Abstract: Improved methods for producing colloidal dispersions of cerium-containing oxide nanoparticles in substantially non-polar solvents are disclosed. The cerium-containing oxide nanoparticles of an aqueous colloid are transferred to a substantially non-polar liquid comprising one or more amphiphilic materials, one or more low-polarity solvents, and, optionally, one or more glycol ether promoter materials. The transfer is achieved by mixing the aqueous and substantially non-polar materials, forming an emulsion, followed by a phase separation into a remnant polar solution phase and a substantially non-polar organic colloid phase. The organic colloid phase is then collected.

Abstract: A method for forming spheroidal particles including inducing flow of a slurry of particles and a reactant through one or more orifices, detaching an amount of the slurry from the slurry flow following exit from the one or more orifices, the detached amount forming a slurry body, forming the slurry body into a spheroidal shape, contacting the spheroidally shaped slurry body with a coagulation solution to form a stabilized spheroidal particle and drying and/or sintering the stabilized spheroidal particle.

Abstract: A gel which contains a component of a catalyst is prepared. The gel and another gel are arranged in layers at a position which faces open ends of exhaust flow passages of a wall flow-type particulate filter. Next, a pusher is used to push and pack the gel and other gel through the open ends to the insides of the exhaust flow passages. Next, the particulate filter is made to dry to thereby make the partition walls carry the catalyst on their surfaces or in their micropores.

Abstract: Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

Abstract: Formulations useful for preparing hydrous hafnium oxide gels contain a metal salt including hafnium, an acid, an organic base, and a complexing agent. Methods for preparing gels containing hydrous hafnium oxide include heating a formulation to a temperature sufficient to induce gel formation, where the formulation contains a metal salt including hafnium, an acid, an organic base, and a complexing agent.

Abstract: Formulations useful for preparing hydrous cerium oxide gels contain a metal salt including cerium, an organic base, and a complexing agent. Methods for preparing gels containing hydrous cerium oxide include heating a formulation to a temperature sufficient to induce gel formation, where the formulation contains a metal salt including cerium, an organic base, and a complexing agent.

Abstract: Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

Abstract: Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI2, CuI, or Bi5O7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425° C. to 550° C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500° C. (below the silver iodide sublimation temperature of 500° C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

Abstract: This invention provides a dripping nozzle device to produce ADU particles with good sphericity, a device for recovering a feedstock liquid to prepare a uniform feedstock liquid, a device for supplying a feedstock liquid to form drops with a uniform volume, a device for solidifying the surfaces of drops so that the drops will not deform easily when they fall onto and hit the surface of an aqueous ammonia solution, a device for circulating an aqueous ammonia solution so that the uranyl nitrate in the drops can be changed to ammonium diuranate completely, to such an extent that uranyl nitrate in the center of each drop is changed to ammonium diuranate, and an apparatus for producing ammonium diuranate particles with good sphericity. The dripping nozzle device is provided with a single vibrator to vibrate nozzles simultaneously. The device for recovering a feedstock liquid recovers the feedstock liquid remaining in the nozzles and mixes it with a fresh feedstock liquid.

Abstract: Locally deposited polymer depots are used as a vehicle for the immobilization and local delivery of a radionuclide or radiopharmaceutical. Radionuclides are incorporated in their elemental forms, as inorganic compounds, or are attached to a larger molecule or incorporated into the polymer, by physical or chemical methods. Ancillary structure may be employed to control the rate of release. Standard radionuclides which have been used for local radiotherapy may be used, such as radionuclides of iodine, iridium, radium, cesium, yttrium or other elements.

Abstract: The present invention relates to a process for nuclear waste disposal. In it, a glass forming mixture including an aqueous solution of one or more metal alkoxides, alcohol, and solubilized, low level radioactive waste having a pH effective to hydrolyze the one or more metal alkoxides is formed. The one or more metal alkoxides in the glass forming mixture are converted to a network of corresponding one or more metal oxides. A gel is then formed from the glass forming mixture containing the network of one or more metal oxides. The gel is dried and sintered under conditions effective to form a densified glass.

Abstract: Pollution remedial compositions are disclosed which comprise a soluble silicate, a surfactant, a polyol, and water.

Abstract: A process of preparing a fuel pellet for a nuclear reactor, comprising washing the gel particle using an organic solvent miscible with water to substitute the organic solvent for the water, removing the organic solvent, moistening again the dry gel particle, followed by press molding and sintering.

