67 results about "Nuclear reprocessing" patented technology

Synthesis of water soluble half amide oxide is carried out by taking 1,4-dioxane as solvent at low-temperature, diethylene acid anhydride reacting with dimethylamine, diethylamine, and dipropylamine under pyridine, the products concentrating, chlorhydric acid acidifying, crystallizing precipitating, and annealing purifying in water or alcohol and diethyl ether mixed solvent. Its advantages include good water solubility, high output, and simple process.

The present invention uses alkyl-3-oxyalutaramine aqueous solution whose substituted alkyl carbon chain length is less than 3 carbons as back extractant, and utilizes the multistage cross current back extraction and/or continuous counter-current back-extraction process to separate and extract superuranium element and lanthanide from organic phase of tributyl phosphate and trialkyloxyphosphine in which the actinoid and lanthanide are extracted.

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 belongs to the speciality of nuclear fuel reprocessing, and discloses a plutonium purification and concentration method utilizing dioxime imide as a reduction and reextraction agent. The method comprises the two steps of extraction and reextraction. In the step of extraction, Pu(IV) and U(VI) in a salpeter solution are extracted to enter an organic phase 30%TBP-kerosene solution, so that uranium and plutonium are separated from other components. In the step of reextraction, dioxime imide is utilized as the reduction and reextraction agent, the salpeter solution containing dioxime imide is utilized as a water phase to quickly reduce Pu(IV) in an organic phase into Pu(III) to enter the water phase, and therefore plutonium purification and concentration are achieved, wherein the volume ratio of the organic phase to the water phase is 10:1-1:1. The method has the beneficial effects that the single-stage two-phase contact time is short, the amount of generated radioactive liquid waste is small, and solvent radiolysis damage is small.

The invention belongs to the technical field of nuclear fuel post-treatment, and relates to a uranium purification method for simultaneously removing neptunium and plutonium in the nuclear fuel Purex post-treatment process. The uranium purification method sequentially comprises the following steps that firstly, evaporation and concentration pretreatment is conducted, a uranium rough product from the Purex post-treatment procedure uranium plutonium codecontamination separating cycle is subjected to evaporation and concentration pretreatment, accordingly, Np(IV) in the uranium rough product is adjusted to Np (V) and Np (VI); secondly, a reducing agent is adopted for reducing, a concentrated solution obtained in the first step is cooled to the room temperature, then the reducing agent is added, then Pu (IV) and Pu (VI) in the concentrated solution are reduced into Pu (III), and Np (VI) is reduced into Np (V); and thirdly, organic solvent extraction is conducted, the acidity of the solution obtained in the second step is adjusted, extraction is conducted in an extractor with an organic solvent containing tributyl phosphate in a multi-level manner, wherein the organic solvent containing the tributyl phosphate is divided into multiple parts to be added at different extraction levels. By utilization of the uranium purification method, the Np and the Pu can be removed at the same time through one uranium purification cycle, and qualified uranium products are obtained.

The invention relates to a modified mesoporous silica adsorbent and a preparation method and application thereof. Under the catalysis of potassium carbonate and potassium iodide, aza-macrocyclic ligand is modified onto the surface of chlorine group functional mesoporous silica through nucleophilic substitution, and the adsorbent is prepared; palladium recycling and enrichment under different nitric acid concentrations are achieved through the adsorbent, and the selective recovery performance of palladium in highly-active effluent simulated feed liquid is evaluated. According to the modified mesoporous silica adsorbent, mesoporous nanomaterials are combined with the supermolecule recognition performance of the aza-macrocyclic ligand, the problems that selectivity is poor, operation is complex, the raw material cost is high and the stability is poor in the traditional treatment technology for palladium in highly-active effluent are solved, and the adsorbent has the advantages of being good in selectivity, large in adsorption capacity, high in adsorption speed and easy to recycle for palladium within the wide nitric acid concentration range, and has the great practical significance in the aspects of nuclear fuel reprocessing and hydrometallurgy.

