205 results about "Uranium oxide" patented technology

Improved methods for the extraction or dissolution of metals, metalloids or their oxides, especially lanthanides, actinides, uranium or their oxides, into supercritical solvents containing an extractant are disclosed. The disclosed embodiments specifically include enhancing the extraction or dissolution efficiency with ultrasound. The present methods allow the direct, efficient dissolution of UO2 or other uranium oxides without generating any waste stream or by-products.

In a method of producing large-grained nuclear fuel pellet, Cr-compound contained in an uranium oxide green pellet is reduced to Cr phase at 1,470° C. or below and maintained to the Cr phase, and the uranium oxide green pellet containing the Cr-compound is then sintered at 1,650° C.-1,800° C. in a gas atmosphere of oxygen potential at which Cr element in the uranium oxide green pellet becomes liquid phase.

The invention relates to a process for reprocessing a spent nuclear fuel and for preparing a mixed uranium-plutonium oxide, which process comprises: a) the separation of the uranium and plutonium from the fission products, the americium and the curium that are present in an aqueous nitric solution resulting from the dissolution of the fuel in nitric acid, this step including at least one operation of coextracting the uranium and plutonium from said solution by a solvent phase; b) the partition of the coextracted uranium and plutonium to a first aqueous phase containing plutonium and uranium, and a second aqueous phase containing uranium but no plutonium; c) the purification of the plutonium and uranium that are present in the first aqueous phase; and d) a step of coconverting the plutonium and uranium to a mixed uranium/plutonium oxide. Applications: reprocessing of nuclear fuels based on uranium oxide or on mixed uranium-plutonium oxide.

A method of declogging at least one filter of a plant for manufacturing uranium oxide from uranium hexafluoride, including separating, from the wall of the filter, uranium oxyfluoride particles deposited, by a stream of inert gas such as nitrogen, injected into the filter, in a counter-currentwise direction to the flow of hydrofluoric acid.

Disclosed is a method of in-situ monitoring a reduction process of uranium oxides by lithium metal, wherein a conversion yield of uranium metal from uranium oxides upon production of uranium metal through a reaction of uranium oxides (UO_x, x≦3) with lithium metal in the presence of a high-temperature molten salt is measured according to an electrochemical analysis based on an oxidation of an oxygen ion and a reduction of a lithium ion dissociated from lithium oxide obtained as a by-product of the reaction, by use of a measuring device composed of a potentiostat/galvanostat and a reactor provided with an anode and a cathode. The in-situ monitoring method of the current invention is advantageous in terms of fast and simplified measuring techniques, by directly measuring the reduction process of uranium oxides at the anode and cathode connected to the potentiostat/galvanostat in the presence of the high-temperature molten salt.

The invention belongs to the technical field of dissolution and purification treatment technologies for ammonium biuranate in the uranium purification and conversion process, in particular to the technologies of dissolution of the ammonium biuranate, extraction and purification treatment of a uranyl nitrate solution and the like and provides a uranium purification method for the ammonium biuranate. The uranium purification method for the ammonium biuranate comprises the following steps that (1) dissolution is conducted so as to prepare a qualified uranyl nitrate solution for extraction and purification; (2) extraction is conducted, the solution to be extracted and an extraction agent are mixed and conduct matter transfer in a pulse extraction column, uranium in the material is carried into the extraction agent, and an organic phase containing the uranium is obtained; (3) washing is conducted, the organic phase containing the uranium and a washing agent are fully mixed and conduct matter transfer in a pulse washing column, impurity elements in the organic phase containing the uranium are rewashed and enter a water phase, the purity of the uranium is further improved, and the purified organic phase containing the uranium is obtained; and (4) reverse extraction is conducted, the purified organic phase containing the uranium and the washing agent are fully mixed and conduct matter transfer in the pulse washing column, the uranium enters the water phase again through reverse extraction, the purified uranyl nitrate solution is obtained, and an uranium oxide is prepared from the purified uranyl nitrate solution through concentration and denitration manners.