Abstract: Process for preparing a radionuclide generator for producing Tc-99m or Re-188. A clear solution containing a metallic cation and an anion comprising W-188 or Mo-99 is provided. The metallic cation is present in the solution as a dissolved complex of the metallic cation and a complexing agent and/or the anion being present in the solution as a dissolved complex of the anion and a complexing agent. The dissolved complex(es) are decomposed to form a slurry containing a precipitate of the metallic cation and the anion. The precipitate is transferred to an elutable container of a radionuclide generator.

Abstract: A continuous denitration apparatus by microwave heating comprises an oven into which microwaves are applied and a cylindrical rotary drum disposed horizontally and rotatably disposed inside the oven. The rotary drum has the outer peripheral surface composed of a honeycomb or porous sheet-like microwave dielectric member. Inside the oven are disposed a trough for storing a nitrate solution of a nuclear fuel material below the rotary drum so as to immerse a lower portion of the outer peripheral surface of the rotary drum in the solution stored inside the trough, a heater above the rotary drum so as to heat the outer peripheral surface of the rotary drum opposing the heater, and a scraper for scraping off a denitrated product formed on the outer peripheral surface of the roatray drum between the heater and the trough.

Abstract: A ferric hydroxide colloid having an alpha-emitting radionuclide essentially on the outer surfaces and a method of forming same. The method includes oxidizing a ferrous hydroxide to ferric hydroxide in the presence of a preselected radionuclide to form a colloid having the radionuclide on the outer surface thereof, and thereafter washing the colloid, and suspending the washed colloid in a suitable solution. The labelled colloid is useful in cancer therapy and for the treatment of inflamed joints.

Abstract: A process for the embedding or containment of waste materials, especially of radioactive waste materials, by embedding in a ceramic matrix through the intermediary of the chemical precipitation of a hydrosol of the matrix, which is formed through the reaction of a nitrate with ammonia or ammonium hydroxide. In this process, the hydrosol is trickled in a pouring column into a precipitation bath, subsequent to the introduction of the material which is to be embedded, in which the hydrosol particles which are formed during the trickling gelatinize into gel particles. The gel particles are removed from the precipitation bath, washed, dried and finally sintered.

Abstract: A method 20 for decontaminating plastic products and materials which have become radioactively contaminated. The treatment method 20 involves dissolving such plastics in a dissolution tank 28 in an organic solvent and treating the resulting solution by a solvent extraction technique in column 36 to remove particulate and dissolved radioactive contaminants from the plastic. The contaminants can be buried in a low level radioactive waste site and the separated plastic material can be disposed of in a sanitary landfill or recycled into other plastic products.

Abstract: The invention relates to a process for improving the effectiveness of the removal of zirconium from an aqueous, zirconium-containing nuclear fuel and/or fertile material solution in nitric acid in a liquid-liquid extraction process.Accordingly, the object of the invention is to improve the removal of zirconium from the reprocessing solutions and, at the same time, to simplify the course of the process. The invention seeks to improve decontamination of the uranium product and the plutonium product while, at the same time, reducing the outlay involved.According to the invention, this object is achieved in that, in a process step carried out before the first extraction of the nuclear fuels and/or fertile materials, the zirconium is converted from the dissolved state into a filterable or centrifugable solid phase by the use of an adsorbent from the group of inorganic ion exchangers and is removed from the aqueous solution together with the adsorbent.

Abstract: Not readily dissolvable nuclear fuel is dissolved in an airtight closed vessel or autoclave. Nitric acid is evaporated and condensed in the vessel and the nitric acid condensate trickles over the nuclear fuel disposed outside of the nitric acid undergoing evaporation.

Abstract: An improved method of disposing of radioactive or hazardous liquid hydrocarbon compositions comprises placing the liquid hydrocarbons in a container and adding an organic ammonium montmorillonite in ratios of between about 1:2 and 3:1, by volume, respectively. The organic ammonium montmorillonite has at least 10 carbon atoms. A polar organic compound having between 1 and about 10 carbon atoms is also preferably added with the montmorillonite. Where the liquid waste material includes 5% or more of water in addition to the liquid hydrocarbon a minus 200 mesh sodium montmorillonite is also preferably added.