The method for synthesizing water-soluble oxabisamide uses diethylene glycol as raw material, makes said raw material undergo the processes of oxidation and chloridization to obtain diethylene aryl chloride, then makes it react with dialkylamine to obtain bisamide, in the presence of pyridine uses 1,4-dioxane as solvent, and makes diethylene acyl chloride respectively react with dimethylamine or diethylamine or dipropylamine under the condition of low temperature, and makes the obtained product undergo the processes of filtration and aminolysis to make the product separated and purify so as to obtain oxabisamide compound.

The invention relates to a nuclear fuel reprocessing method in order to solve the problems that more devices are adopted, actual operation is complex, excessive nitric oxide is consumed, the economy is low and the like when a U/Pu separation section of the Purex process adopts nitric oxide for price adjustment, and provides a BP price-adjusting, degassing and acid-adjusting integrated device and method in a Purex process. The device comprises a filler column and a feed liquid receiving tank, wherein a BP feed liquid inlet and an NO2 feed port are formed in the filler column, a tail gas outlet is formed in the top, and a feed pipe extending to the feed liquid receiving tank is arranged on the feed liquid receiving tank. The method comprises steps of Pu price adjustment, nitrous acid removal and acid adjustment. According to the BP price-adjusting, degassing and acid-adjusting integrated device in the Purex process, nitric oxide price adjustment equipment is successfully simplified, the overall design of the adopted technological process is remarkably optimized, the nitric oxide consumption and the tail gas treatment quantity are remarkably reduced, the economy is remarkably improved, and a large quantity of costs is saved.

The invention belongs to the technical field of nuclear fuel post-treatment and discloses a carbon-based supramolecular recognition material and a preparation method thereof and application of the carbon-based supramolecular recognition material in adsorption and separation of Cs. A polymer is polymerized to millimeter-level macroporous-mesoporous carbon spheres to form a carbon-based carrier; thepolymer in a carbon-based carrier has mass percentage of 0-10%; the carbon-based carrier and calix[4]-arene-biscrown-6 are used as raw materials subjected to vacuum solvent evaporation induced self-assembly to prepare the carbon-based supramolecular recognition material; the mass of the carbon-based carrier in the carbon-based supramolecular recognition material is 0.5-40 times that of calix[4]-arene-biscrown-6; the calix[4]-arene-biscrown-6 has a structural formula shown as in (1). The carbon-based supramolecular recognition material disclosed herein has the structure well retained in a strong acid solution, has excellent adsorption selectivity for cesium ions, and is applicable to high-selectivity adsorption separation of cesium from strong-acidity high-emission waste liquid.

The invention belongs to the technical field of liquid-phase sampling in a hot cell extraction process in the field of nuclear fuel after-treatment and discloses an instantaneous sampling apparatus for a radioactive bench test. The apparatus is arranged on the bottom of a radioactive bench and includes a sampling pipeline, a first three-way valve, a second three-way valve, an exhaust pipe, a sampling bottle bracket, a fixing sleeve tube, a fixing sleeve tube supporter and a sampling bottle bracket handle. The sampling bottle bracket has a drawable structure and is provided with a sampling bottle trough and the sampling bottle bracket handle. The fixing sleeve tube supporter is arranged above the sampling bottle bracket. The fixing sleeve tube is welded to the fixing sleeve tube supporter. The fixing sleeve tube is arranged on a corresponding position above the sampling bottle trough and can be matched with and connected to the sampling pipeline. The sampling pipeline includes an organic phase sampling tube and a water phase sampling tube. The sampling apparatus has simple structure and is easy to operate, saves space on the bench, enables the bench to be hoisted conveniently, and can effectively avoids retention and overflow of a sample liquid.

