184 results about "Fuel reprocessing" patented technology

The invention discloses a preparation method of a porous-fuel-core inert matrix dispersion fuel pellet. The preparation method mainly comprises the steps of preparation of porous fuel core beads, preparation of inert matrix slurry, preparation of composite particles, pressing of a core, pressing of a core shell and preparation forming. According to the method, TRISO particles in an existing pellet is replaced with a porous fuel core, so that uraniumu abundance in the pellet is improved by four times or more, neutron economy of the fuel pellet can also be obviously improved, and power station operation cost and spent fuel reprocessing cost are reduced; moreover, a process period can also be greatly shortened, batch preparation of the pellet is implemented, and production cost is obviously reduced; the prepared inert matrix dispersion fuel pellet is higher in compactness, smaller in crystal grain size, fewer in structural defects and more excellent in high-temperature stability and irradiation resistance.

The invention discloses application of a two-dimensional transitional metal carbide nanosheet as a radionuclide adsorbent. Due to the fact that two-dimensional transitional metal carbide nanosheet MXene material has an absorption effect on radionuclide, the two-dimensional transitional metal carbide nanosheet can be utilized as the adsorbent for absorbing radionuclide.Compared with other inorganic nanometer adsorption material, MXene material, featuring a nano layered structure, can effectively increase adsorption capacity of radionuclide. In the meantime, storage space of the adsorbent is greatly reduced due to large density of MXene material.Accordingly, MXene material has potential application value in the field of irradiated fuel reprocessing, control over radioactive pollutants discharged by a nuclear power station, a hospital, a factory and others.

The invention belongs to the technical field of fuel-short after-treatment and discloses a plutonium purification and concentration method. The method comprises the following steps: extracting tetravalent plutonium in a nitric acid solution into a 30% TBP (tributyl phosphate)-kerosene solution with relatively small volume for purification and concentration in a centrifugal extractor, reducing tetravalent plutonium into trivalent plutonium by using a dimethylhydroxylamine-containing nitric acid solution with relatively small volume, performing back extraction on trivalent plutonium, and further purifying and concentrating plutonium in a water phase. In the whole concentration process, a plutonium replenishment and extraction link is removed. The method is simple in step, short in time and small in solvent radiation effect; the plutonium back-extraction yield reaches up to 99.98% while the plutonium concentration is increased by more than 10 times.

The invention discloses a method for synthesizing diamide podand extraction agent, wherein chloro-carbonic ester and diglycolic anhydride react to generate mixed anhydride under the action of tertiary amine, and then the reaction is performed with amine to generate the diamide podand extraction agent. The method has mild reaction condition, can be performed at a low temperature, and has high reaction speed and short consumed time; the product purification operation is simple and easy; the obtained diamide podand extraction agent can meet the extraction purity requirement, and is beneficial to establishing flow that the diamide podand is used for treating high-level liquid waste in spent fuel reprocessing plant; the used chloro-carbonic ester is easy to prepare and has a low price, therefore, the cost for preparation of a great amount of extraction agent is greatly reduced; and besides, the yield of the extraction agent is high, so that the method is very suitable for industrial production and application.

The invention relates to the technical field of nuclear fuel post-treatment, and discloses a bismuth-based functional material capable of adsorbing gaseous iodine as well as a preparation method and application thereof. The method includes dissolving a bismuth salt and polyacrylonitrile into a solvent and mixing to form a precursor solution; carrying out electrostatic spinning to obtain a fiber membrane; and then pre-oxidizing in an air atmosphere and carbonizing in an inert gas atmosphere to obtain the bismuth-based functional material. According to the material, a carbon nanofiber membrane is used as a carrier, and metal bismuth nanoparticles are uniformly attached to fibers, so that rich active sites are provided for chemical adsorption of iodine, the adsorption capacity can reach 560 mg/g, and gaseous iodine can be effectively adsorbed and separated. Meanwhile, the material is simple in preparation method and low in raw material cost, and more importantly, compared with most of powdery adsorbents, the material has a macroscopic membrane form structure, is good in flexibility and high in thermal stability, is expected to be applied to large-scale industrial application in the spent fuel aftertreatment process and has a wide prospect.

