69 results about "Fuel reprocessing" patented technology

The invention relates to a crown ether covalent organic framework material as well as a preparation method and application thereof. The preparation method comprises the following steps: dissolving a crown ether monomer and an amino monomer in an o-dichlorobenzene solvent, adding a catalyst acetic acid aqueous solution, and uniformly dispersing to form a reaction solution; the method comprises the following steps: freezing a reaction solution through a liquid nitrogen bath, vacuumizing, filling nitrogen, carrying out a heating reaction, and carrying out post-treatment operations such as washing and drying on a reaction product, so as to obtain the crown ether covalent organic framework material BADT-DB18C6-COF. The crown ether covalent organic framework material is formed by condensation of a crown ether aldehyde monomer and an amino monomer, the crown ether monomer is 4, 4 ', 4' ', 4 '''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde, and the crown ether monomer is 4, 4 ', 4' ', 4'''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde. The amino monomer is 4 ', 5'-bis (4-aminophenyl)-[1, 1 ': 2', 1 ''-terphenyl]-4, 4''-diamine. The crown ether covalent organic framework material BADT-DB18C6-COF prepared by the invention is mainly used as an adsorbent, is used for adsorbing and recovering uranium, thorium and americium ions from a solution, shows good adsorption capacity on radionuclides U-238, Th-232 and Am-243, and is particularly suitable for the field of spent fuel post-treatment in the nuclear industry.

The present application relates to the technical field of spent fuel reprocessing, in particular to a 1AF metering tank gas damper adjusting device and method, the 1AF metering tank gas damper adjusting device comprising a fixed rack and a plurality of stirring mechanisms, the fixed rack being arranged directly above the 1AF metering tank, and the stirring mechanisms being movably arranged on one side of the fixed rack close to the 1AF metering tank. Through the adjustment of the fixed rack and the stirring mechanisms, the horizontal position and the vertical position of the stirring mechanisms can be quickly adjusted, and then the stirring effect of the gas damper in the 1AF metering tank with different number and arrangement position of the gas damper can be tested, so that the number and position arrangement of the gas damper in the 1AF metering tank can be confirmed and fixed, and the type, number and position arrangement of the stirring device in the 1AF metering tank with 1:1 scale of the commercial spent fuel reprocessing plant for the accurate metering of the 1AF liquid volume are provided with reference, the traditional fixed welding method is avoided, and the manpower, material resources and financial resources for the 1AF metering tank gas damper adjustment test are saved.

The utility model discloses an electrochemical test experimental device for boiling nitric acid corrosion in a high-temperature reduced-pressure environment, which relates to the technical field of electrochemical test and consists of a reduced-pressure boiling corrosion reaction system, an electrochemical test system, a heating system, an acid gas absorbing and drying system, a vacuum degree control system and an automatic integrated control system. The decompression boiling corrosion reaction system is connected with the acid gas absorption and drying system, the vacuum degree control system and the heating system are respectively connected with the acid gas absorption and drying system and are in contact with the decompression boiling corrosion reaction system, and an automatic integrated control system is adopted to control the experimental device. The experimental device disclosed by the utility model can test and simulate the electrochemical corrosion behavior of high-temperature reduced-pressure boiling nitric acid in the environment of the evaporator for spent fuel post-treatment, and can realize weight loss and electrochemical test of various materials under the condition of reduced-pressure boiling strong corrosive solution at the same time.

The application discloses a check valve and a pipeline system of a nuclear fuel reprocessing plant. The check valve comprises a valve body, a valve clack, an elastic rod and a control assembly. The valve body is installed in a pipeline and is internally provided with a fluid cavity. One end of the valve body is provided with an inlet, and the other end is provided with an outlet. The inlet faces the upstream of the valve body, and the outlet faces the downstream. The inlet and the outlet are both communicated with the fluid cavity. The valve clack is located in the fluid cavity of the valve body, and the upper end surface of the valve clack is attached to the inner wall of the upper end of the fluid cavity to block the inlet. One end of the elastic rod is rotationally connected with the valve body, and the other end is rotationally connected with the valve clack. The elastic rod drives the valve clack to move in the vertical direction through swinging. The control assembly is electrically connected with the elastic rod, is used for monitoring the rotation angle of the elastic rod, and judges the vertical position of the valve clack according to the rotation angle information of the elastic rod to obtain the opening degree information of the valve. The check valve has the function of remotely monitoring the opening degree of the valve.