The invention discloses a method of preparing a ceramic grade uranium dioxide ball. The method sequentially comprises the following steps: carrying out ball milling on uranium oxide, wherein the particle size of the uranium oxide after ball milling is 0.1-0.3 micron; adding the uranium oxide into a polymer monomer, and uniformly stirring to obtain polymer slurry; producing slurry drops by virtue of the polymer slurry; adding the slurry drops into a dispersion column which contains heated carbon tetrachloride or liquid paraffin, and descending the slurry drops to obtain a green body ball in the dispersion column under the surface tension, wherein the temperature in the dispersion column is 80-90 DEG C; washing the green body ball by adopting an ethanol solution; drying the green body ball, and removing an organic matter on the surface of the green body ball to obtain a dry calcined ball; and carrying out reduction sintering on the dry calcined ball in a hydrogen atmosphere to obtain the ceramic grade UO2 ball. The method is simple in whole process, convenient in realization, small in introduced impurities, high in purity of the prepared UO2 ball and beneficial to popularizing and applying the UO2 ball in the field of nuclear technology.

The invention belongs to the technical field of uranium ore water smelting technology, and specifically adopts anion exchange resin to adsorb D382, loaded resin for secondary adsorption (uranium-loaded resin absorbs uranium in the leaching qualified liquid), ammonium chloride plus sodium carbonate leaching It is a process for recovering uranium from the weakly alkaline leaching solution with high chlorine and high salinity by washing and heating precipitation. It effectively solves the problem of effectively adsorbing uranium from the weak alkaline leaching solution with high chloride ion and high salinity as high as 6g.L-1 and about 10g.L-1 salinity. Rinse the loaded resin with NH4Cl+Na2CO3 solution to obtain a satisfactory rinsing effect. The secondary adsorption of the loaded resin (the loaded resin is in contact with part of the eluting qualified solution) effectively increases the uranium concentration of the eluting qualified solution. After the eluting qualified solution is heated to drive off ammonium and carbon dioxide, uranium oxide precipitates are obtained. The precipitation process reduces the consumption of reagents and fresh water, simplifies the precipitation process, and at the same time obtains uranium products with good filtration and dehydration performance. The product obtained under the test conditions contains 79.96% uranium, 0.046% chloride ion and 2.91% sulfate radical.

The present invention relates to a SIMS measurement method for oxygen isotope in an insulator core material. The process comprises: transferring uranium oxide micro-particles onto a graphite carbon sheet to prepare a sample; debugging equipment required during the measurement process; measuring oxygen isotope of the uranium oxide micro-particles, and calculating the <18>O/<16>O ratio; correcting the measurement value, and calculating the uncertainty; and replacing the standard sample and uninstalling the program. According to the present invention, the chemical treatment of the sample is not required, and the measurement can be directly performed, such that the analysis speed is rapid, the sample preparation is simple, and the sample consumption is low; SIMS has the micro-zone analysis and depth profiling capability, and can perform oxygen isotope measurement on different regions and places having different depths of the sample; and SIMS performs measurements by receiving O<-> without conversion of CO2 gas so as to reduce the intermediate link, reduce the introduction factor of the final result uncertainty, and meet characteristics of rapid and accurate analysis of the nuclear forensics.

The invention relates to a process for reprocessing spent nuclear fuel which, among other advantages, does not require a plutonium-reducing stripping operation.

This process finds particular application in the processing of uranium oxide fuels and uranium and plutonium mixed oxide fuels.

The invention relates to a soft-light colored glass, and a preparation method and a use thereof. The soft-light colored glass is prepared through using the following raw materials, by weight, 50-70 parts of silica powder, 10-16 parts of sodium carbonate, 12-22 parts of feldspar powder, 10-24 parts of aluminum hydroxide, 1-12 parts of zinc oxide, 4-24 parts of calcium carbonate, 1.4-12 parts of borax, 8-16 parts of sodium fluosilicate, 0.8-5.8 parts of potassium carbonate, 0.4-4 parts of sodium nitrate and 6.6-70 parts of an auxiliary material, and the auxiliary material comprises several selected from titania, barium carbonate, red lead, calcium phosphate, barium sulfate, magnesium oxide, fluorite, arsenic trioxide, antimony oxide, cobalt oxide, copper oxide, chrome oxide green, sodium antimonate, cadmium yellow, sulfur, uranium oxide, cerium oxide, cadmium red, cadmium sulfide, selenium powder and yttrium aluminum garnet. The soft-light colored glass can solve a blue light polluting problem and has a long service life, and required raw materials do not pollute the environment, and the soft-light colored glass can be directly processed to form a lamp tube, a lamp cover or a lamp for use and avoids the easy color change and shedding problems of a coat of coated glass.