The invention belongs to the technical field of nuclear waste treatment and recycling, and relates to a method for extracting and recycling neptunium from spent fuel aftertreatment waste liquid. The method comprises following optional order-replaceable repeatable steps, and water phase back-extraction matte is finally collected. The steps comprise 1, neptunium in a neptunium-containing water phase solution is oxidized to be hexavalent, and an organic solvent containing dimethyl phosphonate (1-methyl heptyl) ester is added under the strong acid condition for extraction; organic phases are collected, a reducing agent is added, reextraction is carried out under the dilute acid condition, and therefore hexavalent neptunium in the organic phases is selectively reduced to be pentavalent and reextracted to enter a water phase; 2, a reducing agent is used for reducing neptunium in the water phase solution to be tetravalent, and an organic solvent containing dimethyl phosphonate (1-methyl heptyl) ester is added under the strong acid condition for extraction; and organic phases are collected, reextraction is carried out under the dilute acid condition, and therefore tetravalent neptunium in the organic phases is reextracted to enter the water phase. By means of the method, a neptunium nitrate solution with standard purity can be prepared.

The invention relates to the critical safety control technologies of nuclear fuel reprocessing plant dissolvers, in particular to a critical safety control method for a dissolver with an annular solid neutron poison partition layout. The dissolver includes a big hanging basket disposed in a dissolver intramural dissolution solution. The big hanging basket is internally provided with a solid neutron poison partition structure, which comprises an annular neutron poison layer. The inner side and the outer side of the neutron poison layer are respectively provided with a poison outer coating shell and a poison inner coating shell, and the poison inner coating shell is internally provided with a neutron moderation material. The method provided by the invention improves the cross-sectional area and loading capacity of the reprocessing plant key technological equipment dissolver, and can solve the critical safety control problem of reprocessing plants.

The invention belongs to the technical field of spent fuel reprocessing, and discloses an analysis method for a trace oxalic acid root in a mother liquor of plutonium oxalate precipitation. The analysis method comprises the following steps that: (1) the mother liquor of plutonium oxalate precipitation is diluted by 10 to 20 times, and a reducing agent is added to destroy potassium permanganate in the mother liquor of plutonium oxalate precipitation; (2) the test solution prepared in the step (1) is placed on an IC - H column to remove the positive ions in the solution; (3) the solution collected in the step (2) is placed into a container and heated under a temperature condition of less than 60 DEG C, and a gas is used to purge the upper surface of the solution to full dry at the same time; (4) a beaker is washed with deionized water for many times, the washing liquor is collected for fixing a volume, and the content of the oxalic acid root in the solution is measured by an ion chromatography. The analysis method has the characteristics of small interference of matrix, high accuracy and simple device adopted.

The invention discloses a device and method for treating radioactive organic waste liquid by using supercritical water oxidation. The device for treating the radioactive organic waste liquid by using the supercritical water oxidation is integrated with a high pressure liquid pump, a high pressure gas pump, a supercritical water oxidation kettle, a heating system (tube furnace), a one-way valve, a back-pressure valve and a radioactive operation glove box to form a supercritical water oxidation device; by using a continuous operation method, the organic waste liquid is oxidized to rapidly and efficiently transform the radioactive organic waste liquid into radioactive waste water and carbon dioxide. At present, the radioactive organic waste liquids to be treated in time generally come from nuclear related units; the typical of the radioactive organic waste liquids to be treated in time is waste oil generated by nuclear power plant workshops, tributyl phosphate (TBP) + diluting agent in the reprocessing process of spent nuclear fuel, scintillation liquids generated in the nuclear medicine process and so on. According to the device and method for treating the radioactive organic waste liquid by using the supercritical water oxidation, by using the supercritical water oxidation technique, the organic waste liquid containing radioactive nuclide is completely oxidized, so that the organic waste liquid containing radioactive nuclide is completely transformed into carbon dioxide, water and inorganic salts; therefore, the organic waste liquid containing radioactive nuclide becomes common radioactive waste water to be convenient to treat.