The application discloses a method for reducing the content of plutonium and neptunium in a uranium product solution in a spent fuel reprocessing PUREX process, which comprises the following steps: adjusting the 2DF material liquid, washing the extraction organic phase by using a washing agent with complexing and reducing double effects in an extraction device, and reducing the content of neptunium and plutonium in the uranium-containing organic phase, so as to obtain a uranium product solution with greatly reduced content of plutonium and neptunium through back extraction. 2 The method has the advantages of simple process, and the decontamination factor of plutonium and neptunium is 10 4 orders of magnitude after 2D tank extraction and washing, and the content of plutonium and neptunium in the uranium product solution is greatly reduced after only one extraction cycle of the uranium line, so that the requirements for the operation control precision and the solvent quality of the process are reduced, the probability of unqualified uranium product solution due to the excessive content of plutonium and neptunium is greatly reduced, the quality of the uranium product is improved, the method has a remarkable removal effect on the fission products such as zirconium, niobium and ruthenium in the uranium product, and can be effectively applied to the PUREX process.

The application discloses a kind of nuclear spent fuel reprocessing equipment magnetic powder sealing device and method, including shell, first pole shoe, second pole shoe, magnetic component and magnetic conducting component, two end surfaces of magnetic conducting component in the axial direction of inner cavity form sealing gap between first pole shoe and second pole shoe respectively.When injecting magnetic powder into sealing gap, make the axial direction of inner cavity parallel with the direction of gravity, and inner cavity opening is upward.Assemble, first pole shoe is assembled in the inner cavity of shell first, the pole tooth of first pole shoe is upward, magnetic powder is injected to the pole tooth of first pole shoe, then magnetic component and magnetic conducting component are assembled in inner cavity, and magnetic powder is injected above magnetic conducting component.After second pole shoe and end cover are assembled in the inner cavity of shell, magnetic powder is uniformly distributed by rotating magnetic conducting component driven by rotating shaft.In the process of assembling, magnetic component is not magnetized or weakly magnetized, finally, magnetic component is energized or magnetized, so that the uniformity of magnetic powder injection and the sealing performance of device can be improved.

The utility model relates to the field of spent fuel post-treatment corrosion electrochemistry, in particular to an electrochemical corrosion sample used in a severe environment. The sample is a cylinder with one end being a hemisphere, a hole is formed in the center of the flat end face of the sample, an internal thread I is arranged in the hole, the side, where the internal thread I is located, of the sample is fastened to one end of the connecting screw through threads, and a gasket I for enhancing sealing performance is arranged at the joint of the sample and the connecting screw. The other end of the connecting screw rod is fastened with the bolt through threads, and a gasket II for enhancing the sealing performance is arranged at the joint of the connecting screw rod and the bolt; and a glass tube is sleeved outside the connecting screw rod, one end of the glass tube is tightly pressed on the gasket I, and the other end of the glass tube is tightly pressed on the gasket II. The device can be suitable for preparing an electrochemical sample which needs to work in a severe environment, the electrochemical corrosion behavior of a material in the severe environment is researched, and the situation that the sample prepared by the existing electrochemical sample preparation method is damaged by an electrolyte solution can be effectively avoided.

