35 results about "Plutonium" patented technology

Fuel, intended to be irradiated in a fast neutron nuclear reactor, comprising grains of at least one metal selected from U, Pu or Th, the grains being compressed in a volume, the fuel being porous, the fuel being characterized in that it comprises open intergranular pores, forming an interconnected porosity, the volume fraction of porosity preferably being between 3% and 30%, more than 50% or 80% of the porosity being an open intergranular porosity,

The invention relates to the technical field of nuclear material analysis, in particular to a plutonium solution specific activity determination method and application, and the plutonium solution specific activity determination method comprises the following steps: taking a plutonium solution with known specific activity A1 as a standard solution, determining the liquid flash counting rate of the standard solution, and recording the liquid flash counting rate as C1; obtaining a second-order conductance spectrum of the standard solution by using an ultraviolet-visible absorption spectrum method, determining the peak area of a characteristic absorption peak of tetravalent plutonium in the second-order conductance spectrum of the standard solution, and recording the peak area as S1; measuring the liquid flash counting rate of the to-be-measured solution, and recording the liquid flash counting rate as C2; obtaining a second-order conductance spectrum of the to-be-detected solution by using an ultraviolet-visible absorption spectrometry, determining the peak area of a characteristic absorption peak of tetravalent plutonium in the second-order conductance spectrum of the to-be-detected solution, and recording the peak area as S2; and calculating the specific activity A2 of the solution to be detected according to the following formula. According to the method, liquid scintillation counting and ultraviolet-visible spectrophotometry are combined, so that combination of activity response and concentration response is realized, and acquisition of isotope composition of a plutonium solution to be detected and complex sample pretreatment are not needed.

A method for determining the quantity of plutonium in a radioactive sample in the presence of curium, comprising the steps of: Measuring (301) a train of electrical pulses using a radiation detection system, Determining (302), respectively in a first and a second time window, a first and a second number of second-order pulse coincidences between the electrical pulses supplied by the detectors, The second time window being formed by the union of the first time window and a third time window, The first and third time windows being chosen such that the number of second-order pulse coincidences related to plutonium is greater than that related to curium in one of the two windows between the first and third time windows and less than that related to curium in the other window,Determine (304) the quantity of plutonium by applying a nonlinear regression model taking as parameters: the total number of measured pulses, the first number of second-order pulse coincidences, and the second number of second-order pulse coincidences, the model having been previously trained on a set of values ​​simulated from a numerical model. Figure 3,

The application discloses an off-site transportation container for plutonium dioxide powder product cups, which comprises an outer container, a shock absorber, an inner container and a holder, the inner container is arranged in the outer container, the shock absorber is arranged between the inner wall of the outer container and the outer wall of the inner container, and the holder is arranged in the inner container, the inner wall of the body of the outer container is filled with heat insulation material, the shock absorber is used for buffering the vibration of the inner container, the inner wall of the body of the inner container is filled with a shielding neutron poison, and the holder is used for supporting and limiting a plutonium dioxide powder product cup hanger. The technical scheme of the application can effectively reduce the influence of vibration and thermal shock in the transportation process on the plutonium dioxide powder product cups in the container, the shock absorption performance and the heat insulation effect are good, the product transportation requirements are met, and the safe transportation of radioactive substances is ensured.

The invention belongs to the technical field of alloy smelting and injection casting, and particularly relates to a crucible for plutonium alloy injection casting, a heat preservation system and a casting method. According to the device, a crucible is installed on a crucible base, a protrusion is arranged at the center of the bottom of the crucible, a groove is formed in the center of the upper end face of the crucible base, and the protrusion is matched with the groove; heat insulation sleeves are arranged on the outer side and below the crucible; groove holes are formed in the bottom of the crucible and the bottom of the crucible base, the lower heat insulation sleeve is installed above the top cover, and the nanometer heat insulation layer is arranged under the lower heat insulation sleeve and located between the furnace bottom cover and the top cover. The penetrating piece is located in the center of the furnace bottom cover, penetrates through the nanometer heat insulation layer and the top cover and then is embedded into the lower heat insulation sleeve, and the thermocouple penetrates through the center of the penetrating piece and is inserted into a groove hole in the bottom of the crucible. The problems that in the plutonium-containing alloy injection casting process, a crucible and a heat preservation system are complex in structure, poor in durability, low in automation degree, large in heat loss, inaccurate in temperature control, difficult to maintain and the like can be solved.

The invention provides a fast reactor spent fuel separation method, which comprises the following steps: a first extraction step: mixing fast reactor spent fuel, a first extraction agent and a first detergent, extracting radioactive elements in the fast reactor spent fuel into the first extraction agent to form a first organic phase, the radioactive elements at least comprising plutonium element and uranium element; a first back extraction step: mixing the first organic phase and a first back extraction agent, so that radioactive elements in the first organic phase enter the first back extraction agent to form a first back extraction phase; a second extraction step: reducing the acidity of the first back extraction phase, and mixing the first back extraction phase with a second extraction agent and a first reducing agent, so that uranium in the first back extraction phase enters the second extraction agent to form a second extraction phase, and a separation step: separating a plutonium product from the first back extraction phase, and separating a uranium product from the second extraction phase. According to the separation method, the radioactive elements can be separated from the fast reactor spent fuel through selective extraction and back extraction technologies, and the yield of the radioactive elements is increased.