The invention relates to a nuclear power technology, in particular to a multi-reactor combined nuclear fuel cycle utilization method which is applied to the management design of first core fuel for a nuclear plant. A plurality of reactors with fuel elements which can be interchanged are combined to operate the cycle utilization of nuclear fuel, parts of fuel elements are used in a plurality of reactors in sequence. The invention has the advantages that the utilization ratio of the fuel for the newly-built the nuclear power station at the earlier stage can be improved to the level during the equilibrium cycle at the later stage, and the fuel purchasing cost and the post treatment cost of the fuel are saved; the long cycle refueling of a refueling shutdown reactor type can be realized in the first core. The method also can be used for the optimization of the management design of the reactor core fuel among operating reactors, thereby improving the fuel utilization ratio. The method of the invention is suitable for both the refueling shutdown reactor type and a continuous refueling reactor type and suitable for a pressurized water reactor, a boiling water reactor, a heavy water reactor and a high temperature gas cooled reactor.

The invention discloses a power generation and nuclear fuel processing disposal integrated nuclear energy base based on closed cycle. The power generation and nuclear fuel processing disposal integrated nuclear energy base comprises a nuclear energy power generation area, a spent fuel and waste disposal area, and a waste final disposal area. The nuclear energy power generation area comprises a reactor workshop, a nuclear auxiliary workshop, and a fuel workshop. The spent fuel and waste disposal area comprises a spent fuel and waste disposal workshop and a waste processing workshop. The waste disposal area comprises a commissioned near surface disposal warehouse, a decommissioned near surface disposal warehouse, a middle depth disposal warehouse, and a deep geological disposal warehouse. The nuclear energy base is divided into the power generation area, the fuel and waste disposal area, and the waste final disposal area according to functions, and the planning, the siting, and the investment of the whole nuclear energy base are realized in a unified manner, and therefore planning is facilitated, and factory site resources are saved. The spent fuel and the wastes are processed and disposed in the neighborhood, and therefore radiation risks caused by long distance transport are prevented.

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 belongs to the technical field of nuclear material extraction and recovery, and relates to a method for extracting and recovering plutonium from a plutonium-containing nitric acid solution. The method comprises the following steps: oxidizing plutonium in the spent fuel reprocessing plutonium-containing nitric acid solution to tetravalent with an oxidant, adding an organic solvent containing DMHMP, Di(1-methyl heptyl)methyl phosphonate for extraction under strong acidic conditions, collecting organic phase, adding a reducing agent, performing back extraction under dilute acidic conditions to selectively reduce the tetravalent plutonium in the organic phase to trivalent, and performing back extraction to form the aqueous phase. The method for extracting and recovering plutoniumfrom the plutonium-containing nitric acid solution can efficiently separate plutonium from other impurities, such as uranium, fission product elements of plutonium, strontium, and cesium, and concentrate plutonium to obtain the plutonium nitrate product solution with standard purity.

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.

A spent fuel reprocessing method has a dissolution step of dissolving the spent fuel in nitric acid solution, an electrolysis/valence adjustment step of reducing Pu to trivalent, maintaining the pentavalent of Np, a uranium extraction step of collecting UO₂ by bringing the fuel into contact with organic solvent and extracting hexavalent U by means of an extraction agent, an oxalic acid precipitation step of causing MA and the fissure products remaining in the nitric acid solution to precipitate together as oxalic acid precipitate, a chlorination step of converting the oxalic acid precipitate into chlorides by adding hydrochloric acid to the oxalic acid precipitate, a dehydration step of synthetically producing anhydrous chlorides by dehydrating the chlorides in a flow of Ar gas, and a molten salt electrolysis step of dissolving the anhydrous chlorides into molten salt and collecting U, Pu and MA at the cathode by electrolysis.