The application discloses a method for determining the concentration of hydroxylamine nitrate, preparing verification set and correction set samples in the concentration range of spent fuel reprocessing process, collecting near-infrared absorption spectra of the verification set and the correction set, establishing a quantitative correction model and verifying to obtain a quantitative model, performing near-infrared spectrum detection on a solution of the spent fuel reprocessing process, not needing to pretreat the solution, and simultaneously obtaining the concentration of hydroxylamine nitrate by processing the absorption spectra of the corresponding wavelengths by the quantitative model; the method is simple, improves the analysis efficiency, and is high in accuracy; and the method provides effective data support for online detection of the concentration of hydroxylamine nitrate.

The application discloses a high-strength, high-corrosion-resistance and low-cost titanium alloy suitable for spent fuel reprocessing and a preparation method thereof, and belongs to the field of structural materials for spent fuel reprocessing. The chemical composition of the titanium alloy comprises the following components in percentage by mass: Al 4%, Zr 3%, V 0.5%-5.0%, H <0.008%, O <0.08%, N <0.04%, C <0.08%, and the balance Ti. The preparation method comprises the following steps: raw material drying and preheating, vacuum / argon protection smelting, overturning remelting, casting and hot forging. The forging breakdown temperature is 1000 DEG C, the final forging temperature is greater than or equal to 880 DEG C, the cumulative deformation is 40+ / -5%, the strain rate is 3 s ‑1 ~8 s ‑1 . The alloy has excellent corrosion rate and high yield strength in 6 mol / L HNO3 containing oxidizing ions at 100 DEG C, and is suitable for corrosion-resistant load-bearing components of spent fuel reprocessing equipment.

The application discloses an I-129 and Kr-85 online continuous monitoring system and method, relates to the technical field of spent fuel reprocessing radiation monitoring, and comprises a Kr-85 sampling detection device, an input of a first I-129 sampling detection device, and an output of a Kr-85 sampling detection device; a second I-129 sampling detection device, an output of a Kr-85 sampling detection device, and an input of a flushing interface; and a flushing interface, an output of a first I-129 sampling detection device and a second I-129 sampling detection device, and an input of an electrical control box. The application controls the on-off state of solenoid valves YV2-YV5 through the electrical control box. The application can remove Kr-85 in sampling gas through air flushing and differential measurement, and can realize I-129 and Kr-85 online continuous measurement when the concentration difference between Kr-85 and I-129 is more than 10 9 times.

The utility model belongs to the technical field of equipment heating, and particularly relates to a device suitable for pulse extraction column plate section heating. The device comprises a plurality of heating pipelines arranged in parallel; the heating pipeline is arranged around the plate section of the 2BX column and is parallel to the plate section; a preset distance is formed between the heating pipeline and the outer wall of the plate section; and hot water with a preset temperature is introduced into the heating pipeline. Heat transfer is used in a heating mode, water is used as a medium, cost is low, recycling can be achieved, damage is not prone to occurring, maintenance is avoided, radiation injury to personnel is greatly reduced, and the problem that the temperature of 2BX pulse column plate section feed liquid of a spent fuel reprocessing plant cannot meet the design requirement is solved.

The invention relates to the technical field of spent fuel aftertreatment process flow monitoring, and discloses a spent fuel dissolution degree monitoring system and method.The system comprises a radiation collection module, an automatic control module, a signal processing module and a process master control module, and the process master control module is used for collecting state signals of a loader; the signal processing module is used for controlling the automatic control module to switch the shielding thickness in the radiation acquisition module on the basis of the state signal of the loader and in combination with external preset measurement parameters, and then calculating the solubility of the spent fuel through energy spectrum analysis on the basis of the electric signals output by the radiation acquisition module under different shielding thicknesses. The invention provides a measuring system for monitoring the content of undissolved substances in the technical field of spent fuel post-treatment process equipment operation monitoring. The measuring system can specifically detect the content of the undissolved substances in spent fuel.

