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 invention belongs to the field of nuclear fuel reprocessing, and in particular relates to a device and a method for preparing tetravalent uranium by dynamic membrane electrolysis. The device consists of a cation exchange membrane, an anode, a cathode, anolyte, catholyte, a peristaltic pump, a gasket a, an electrode frame a, an anolyte outlet, a catholyte outlet, an anolyte inlet, a catholyte inlet, an electrode frame b and a gasket b, wherein the electrode frame a, the gasket a, the cation exchange membrane, the gasket b and the electrode frame b are arranged on the device from left to right and superposed layer upon layer to form a sealed space; each of the gasket a and the gasket b has a quadrilateral structure; the anode is a titanium-coated ruthenium iridium mesh; the cathode is a titanium mesh; the distance between the cathode and the anode is 6 to 15mm; and the cation exchange membrane is a perfluorosulfonated cation exchange membrane. By the device and the method for preparing the tetravalent uranium by dynamic membrane electrolysis, the content of the tetravalent uranium in a product is ensured, and the tetravalent uranium can be stably and quickly prepared; and the device has a simple structure.

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 provides a method for extracting technetium from a nuclear fuel post-treatment waste liquid. The method comprises the following steps of (1), adjusting nitric acid concentration of nitric acid waste liquid containing high technetium acid radical to 2 mol/L; (2), extracting the high-technetium acid radical ions in the waste liquid, and separating to obtain an organic phase; (3), washing the organic phase obtained in the step (2) by adopting an oxalic acid-containing nitric acid solution; and (4), reversely extracting technetium in the organic phase by using an alkaline aqueous solution to obtain a technetium product solution. According to the method, extraction and purification of technetium in the nuclear fuel post-treatment nitric acid waste liquid are effectively realized;the method for extracting technetium is simple to operate, and can be applied to a traditional PUREX flow post-treatment plant, the concentration and the concentration of nitric acid of technetium inthe treated object can be higher, the composition of the waste liquid can be relatively complicated, and the generation amount of a radioactive waste in the treatment process is smaller.

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 invention belongs to the technical field of nuclear fuel reprocessing and discloses an oxidizing valence-adjusting equipment for oxidizing Pu (III) to Pu (IV). The equipment mainly comprises an oxidizing tower, a feed liquid receiving tank and a NOX storage tank, wherein a feed liquid outlet is formed at the upper end of the oxidizing tower, a Pu (III) feed liquid inlet and a NOX feed inlet are formed at the lower end, and the feed liquid outlets at the upper parts of the feed liquid receiving tank and the oxidizing tower are connected by virtue of a pipeline; the main body of the oxidizing tower and packing in the tower are both radiation-resistant and nitric acid corrosion-resistant materials; a heating device is arranged on the outer tower wall of the oxidizing tower; the Pu (III) feed liquid enters the oxidizing tower from the Pu (III) feed liquid inlet at the lower part of the oxidizing tower; NOX gas provided by the NOX storage tank enters into the oxidizing tower from the NOX feed inlet at the lower part of the oxidizing tower separately or together with carrier gas; the valence-adjusted feed liquid enters the feed liquid receiving tank from the feed liquid outlet at the upper part of the oxidizing tower. The equipment is large in processing scale, high in NOx utilization ratio, safe and stable in valence-adjusting process and low in processing burden of nitric oxide tail gas.

The invention discloses a method for electrolytically extracting rare earth through liquid cathode molten salt and preparing lead rare earth alloy, and belongs to the technical field of nuclear fuel reprocessing. According to the technical scheme, lithium chloride, potassium chloride, and rare earth chloride are put into an alumina crucible, a mixture is subjected to temperature rise and is meltedto molten salt, a small crucible containing a lead ingot is put into the alumina crucible, and heat preservation is carried out; a cathode, a reference electrode and an auxiliary electrode are inserted into the molten salt to be connected with an electrochemical workstation; electrochemical workstation cyclic voltamogram and timing potential are used for measuring rare-earth reduction peak potential and alloy forming current; according to the reduction peak potential of the rare earth on the liquid-state lead electrode and the forming current of the lead rare earth alloy, constant potential electrolysis and constant current electrolysis are carried out for extracting the rare earth; the small crucible is cooled under the argon protection, an electrolysate is taken out, ethyl alcohol and deionized water are used for flushing, low-temperature drying is carried out, and the lead rare earth alloy is obtained. Compared with a method for extruding the rare earth through a solid electrode and obtaining the lead rare earth alloy, the method is short in flow process, and high in extraction rate, and the rare earth extraction rate is up to 97.2%.