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: A stable acidic and alkaline metal oxide sol is formed by hydrolyzing a metal salt precursor in the presence of a stabilizing component chosen from the group consisting of surfactants having an HLB of at least 8, dispersing carboxylate polymers, and mixtures thereof, then alkalizing by adding a water-soluble amine compound, preferably diethylaminoethanol.

Abstract: N-(3-(3-.sup.125 iodo, 4-hydroxybenzyl)propionyl)phosphatidylethanolamine[.sup.125 I-phenylpropionyl-PE], a nonexchangeable iodinated lipid of high specific radioactivity is disclosed, with a method for its synthesis as well as the synthesis of other iodinated lipids using a technique involving acylation of the amine nitrogen of a phospholipid, such as phosphatidylethanolamine, with an [.sup.125 I] or [.sup.131 I] substituted cyclic organic acid, for example, monoiodinated Bolton-Hunter reagent and purification by thin layer chromatography. The iodinated lipid may be quantitatively incorporated into vesicles for use as an accurate liposome tracer in in vitro and in vivo assays that require high levels of sensitivity unobtainable with most other lipid-labeling techniques.

Abstract: A process of recovering cesium ions from mixtures of ions containing them and other ions, e.g., a solution of nuclear waste materials, which comprises establishing a separate source phase containing such a mixture of ions, establishing a separate recipient phase, establishing a liquid membrane phase in interfacial contact with said source and recipient phases, said membrane phase containing a ligand, preferably a selected calixarene as depicted in the drawing, maintaining said interfacial contact for a period of time long enough to transport by said ligand a substantial portion of the cesium ion from the source phase to the recipient phase, and recovering the cesium ion from the recipient phase. The separation of the source and recipient phases may be by the membrane phase only, e.g., where these aqueous phases are emulsified as dispersed phases in a continuous membrane phase, or may include a physical barrier as well, e.g.

Abstract: Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Abstract: A cyclical and automatable process and apparatus for continuously removing undesired components from solid particles by means of a solvent, the process including maintaining an expanded bed of solid particles in a vertical column, supplying a suitable solvent to the bed to flow through the bed either upwardly or downwardly in a selectably alternating manner, supplying solid particles to the top of the column while the solvent flows upwardly therethrough and interrupting the supply of solid particles and removing a bottom portion of the bed by the solvent when it flows downwardly.

Abstract: Sorbitol, or another polyalcohol such as erythritol, dulcitol or xylitol, is added to a solution containing uranyl nitrate which may also contain another heavy metal, such as throium or plutonium, prior to preneutralization with ammonia in order to provide a highly viscous solution that can be preneutralized to a great extent without premature precipitation of uranium. The high viscosity makes possible the formation of favorably large drops when the solution is dripped into an ammonia containing bath for external gelification of the drops. According to the pouring apparatus used, the particles after washing, drying and sintering have a diameter between 0.6 and 1.5 mm. The polyalcohol is added to a hydrosol containing from 1.5 to 2 moles per liter of heavy metal, the polyalcohol being added until concentration of one-third mole of polyalcohol per mole of heavy metal is reached. In certain cases up to four moles of ammonium nitrate per liter are added.

Abstract: Nuclear fuel is made by contacting an aqueous solution containing an actinide salt with an aqueous solution containing ammonium hydroxide, ammonium oxalate, or oxalic acid in an amount that will react with a fraction of the actinide salt to form a precipitate consisting of the hydroxide or oxalate of the actinide. A slurry consisting of the precipitate and solution containing the unreacted actinide salt is formed into drops which are gelled, calcined, and pressed to form pellets.

Abstract: Radioactive organic liquids are converted to a form suitable for burial by the process wherein the liquid is contacted with insoluble, swellable polymer particles to form swollen gelled particles which are dispersed in an unsaturated polyester, vinyl ester resin or mixture thereof which is then cured to a solid state with the gelled particles encased therein.

Abstract: An improved method of disposing of radioactive or hazardous liquid hydrocarbon compositions comprises placing the liquid hydrocarbons in a container and adding an organic ammonium montmorillonite in ratios of between about 1:2 and 3:1, by volume, respectively. The organic ammonium montmorillonite has at least 10 carbon atoms. A polar organic compound having between 1 and about 10 carbon atoms is also preferably added with the montmorillonite. Where the liquid waste material includes 5% or more of water in addition to the liquid hydrocarbon a minus 200 mesh sodium montmorillonite is also preferably added.