The application provides a spent fuel reprocessing high-temperature nitric acid full stress corrosion simulation experiment device, and relates to the technical field of spent fuel reprocessing. The device comprises an electric control system and a load loading system, a corrosion reaction system, an electrochemical test system, a temperature control system and a tail gas treatment system connected with the electric control system. The load loading system is used for loading various loads. The corrosion reaction system is composed of a glass reaction tank and a reaction tank cover. The reaction tank contains nitric acid solution for corrosion reaction of the sample. The sample penetrates the glass reaction tank and is sealed by a sealing kit. The electrochemical test system is arranged in the reaction tank cover and used for electrochemical test of the sample corrosion reaction. The temperature control system is used for adjusting and controlling the temperature of the sample corrosion reaction of the load loading system. The tail gas treatment system realizes treatment of the tail gas generated in the sample corrosion process. The device solves the problem that the stress corrosion test under high-temperature nitric acid cannot be carried out at present.

The invention provides a compound as well as a preparation method and application thereof. The compound is represented by the following formula (I) wherein R is a C1 to C14 alkyl group. The compound can efficiently extract tetravalent plutonium in a spent fuel post-treatment process, and is wide in applicable nitric acid concentration, fast in extraction kinetics and short in extraction equilibrium time.

The present application relates to the nuclear technology field, specifically to a method for separating thorium from a thorium and lanthanum mixture. The method comprises the following steps: in a three-electrode system, pulse constant current controlled potential electrolysis is performed on a molten salt system containing mixed ions of thorium and lanthanum series elements, and the deposition rate difference of thorium and lanthanum series element ions during pulse constant current electrolysis and the electrolytic reaction limitation of the control potential on the lanthanum series element ions are utilized to achieve efficient separation of thorium from the thorium and lanthanum mixture. The molten salt system containing mixed ions of thorium and lanthanum series elements is obtained by adding the thorium and lanthanum mixture into a chloride molten salt, mixing uniformly, and passing in chlorine gas mixture to chlorinate the mixture and dissolve it into the chloride molten salt. The method has the advantages of simple process, strong radiation resistance, no additional radioactive waste, and low cost, and is suitable for industrial application of molten salt reactor spent fuel reprocessing.

This invention provides a simulation method for the generation of organic iodine based on the Kolbe electrolysis reaction, belonging to the field of spent fuel reprocessing technology. This method introduces iodine into the Kolbe electrolysis reaction and compares the intrinsic ability of different forms of iodine to react with alkyl radicals to generate organic iodine. Compared with traditional simulation experiments under irradiation conditions, the method of this invention can conveniently obtain alkyl radicals through electrochemical reactions, eliminates the interference of water radiolysis products in the irradiation environment, and provides a clear understanding of the organic iodine conversion capabilities of different forms of iodine; moreover, it does not require specific irradiation equipment and shielding measures, facilitating real-time control and sampling analysis of the reaction, and has better simplicity and operability.

The invention discloses a method for solidifying radionuclide technetium in a solution and application of the method, and belongs to the technical field of spent fuel post-treatment. The curing method comprises the following steps: firstly, acidifying a technetium-containing solution by using nitric acid, then adding the acidified technetium-containing solution into a sodium metaaluminate solution, and reacting under a low-temperature condition to ensure that technetium is covalently bonded with high-activity mu < 1->-OH in a curing product in a form of pertechnetate (TcO4 <->) and is cured in pseudo-boehmite-like mixed crystals with stable crystal structures; the curing method is mild in operation condition and simple in technological process, overcomes the defects that in a traditional cement curing and glass curing method, the size of a cured product is increased, technetium is prone to volatilization and the like, has the advantages of being small in cured product size, high in curing efficiency, low in cost and environmentally friendly, the maximum curing amount can reach 100 mg / g, and the curing method is suitable for industrial production. The method is suitable for treating technetium-containing acidic nucleus waste liquid generated in the spent fuel post-treatment process.