The invention discloses a method for selective extraction separation of trivalent actinides and lanthanides in fuel-short after-treatment. The method is implemented through the steps that NTAamide(n-Oct) serves as an extraction agent, kerosene serves as a diluting agent, a water-soluble ligand TEE-BisDGA serves as a detergent, a hydroxycarboxylic complexing agent or ammoniacarboxylic complexing agent or a nitric acid solution serves as a stripping agent, and a multi-stage extraction technology is utilized, firstly, a large number of actinide elements and part of lanthanide elements are extracted from feed liquid through the extraction agent, then the lanthanide elements in a loaded organic phase are removed through the detergent, finally the actinide elements are stripped from the organicphase through the stripping agent, thus selective extraction separation of the actinide elements and the lanthanide elements can be realized, the recovery rate of the actinide elements exceeds 99%, and the recovery purity can reach 99.9% or above.

The invention belongs to the technical field of irradiated fuel reprocessing and discloses a uranium and plutonium separation technology in a Purex process. According to the technology, in a 1B tank, a reducing agent 1BX and a fill extraction agent 1BS are used for reducing Pu (IV) in 1BF feed liquid into Pu (III), so that plutonium enters a water phase, and uranium is reserved in an organic phase; therefore, the uranium and the plutonium are separated. The used reducing agent 1BX comprises 0.1-0.2mol/L of N-hydroxyl-1-hydrazine carbomite, 0.35-0.5mol/L of NO<3-> and 0.35-0.5mol/L of H<+>; the fill extraction agent 1BS is 30% TBP- kerosene containing 0.6-1mol/LHNO3. The technology has the characteristics of high plutonium yield, good uranium and plutonium separation effect and high technology safety factor without needing adding a holding reducing agent.

The invention belongs to the technical field of analytical chemistry, and discloses a system for concurrently measuring a uranium concentration and a nitric acid concentration during a spent fuel post-treatment process. The system comprises two parts such as a hardware system and a software system, wherein the hardware system comprises a sampling analysis system, an instrumentation system, a data acquisition and control system, an industrial control machine and a peripheral equipment system, the sampling analysis system comprises a U-shaped oscillation tube density meter and a electrode-free conductivity analyzer, analog signals of the measured uranium concentration and the measured nitric acid concentration are read by the instrumentation system, are converted into digital signals through the data acquisition and control system, and are transmitted to the industrial control machine through a data transmission line, and the uranium concentration and the nitric acid concentration in the solution are directly calculated through a UraniumAcid1.0 software. With the system, the uranium concentration and the nitric acid concentration can be rapidly and concurrently measured.

The invention belongs to the technical field of analytical chemistry, and relates to a method for analyzing trace plutonium content in a large amount of uranium in a nuclear fuel aftertreatment process. The analysis method sequentially comprises the following steps: (1) taking a sample with a certain volume, and adding concentrated nitric acid to adjust the concentration of nitric acid to 1-2 mol/L; (2) adding a sodium nitrite solution into the sample, uniformly mixing and standing for a period of time; (3) adding an extraction agent into the sample for extraction, performing centrifugal separation, removing a water phase, and retaining an organic phase; (4) washing the organic phase with a washing solution for 2-5 times, removing the water phase after each washing, and retaining the organic phase; and (5) diluting the organic phase with isopropanol, and analyzing the diluted organic phase with ICP-MS provided with an organic sample injection system. By utilizing the method for analyzing the trace plutonium content in the large amount of uranium in the nuclear fuel post-treatment process, the trace plutonium content in the large amount of uranium in the nuclear fuel post-treatmentprocess can be simply, conveniently, quickly and accurately analyzed with less required sample amount and little interference.

The invention discloses a method for directly dissolving uranium dioxide or spent fuel oxides with ionic liquid. UO2 or spent fuel oxides are dissolved in iron (III)-containing ionic liquid directly, and no additional strong oxidants are needed to be added into the ionic liquid system. The method has advantages of high dissolution efficiency, good economical efficiency, green, environmental protection and the like. The method can be used for dissolution process of spent fuel reprocessing and also can be used for preparation and purification of uranium mine concentrates.

