35 results about "Minor actinide" patented technology

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

The invention provides a method for reprocessing a spent nuclear fuel which can reprocess highly purified uranium without arranged with a step of separating Pu from solely from a spent nuclear fuel, and solely separated reprocessing of the reprocessed Pu is more difficult and has an excellent repression effect to nuclear proliferation. In a method for reprocessing a spent nuclear fuel by dissolving a spent nuclear fuel in an aqueous nitric acid solution and separating and recovering nuclides contained in the resulting fuel solution by solvent extraction comprising an electrolytic valence adjustment step by using the valence of plutonium contained in the fuel solution as a parameter until the valence of plutonium becomes 3 in which nuclides contained in the fuel solution is electrolyticallyreduced without removing fission products or minor actinides; a nuclide separation step in which, by using an extraction solvent which extracts uranium contained in the fuel solution, uranium is distributed from the fuel solution subjected to the electrolytic valence adjustment step to the extraction solvent.

The invention relates to a method for manufacturing a porous fuel including uranium, optionally plutonium, and at least one minor actinide, including the following consecutive steps: a) a step of compacting, into pellets, a mixture of powders including uranium oxide, optionally plutonium oxide, and at least one minor actinide oxide, at least a portion of the uranium oxide being in the form of triuranium octaoxide (U3O8), the other portion being in the form of uranium dioxide (UO2); and b) a step of reducing at least one portion of the triuranium octaoxide (U3O8) into uranium dioxide (UO2).

The invention discloses a preparation method of a kularite ceramic solidifying body. The preparation method is characterized in that Ce2(C2O4)3.10H2O, An2 (C2O4)3.10H2O / oxide of An (An=Gd, Pr and Eu) and NH4H2PO4 are taken as raw materials; the high-quality kularite ceramic solidifying body is prepared according to formulation designing and the steps such as fine grinding, drying, granulating, forming, adhesive removing and vacuum hot-pressing sintering at a lower temperature. The kularite ceramic solidifying body prepared according to the preparation method is high in heat stability, mechanical stability, chemical stability and anti-radiation stability, is applicable to the safe solidification treatment of high-level waste, particularly applicable to the safe solidification treatment of high-level minor actinides waste which is high in radioactivity and toxicity, and long in half life period, and also applicable to the engineering application of high-level waste ceramic solidification treatment needing remote operation.

A spent nuclear fuel is reprocessed by dissolving a spent nuclear fuel in an aqueous nitric acid solution and separating and recovering nuclides contained in the resulting fuel solution by solvent extraction. A spent nuclear fuel reprocessing method includes: an electrolytic valence adjustment step in which nuclides contained in the fuel solution is electrolytically reduced without removing fission products or minor actinides until valence of plutonium is at a level at which solvent extraction efficiency is low by using the valence of plutonium contained in the fuel solution as a parameter; and a nuclide separation step in which, by using an extraction solvent which extracts uranium contained in the fuel solution, uranium is distributed from the fuel solution subjected to the electrolytic valence adjustment step to the extraction solvent.

The invention discloses a method for separating element palladium from high-level waste, which includes the steps: adding concentrated nitric acid into nitrate solution of the high-level waste to adjust the nitric acid concentration to be 2mol / L; and allowing the high-level waste nitrate solution with the nitric acid concentration adjusted to pass through a chromatographic column filled with adsorbent to enable element palladium and sub-actinide elements to be adsorbed by the chromatographic column filled with the adsorbent; using nitric acid aqueous solution of thiourea to elute the chromatographic column with the adsorbed element palladium and sub-actinide elements, and eluting the element palladium out in a nitrate form, wherein the thiourea concentration of the nitric acid aqueous solution of thiourea is 0.2mol / L, and the nitric acid concentration is 0.1mol / L; and after the element palladium is eluted out in a nitrate form, using saltpeter solution with the pH (potential of hydrogen) being 6.5 to elute the chromatographic column with the adsorbed sub-actinide elements and eluting the sub-actinide elements out in a nitrate form. The method is simple, efficient, high in selectivity of the chromatographic column and effective in separation.

The invention discloses a method for separating an element Pd from the high-level radioactive waste, which comprises the following steps: (1) adding concentrated nitric acid in nitrate solution of the high-level radioactive waste which separates minor actinides for adjusting the concentration of the nitric acid to 1.7-2.3 mol / L; (2) leading the nitrate solution of the high-level radioactive wasteafter the adjustment of the concentration of the nitric acid to pass through a column filled with an absorbent, and leading the element Pd to be absorbed by the column filled with the absorbent; and (3) using nitric acid water solution of thiourea to elute the column filled with the element Pd, and eluting out the element Pd in the form of nitrate, wherein the concentration of the thiourea in thenitric acid water solution of the thiourea is 0.1-0.3 mol / L, and the concentration of the nitric acid is 0.05-0.15 mol / L. The method is simple and high-efficient, the selectivity of the column is high, and the separation effect is good.

The present invention discloses a method for preparing a fluorapatite ceramic solidified body. The method is characterized by comprising the steps of: taking and mixing 23-76 wt.% of calcium pyrophosphate, 8-26 wt.% of calcium fluoride, 0. 02-52wt.% of samarium oxide (europium oxide, gadolinium oxide), 0-15 wt.% of sodium carbonate, 0-16 wt.% of silica to obtain a mixture; grinding, drying, granulating, molding, discharging rubber and conducting vacuum hot pressing sintering on the mixture to obtain a fluorapatite ceramic solidified body. The method not only utilizes the phosphorus in high-level radioactive waste in a recycling way, but also can address the problem of low solubility of phosphorus and minor actinides nuclide in glass solidified body. The fluorapatite ceramic cured body prepared by the invention has excellent preparation geological stability, irradiation stability, mechanical stability, chemical stability and thermal stability, and can be used as a ceramic solidified body for safe treatment of minor actinides high-level radioactive waste with high toxicity, long half-life period and strong radioactivity.