The invention belongs to the technical field of radiation shielding design of radioactive materials in a nuclear fuel post-processing plant, and particularly relates to a material product cup in the nuclear fuel post-processing plant for storing radioactive materials extracted by the nuclear fuel post-processing plant. The material product cup includes an outer cup body (1) which is arranged at the outer layer and can be sealed by an outer cup cover (2) and an inner cup body (3) which is arranged inside the outer cup body (1) and can be sealed by an inner cup cover, wherein the inner cup body(3) is used for storing the radioactive materials, alpha particles, beta particles and other charged particles generated by the radioactive materials in the inner cup body (3) can be shielded completely through the sealed outer cup body (1) and the sealed inner cup body (3), and heat generated by the radioactive materials can be transferred. A variety of radioactive rays radiated by the radioactive materials can be shielded through the product cup, the irradiation dose of bremsstrahlung photons to workers can be reduced, the contact dose rate of the surface of the product cup can be reduced by15-40%, and the operation cost of the nuclear fuel post-processing plant is reduced.

The invention provides an analyzing method of hydroxysemicarbazide in a nuclear fuel post-treatment process. The method includes the following steps: (1) preparing a series of standard aqueous solutions, preparing multiple volumetric flasks for holding a color developing agent and a buffer solution, preparing a first blank sample and a series of standard samples, measuring the absorbance value ofeach standard sample at a same feature absorption peak by using an ultraviolet-visible spectrophotometer by taking the first blank sample as a reference, and making a standard curve of absorbance varying with the concentration of the hydroxysemicarbazide; and (2) taking a to-be-measured sample, removing plutonium, preparing multiple volumetric flasks for holding the color developing agent and thebuffer solution, preparing a second blank sample and a to-be-measured sample, measuring a corresponding absorbance value by using the ultraviolet-visible spectrophotometer by taking the second blank sample as a reference, and substituting the corresponding absorbance value into the standard curve to calculate the concentration of the hydroxysemicarbazide in the sample.

The invention belongs to the technical field of post-processing of nuclear fuel, and in particular relates to an ion chromatography analysis method for trace amounts of oxalic acid in a high-concentration nitric acid solution. Instrument detection and analysis, wherein, before the above-mentioned washing step, solid NaHCO3 is used to pretreat the destroyed plutonium (IV) oxalate precipitation mother liquor, so that the pH value is 6 to 8; the composition of the eluent system is: Na2CO3, NaHCO3 and H3BO3, wherein the concentration of Na2CO3 is 2-10mmol/L, the concentration of NaHCO3 is 2-10mmol/L, and the concentration of H3BO3 is 2-10mmol/L. The ion chromatography analysis method can accurately analyze trace amounts of oxalic acid in high-concentration nitric acid solution.

The invention belongs to the technical field of nuclear fuel post-treatment, and particularly relates to a device and method for continuously generating iodine vapor. The device for continuously generating the iodine vapor includes a U-shaped pipe (2), wherein one end of the U-shaped pipe is connected to a working gas input device, and the other end is used as a gas output end; two ends of the U-shaped pipe (2) face upwards, and a U-shaped section of the U-shaped pipe (2) is arranged in a heating device; and the U-shaped section is filled with a mixture composed of iodine powder and a porous material, and the mixture can generate the iodine vapor after being heated. The device and method provided by the invention solve the problem of agglomeration caused by interaction between iodine molecules due to a too high local iodine concentration during the production of iodine vapor, and can generate the continuous and stable iodine vapor and ensure the stable operation of a simulated iodine-containing tail gas experiment.

The invention relates to the use of butyraldehyde oxime as an anti-nitrous agent in a plutonium stripping operation based on a reduction of this element from oxidation state (IV) to oxidation state (III).

Applications: any nuclear fuel reprocessing process in which employing a compound that has the twofold property of being extractable into an organic phase and of being capable of destroying the nitrous acid therein may be useful, and especially any process including one or more operations for the reductive stripping of plutonium.