The invention belongs to the field of structural materials for spent fuel aftertreatment, and particularly relates to a high-temperature-nitric-acid-corrosion-resistant titanium alloy for spent fuel aftertreatment and a preparation method of the high-temperature-nitric-acid-corrosion-resistant titanium alloy. The titanium alloy comprises, by mass, 1.0%-5.5% of Al, 1.5%-6.0% of Zr and the balance Ti and inevitable impurities, and the mass ratio of Al to Zr is larger than or equal to 0.8 and smaller than or equal to 1.8. The preparation method comprises the steps of vacuum melting and casting, forging and two-stage annealing treatment (in the first stage, the temperature of 750 + / -10 DEG C is kept for 50-70 minutes; and in the second stage, heat preservation is conducted for 4-6 hours at the temperature of 480 + / -20 DEG C. The prepared titanium alloy shows excellent corrosion resistance in a high-temperature nitric acid environment (for example, boiling 3 mol / L HNO3 contains multiple ions), and the average corrosion rate can be as low as 0.0023 mm / y. The alloy is remarkably reduced in cost and suitable for resisting high-temperature nitric acid corrosion in the field of spent fuel post-treatment.

The invention discloses a method and a system for on-site volume reduction of spent fuel of a high-temperature gas cooled reactor. The method comprises the following steps: heating spent fuel graphite nodules under oxygen-containing flow to remove a graphite layer to obtain TRSIO particles and generated tail gas; the TRSIO particles are discharged and then loaded into a container with a shield to achieve spent fuel volume reduction; generated tail gas is cooled and then is discharged after being subjected to filtering, molecular sieve adsorption and efficient filtering in sequence. According to the volume reduction method, the volume of the high-temperature reactor spent fuel is reduced by about 95%, the transportation cost of the high-temperature reactor spent fuel is reduced by about 50% or above, the volume reduction method can be used for removing a high-temperature reactor spent fuel post-treatment graphite layer and treating high-temperature gas cooled reactor decommissioning structure graphite, and radioactive substance minimization is achieved.

The invention discloses an acid-resistant loose nanofiltration membrane as well as a preparation method and application thereof, and belongs to the technical field of water treatment membranes. According to the preparation method, 3, 5-diaminobenzoic acid (DABA) is taken as a water-phase monomer and is subjected to interfacial polymerization with 1, 3, 5-trimesoyl chloride (TMC) on the surface of an ultrafiltration bottom membrane, and the acid-resistant loose nanofiltration membrane is obtained. Carboxyl contained in DABA can reduce the reaction activity of amino and acyl chloride groups in TMC and provide certain steric hindrance, and formation of a loose polyamide network structure is facilitated; meanwhile, the acid resistance of the nanofiltration membrane is enhanced by the contained whole aromatic rings. The prepared loose nanofiltration membrane can efficiently and selectively separate uranyl ions / coexisting inorganic ions under acidic and alkaline conditions, and shows excellent anti-pollution performance. The preparation method provided by the invention is simple in process and easy to operate, and the prepared acid-resistant loose nanofiltration membrane has potential application value for treating actual radioactive spent fuel post-treatment wastewater.

The invention relates to the technical field of spent fuel post-treatment, and discloses a carrier device for conveying a radioactive source in a pipeline, which comprises a protective body for wrapping the radioactive source, the protective body is of a spherical structure, and the carrier device further comprises a fence assembly, two moving assemblies and a driving piece, the fence assembly comprises a beam rod which is horizontally arranged and has elastic toughness and baffle discs fixedly connected to the two ends of the beam rod, the beam rod is located on the upper portion of the baffle disc, the weight of the lower portion of the baffle disc is larger than that of the upper portion of the baffle disc, and a space cavity allowing the protective body to roll and preventing the protective body from disengaging is defined between the beam rod and the baffle disc. The moving assembly comprises wheel shafts, wheel seats fixedly connected with the wheel shafts in a sleeving mode and a plurality of guide wheels evenly distributed in the circumferential direction of the wheel seats, the maximum outer circle diameter defined by the guide wheels is larger than the maximum outer circle diameter of the blocking discs, and the driving part comprises a steel cable connected with one wheel shaft. The transportation stability can be improved.

