered from the waste streams with at least 95% removal and other metals and compounds, such as antimony, sulfur, and selenium are removed and recovered from waste streams with at least 50% removal. The method employs a unique complexing agent comprising a carbamate compound and an alkali metal hydroxide which facilitates the formation of the metals into ionic metal particles enabling them to be readily separated, removed and recovered.

Abstract: A method for recycling metal parts contaminated by radioactive elements, in particular by .alpha.-emitters, includes forming a melt and a slag from the metal parts and then separating the slag from the melt. The radioactive elements are oxidized prior to the formation of the melt and the slag. For that purpose, the contaminated metal parts are exposed to an oxygen-containing atmosphere for a period at a temperature below the melting temperature of the metal parts.

Abstract: A process for producing titanium slag which is low in radioactivity wherein molten titanium slag, produced by smelting ilmenite in the presence of a reductant in a DC electric arc furnace, is separated from molten iron, boron in an amount which is less than 2.5% equivalent B.sub.2 O.sub.3 of the slag is blended with the molten slag which thereafter is allowed to cool to form a glassy phase which contains the bulk of the radioactive elements of the slag before being crushed to particles below about 1 mm, whereafter the radioactive elements are leached to leave a titanium slag product which is low in radioactivity.

Abstract: A process for preparing uranium metal or alloy thereof suitable for use in a metal-based uranium enrichment plant or other use requiring a superdense metal comprising providing a molten metal bath containing the alloy metal and feeding uranium oxide and a reactive metal reductant into the molten metal bath so that the oxide is reduced to elemental uranium and alloying the thus formed uranium with the bath metal.

Abstract: A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product.

Abstract: A process for recovering palladium, rhodium and ruthenium from aqueous solutions deriving from the treatment of nuclear fuels and containing also iron and nickel, by reducing carbonylation with carbon monoxide at a pressure up to 1 atmosphere in a nitric acid solution at a pH of between 2 and 4 and at a temperature of between room and 100.degree. C. and reaction times of from 6 to 100 hours.

Abstract: A dual zone chemical reactor continuously processes metal-containing materials while regenerating and circulating a liquid carrier. The starting materials are fed into a first reaction zone of a vessel containing a molten salt carrier. The starting materials react to form a metal product and a by-product that dissolves in the molten salt that flows to a second reaction zone in the reaction vessel. The second reaction zone is partitioned from, but in fluid communication with, the first reaction zone. The liquid carrier continuously circulates along a pathway between the first reaction zone and the second reaction zone. A reactive gas is introduced into the second reaction zone to react with the reaction by-product to generate the molten salt. The metal product, the gaseous waste products, and the excess liquid carrier are removed without interrupting the operation of the reactor.

Abstract: A method for decontamination of steel components contaminated with radioactive material comprises the steps:(a) providing a mass of material including:(i) a proportion of steel carrying radioactive material; and(ii) a mass of slag forming material;(b) melting the mass of material, to provide a volume of molten steel and a volume of slag, the radioactive material originally present on the steel migrating to the slag; and(c) separating the slag from the molten steel. The mass of slag forming material is selected to provide a predetermined concentration of radioactive material in the slag. The concentration may be selected to be sufficiently dilute to allow disposal of the slag without restriction.