The invention belongs to the technical field of nuclear fuel reprocessing and nuclear chemical engineering, and discloses a sandwich-type inert gas shielding hot chamber.The hot chamber mainly comprises a hot chamber body, a gas circulating and purifying device, an aerosol filtering device, an automatic pressure control system and a material feeding and discharging device, wherein the gas circulating and purifying device and an automatic pressure control system operation panel are located outside the hot chamber body.The hot chamber body mainly comprises an inert gas sealing shell, an anti-seismic sealing shell and a neutron and/or gamma ray shielding shell, wherein a gap is formed between the inert gas sealing shell and the anti-seismic sealing shell, the inert gas sealing shell is filled with high-purity inert gas, and an inert gas inner circulation inlet and an inert gas inner circulation outlet are formed in the diagonal corners of the inert gas sealing shell.The gas circulating and purifying device comprises a fan and a purifying column.The hot chamber has the advantages that insert gas can be controlled, radioactive aerosol can be efficiently filtered, and the requirements for low water and low oxygen can be met.

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 an online monitoring system for detecting hydrazine nitrate and hydroxylamine nitrate simultaneously, and belongs to the technical field of nuclear fuel post-treatment. The online monitoring system comprises a solution extraction system, a potential titration system and a system control center, and uses a standard sodium hydroxide solution to determine the hydrazine nitrate and hydroxylamine nitrate simultaneously, and when a detection result is beyond the detection range of (0.05-0.5)mol/L, the system samples the liquid to be tested in a multiplication or dilution way. Thus, the relative measurement error for concentration determination of hydrazine nitrate and hydroxylamine nitrate is lower than 1%, operation is simple, and the average consumption time of a sample is 60-100s. Compared with a present determination method, the system is characterized by high precision, short time, and simultaneous detection for the contents of hydrazine nitrate and hydroxylamine nitrate.

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 nuclear fuel reprocessing plant radioactive material radiation shielding design, and particularly relates to a nuclear fuel reprocessing plant plutoniumproduct transfer device. The device is used for containing plutonium product cups containing plutonium products in a sealed mode. A shielding layer comprises a first stainless steel shielding layer,a neutron shielding layer and a second stainless steel shielding layer which are sequentially arranged from inside to outside. The first stainless steel shielding layer is used for shielding alpha particles and part of beta rays and gamma rays emitted by the plutonium product cups. The neutron shielding layer is used for shielding neutrons emitted by the plutonium products. The second stainless steel shielding layer is used for shielding the gamma rays and bremsstrahlung photons and secondary photons generated after passing of the neutron shielding layer. By means of the device, neutron rays and the gamma rays can be greatly reduced, the neutron and photon radiation risk generated in the transferring process of the plutonium product cups is greatly reduced, and meanwhile the alpha particles and the beta rays can be completely shielded and the stability of the device is considered.

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 virtual reality technology, and particularly relates to a virtual reality system and method which can economically and truly simulate mechanical operation environment of a nuclear fuel reprocessing plant head end. The system comprises a server, a model processing unit, a data acquisition unit, an I/O data exchange transmission unit, display units, a touch screen and a virtual peripheral; the operation control instructions of the touch screen and the virtual peripheral are transmitted to the model processing unit through the data acquisition unit and the I/O data exchange transmission unit, the model processing unit carries out model optimization and action setting, and transmit and display the processed model data which are processed again through the server in the display units; the display units comprise a two-dimensional display unit and a three-dimensional display unit, and the display method of two and three dimensional integration has strong immersion and more accords with the real work condition. The virtual reality system and method for the nuclear fuel reprocessing plant head end has the advantages of being smart, convenient and fast, good in economy, small in space usage, and capable of being widely applied to the fields such as the facility design optimization, the equipment research and development and the staff training of the nuclear fuel reprocessing plant head end.

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.

The invention belongs to the technical field of post treatment of nuclear fuels, and discloses a method and device for oxidative adjustment of plutonium valence with NOx. According to the method, NOx and oxygen-containing gas are used as oxidants to destroy a reducing agent in Pu (III) liquid and oxidize Pu (III) into PU (IV); the device comprises a Pu (III) oxidation column and a nitrous acid dispelling column, wherein the upper part of the Pu (III) oxidation column is connected with the nitrous acid dispelling column through a liquid delivery pipe; the device is characterized in that a first exhaust discharge port and a second exhaust discharge port are formed in the top ends of the Pu (III) oxidation column and the nitrous acid dispelling column respectively, and an oxidant inlet and a Pu (III) liquid inlet are formed in the lower part of the Pu (III) oxidation column. The method and the device have the advantages of improving the utilization rate of NOx, stabilizing the pressure in the Pu (III) oxidation column and lightening the treatment burden of nitrogen oxide exhaust.