The invention discloses quick-replaceable filter. The quick-replaceable filter is composed of a filter core and an outer sleeve, and the filter core is arranged in the outer sleeve and is connected with the outer sleeve through threads; the filter core is composed of an upper end, a filter net and a lower end, the upper end of the filter net is connected with the upper end, the lower end of the filter net is connected with the lower end, and a liquid inlet and a liquid outlet are formed in the lower end of the outer sleeve. The provided quick-replaceable filter is mainly applied to spent fuel aftertreatment and other environments needing remote operation of a mechanical hand and has the advantages of being easy to replace and preventing liquid leakage.

The invention belongs to the field of nuclear reactor engineering, in particular to a follower component driven by a driving mechanism in a pressure shell and a water reactor adopting the same. The invention realizes the spectral shift control of a reactor and compansates the decrease of reactivity in a burn-up process by a depleted uranium (or thorium) compound follower, and adopts a dense weir and residential uranium compound fuel postprocessed by partially utilizing natural uranium and spent fuel, thereby greatly saving nuclear fuel and decreasing fuel cycle cost. A driving mechanisms driving a great amount of follower components are arranged in the pressure shell, so as to prevent the joint sections of the driving mechanism from penetrating through the upper sealing head of the pressure shell, carry out the spectral shift control under the condition without influencing the reliability of the pressure shell, lower the boron concentration of the water reactor in water in an operation process, avoid the occurrence of the problem of serious corrosion of the joint sections and the upper sealing head of the driving mechanism due to the penetration of boric acid through the upper sealing head of the pressure shell, reduce the possibility of elastic rod accidents and improve the safety.

The invention relates to a long-distance-disassembling-assembling sword-type mechanical arm. The long-distance-disassembling-assembling sword-type mechanical arm is used for a sampling cabinet of a power-reactor fuel-short-after-treatment middle test plant and sampling operation of a plutonium tail-end discharging working box. The inside of a clamp is connected with a nut rod through a rivet, the outside of the clamp is connected with a first sleeve through a countersunk head screw, and the outside of the countersunk head screw is covered with a polytetrafluoroethylene ring; one end of a guiding rod is provided with a screw rod and connected with a nut rod, and the other end of the guiding rod is connected with a core shaft through a pin; the first sleeve and a second sleeve rotate and are pressed through a nut and are connected through a cross recessed pan head screw; a compression spring is arranged at the tail portion of the core shaft and fixed through a spring base, and the spring base is connected with the second sleeve through a thread; a locking screw is arranged on the second sleeve; a trigger is connected with a connecting rod through a rivet, the connecting rod is connected with a rocking rod through a rivet, one end of the rocking rod is fixed at the lower end of a handle, and the other end of the rocking rod goes deep into a square trough at the tail of the core shaft. By means of the long-distance-disassembling-assembling sword-type mechanical arm, the problems that a sword-type mechanical arm trigger falls off, operation is heavy and arduous, the clamp faults, and the spring is lapsed and difficultly replaced are solved.

The invention relates to the technical field of spent fuel post-treatment, and discloses a gaseous iodine adsorption material taking foamed nickel as a framework and a preparation method and application thereof. The preparation method comprises the following steps: putting foamed nickel into a solution containing bismuth salt or silver salt, carrying out solvothermal reaction, taking out the foamed nickel, and washing and drying to obtain the gaseous iodine adsorption material. The foamed nickel is used as a framework, the metal bismuth or silver wraps the surface of nickel through solvothermal reaction, abundant active sites are provided for adsorption of gaseous iodine, and the adsorption capacity of the adsorption material is greatly improved. And the adopted foamed nickel does not contain a microporous structure, so that the gaseous iodine cannot be physically adsorbed, the prepared adsorption material captures the gaseous iodine by virtue of the chemical adsorption effect of bismuth or silver, the risk of iodine migration is not easy to occur in the later period, and the defects of a porous adsorption material in the prior art are overcome.