The invention relates to a fluorination-electrolysis synergistic process applied to post-treatment of a metal spent fuel, which is characterized in that fluorine gas reacts with the pretreated metal spent fuel under a high-temperature condition, uranium, plutonium and a small amount of fluorinated volatile fission products in the spent fuel are converted into gaseous fluoride to be separated and transferred; and the uranium-plutonium alloy is further prepared by using technologies such as condensation trapping and gaseous fluoride reduction. Meanwhile, fluorinated residues containing a large amount of high-radioactivity and high-heat-release fission elements and minor actinide elements are subjected to electrolytic separation through a villiaumite system, and separation and recovery of the minor actinide elements are achieved. The regeneration and reforming preparation of the metal fuel can be realized on the basis of alloy ingots obtained by a gaseous fluoride reduction process and actinide metal alloys obtained by molten salt electrolysis through injection casting supplementary materials. Compared with the prior art, efficient recovery of uranium and plutonium in the metal spent fuel and deep separation and recovery of minor actinide elements are achieved, the process is simple, and the treatment time is short.

The application provides a kind of spent fuel reprocessing process knowledge object processing method, device and equipment, wherein the method comprises: obtaining at least two knowledge objects in spent fuel reprocessing process;The knowledge object includes: at least one spent fuel reprocessing process, at least one equipment corresponding to the spent fuel reprocessing process, at least one process parameter corresponding to at least one equipment of the spent fuel reprocessing process;According to at least two knowledge objects, obtain the knowledge graph chromosome population of spent fuel reprocessing process;According to the preset algorithm, obtain the target chromosome individual vector of the knowledge graph chromosome population;According to the target chromosome individual vector, adjust the preset knowledge base of spent fuel reprocessing process, and obtain the target knowledge base.The application scheme improves the optimization ability of reasoning and problem solving of knowledge object in spent fuel reprocessing process, and improves the efficiency of spent fuel reprocessing process.

A system comprising a turbine assembly, a heat exchanger, and a compressor assembly is disclosed. The turbine assembly includes a turbine housing configured to receive an exhaust gas stream generated by a fuel converter and a turbine rotor configured to extract drive energy from the exhaust gas stream based on its expansion through the turbine assembly. The heat exchanger includes a first flow channel for receiving the exhaust gas stream from the turbine assembly, a second flow channel thermally connected to the first channel for supplying fuel to the fuel converter, and a third flow channel also thermally connected to the first flow channel. The compressor assembly includes a compressor housing for receiving the exhaust gas stream from the first flow channel and a compressor rotor capable of being driven by the drive energy.A process and an energy recovery system are also disclosed.

The invention provides a compound as well as a preparation method and application thereof. The compound is represented by the following formula (I) wherein R is a C1 to C14 alkyl group. The compound can efficiently extract tetravalent plutonium in a spent fuel post-treatment process, and is wide in applicable nitric acid concentration, fast in extraction kinetics, high in extraction efficiency and short in extraction equilibrium time.

The utility model provides an automatic separation device suitable for spent fuel post-treatment process feed liquid. The automatic separation device comprises a control system, a reagent cylinder and a treatment box, wherein the control system is used for controlling all modules to work. The device disclosed by the utility model has the advantages that the control system controls the feeding assembly to input feed liquid into the separation cylinder, a reagent in the reagent cylinder is pumped into the first main separation module through the pipeline, and the reagent and the feed liquid are mixed for extraction. Meanwhile, a related valve body is closed and opened, and a reagent is input into the other main separation module, so that extraction is also carried out in the main separation module. And repeating the steps until all the main separation modules are subjected to extraction work. After the extraction work in the main separation module is completed, the collection bottle collects the useful part of the separated material liquid, and the waste liquid bottle collects the useless part of the material liquid, so that the separation work is completed. And separation is synchronously carried out in different main separation modules according to the time required by extraction, so that the working time is greatly saved, and the separation efficiency can be effectively improved.