The invention provides a method for preparing UO2. The method comprises the following steps: firstly, fully and uniformly mixing U3O8 and NH4HF2 powder, and putting the mixed powder at the bottom of acrucible; then, laying LiCl-KCl salt on the mixed powder to cover the mixed powder; carrying out heating to 500 DEG C to melt the salt and keeping the temperature for 3 hours; and after the U3O8 is fully reacted and dissolved, carrying out constant-potential electrolysis by using a potential of -0.8 v (vs Ag/AgCl) with a molybdenum sheet as a cathode, a graphite rod as an anode and Ag/AgCl as a reference electrode so as to prepare UO2, wherein the result of ICP-AES analysis calculation shows that the extraction rate of uranyl ions after electrolysis for 8 hours reaches 98.5%. On one hand, theinvention provides the method for preparing UO2 by purifying uranium oxide through molten salt electrolysis; and on the other hand, HF gas generated in the process of reaction dissolution can be combined with oxygen ions to remove the oxygen ions in the molten salt in the form of water vapor.

The invention discloses an uranium oxide micro-sphere and a preparation method, a porous structure is provided in the uranium oxide micro-sphere, and the porous structure is composed of nano particles. The invention also comprises a preparation method of the uranium oxide micro-sphere, which comprises the following steps: 1)performing a hydro-thermal reaction on a mixing aqueous solution to obtain a sea urchin-state uranium-containing micro-sphere precursor, wherein the mixing aqueous solution comprises a uranyl nitrate aqueous solution, glycerol and urea; 3)washing and drying the uranium-containing micro-sphere precursor; and 3)placing the dried uranium-containing micro-sphere precursor in a tubular furnace for pyrolysis to obtain the uranium oxide micro-sphere. The preparation method has the advantages of simple process and easy control, and the prepared uranium oxide micro-sphere has the characteristics of small size and has porous structure, and has latent application value in the relative fields of catalysis and nuclear energy.

A method of fabricating metallic fuel from surplus plutonium may include combining plutonium oxide powder and uranium oxide powder to obtain a mixed powder with reduced proliferation potential. The mixed powder may be electroreduced in a bath of molten salt so as to convert the mixed powder to a first alloy. The first alloy may be pressed to remove a majority of the molten salt adhered to the first alloy to form a pressed alloy-salt mixture. The first alloy may be isolated from the salt by melting the pressed alloy-salt mixture. The first alloy may be further processed to fabricate a fuel rod. Accordingly, the metallic fuel produced may be used in a fast reactor system, such as a Power Reactor Innovative Small Module (PRISM).

To furnish stably a sintered nuclear fuel compact of uranium dioxide with a large grain diameter, a fabrication method therefor requires sintering a starting material at two stages in an oxidizing atmosphere and in a reducing atmosphere at relatively low temperatures. The starting material is either a powder mixture of uranium dioxide fresh material powder and triuranium octaoxide in a weight ratio of 75/25 ˜ 55/45 or pulverulent mixed uranium oxides yielding a final O/U ratio equal to that of the powder mixture. The sintering process in the oxidizing atmosphere involves at least one time period of a temperature rise to a higher temperature of 1200 to 1100° C. and at least one time period of a temperature drop to a lower temperature of up to 1080° C. A temperature difference between the higher and lower temperatures is preferred to be 50° C. or more.

A method for manufacturing a porous fuel comprising uranium, optionally plutonium and at least one minor actinide is provided. The method may comprise the following successive steps: a) a step for compacting as pellets a mixture of powders comprising uranium oxide, optionally plutonium oxide and at least one oxide of a minor actinide, at least one portion of the uranium oxide being in the form of triuranium octaoxide U₃O₈, the other portion being in the form of uranium dioxide UO₂; b) a step for reducing at least one portion of the triuranium octaoxide U₃O₈ into uranium dioxide UO₂.