The invention belongs to the field of spent fuel post-treatment, and relates to a neptunium post-treatment salt-free valence adjustment method. The method comprises the following steps: 1, introducing ozone A with a certain concentration into a reactor C containing a solution B, reacting for a period of time to obtain a valence-adjusted solution R, sampling and analyzing the concentration of hexavalent Np, and calculating the oxidation rate of ozone to Np; and 2, mixing the solution R and an extraction agent D in a separating funnel according to a certain ratio, standing for layering for a period of time after short oscillation, separating to obtain upper-layer organic phase feed liquid and lower-layer water phase feed liquid, pre-treating, determining the Np concentration of the organic phase and the water phase by adopting an ICP-MS mass spectrometer, and calculating the Np extraction rate. By adopting the method, pentavalent Np can be stably and efficiently oxidized to hexavalent Np, meanwhile, the salt content of the process can be reduced, equipment and pipeline blockage is avoided, and the extraction rate of Np reaches 99% or above.

The invention relates to the technical field of spent fuel aftertreatment analysis, in particular to a method for determining hexavalent neptunium in an organic phase. A method for determining hexavalent neptunium in an organic phase comprises the following steps: preparing organic phase standard solutions of hexavalent neptunium with a series of concentrations, and determining absorption spectra of the standard solutions; performing second derivative treatment on the absorption spectrum of the standard solution to obtain a second-order guide spectrum of the standard solution; establishing a quantitative correction relation between the peak area of a characteristic absorption peak and the concentration of the hexavalent neptunium according to the second-order guide spectrum of the standard solution; measuring the absorption spectrum of the organic phase solution to be measured; performing second derivative treatment on the absorption spectrum of the to-be-detected solution to obtain a second-order guide spectrum of the to-be-detected solution; and substituting the peak area of the second-order guide spectrum of the solution to be detected at the characteristic absorption peak into the quantitative correction relation, and calculating to obtain the concentration of the hexavalent neptunium in the solution to be detected. The concentration of Np (VI) in an organic phase can be directly determined, the influence of acidity is eliminated or corrected, and the method is rapid and accurate.

The utility model belongs to the field of spent fuel post-treatment, and discloses an automatic loading and unloading type continuous operation rotary dissolver which comprises a container cavity and a rotating wheel scoop arranged in the container cavity, the rotating wheel scoop is hollow and is divided into a plurality of scoop bodies arranged in a fan shape through scoop partition plates, and the scoop bodies are arranged in the container cavity. Each scoop is provided with a charging receiving groove and a dissolving liquid circulation hole, the container cavity is provided with a pellet charging pipe matched with the charging receiving groove, the rotating wheel rotates to the pellet charging pipe and is communicated with any charging receiving groove, pellets are conveyed into the scoop through the pellet charging pipe, and a dissolving liquid in the container cavity enters the scoop through the dissolving liquid circulation hole. The core block is dissolved; aiming at the requirement of continuous production and operation of spent fuel reprocessing plants in China, the dissolver is suitable for a rotary dissolver structure for continuously dissolving spent fuel, and automatic loading and unloading of materials and separation of insoluble substances are realized.

The invention relates to the technical field of spent fuel post-treatment, in particular to a washing method of a post-treatment plant dirty solvent diluent degradation product, which comprises the following steps: (1) washing and separating a post-treatment circulating solvent by using a sodium sulfite solution; and (2) washing and separating the solution treated in the step (1) by using an organic acid solution. The washing method disclosed by the invention is designed around the purification of the degradation product of the pollution solvent diluent in the post-treatment plant, shows remarkable advantages from multiple aspects, has remarkable effects in the aspects of improving the washing effect, optimizing the technological process and the like, and has breakthrough progress compared with the prior art.