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 invention discloses a model training method and device, an analysis method and device, related equipment, a storage medium and a computer program product. The method comprises the steps that multiple sets of sample data are obtained, an initial model is constructed, and each set of sample data in the multiple sets of sample data is related to burnup, initial uranium concentration, cooling time, curium-244 content and plutonium content corresponding to a spent fuel assembly; the initial model is trained through the multiple sets of sample data, a first model is obtained, and the first model is at least used for determining the plutonium content corresponding to the spent fuel assembly based on the burnup, the initial uranium concentration, the cooling time and the curium-244 content corresponding to the spent fuel assembly. The first model obtained through training can be used for accurately determining the plutonium content in the spent fuel assembly.

As one way of preparing for global cooling, we will provide a nuclear reactor that requires less investment capital, is based on the premise that population density will decrease due to population decline, and it is unknown when, where, and what scale an earthquake will occur, and of course does not require a final disposal site, and can convince the media and local residents by reducing the amount of spent nuclear fuel from light water reactors. [Solution] A nuclear reactor in which a large number of allowable alloy-containing nuclear fuel rod assemblies 1, each consisting of a large number of allowable alloy-containing nuclear fuel rods made of heavy noble gas coolant and allowable enrichment plutonium alloy, are arranged in a square lattice pattern in the core of an existing BWR.

At present, the world is suffering from air pollution due to thermal power generation, sea level rise due to atmospheric temperature rise, land sinking due to rising of the sea level, and public nuisance due to nuclear power generation, such as spent depleted uranium, plutonium, and.gamma.-ray radioactive materials. In order to solve these problems, inexhaustible visible-light wave power generation of sunlight radiation energy and a Huang Daozhou converter are integrated.SOLUTION: The secondary coil of the ecliptic transformer according to the present invention may be connected to a country having a shortage of electric power at night due to the rotation of the earth and a country having a shortage of electric power due to a large amount of power consumption even in the daytime via the shortest conductor on the same longitude to transmit and receive electric power.SELECTED DRAWING: Figure 1

The invention relates to a method for determining the amount of plutonium in a radioactive sample in the presence of curium, comprising the following steps: measuring (301) an electrical pulse train by means of a radiation detection system, determining (302) a first number and a second number of second order pulse coincidence between the electrical pulses delivered by the detector, respectively,-the second time window consisting of a union of the first time window and a third time window, the first time window and the third time window are selected such that the number of second order pulses associated with plutonium is higher in one of the first time window and the third time window than the number of second order pulses associated with curium and lower in the other window than the number of second order pulses associated with curium; -determining (304) the amount of plutonium by applying a non-linear regression model having as parameters: the total number of measured pulses, a first number of second order pulse coincidence and a second number of second order pulse coincidence, said model being pre-trained on a set of simulated values obtained using a numerical model.

A method for determining the quantity of plutonium in a radioactive sample in the presence of curium, comprising the steps of: Measuring (301) a train of electrical pulses using a radiation detection system, Determining (302), respectively in a first and a second time window, a first and a second number of second-order pulse coincidences between the electrical pulses supplied by the detectors, The second time window being formed by the union of the first time window and a third time window, The first and third time windows being chosen such that the number of second-order pulse coincidences related to plutonium is greater than that related to curium in one of the two windows between the first and third time windows and less than that related to curium in the other window,Determine (304) the quantity of plutonium by applying a nonlinear regression model taking as parameters: the total number of measured pulses, the first number of second-order pulse coincidences, and the second number of second-order pulse coincidences, the model having been previously trained on a set of values ​​simulated from a numerical model. Figure 3,

The application discloses a kind of measurement method and device of organic phase sample, method includes: determining uranium series working standard solution and plutonium series working standard solution;According to uranium series working standard solution and plutonium series working standard solution, measurement and respectively obtain uranium standard X-ray absorption spectrum and plutonium standard X-ray absorption spectrum;According to uranium standard X-ray absorption spectrum, calculate and obtain uranium standard natural logarithm value, according to plutonium standard X-ray absorption spectrum, calculate and obtain plutonium standard natural logarithm value;According to uranium standard natural logarithm value and the uranium concentration of uranium series working standard solution, obtain uranium working curve, according to plutonium standard natural logarithm value and the plutonium concentration of plutonium series working standard solution, obtain plutonium working curve;According to uranium working curve and plutonium working curve, measure organic phase sample.The scheme of the application does not need to prepare organic phase standard solution, carries out the measurement of organic sample, reduces standard consumption and workload, reduces the contact time with radioactive materials, reduces the dose rate level of operator.

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 present application relates to a kind of spent fuel insoluble residue processing and analysis method, comprising extracting residue, drying and weighing, strengthening dissolution, analyzing residue heavy uranium plutonium content and the like steps;Among them, through blank filter membrane contrast and eluent double baseline control, the interference influence of the filtration process in residue analysis is significantly reduced, so that nuclear material weighing result is more quantitative and reliable, conservative and reliable.Through in small sample amount, high radiation, complex multiphase matrix condition, using optimized mixed acid system proportioning and reaction time / temperature window, the full dissolution of trace target nuclide uranium / plutonium is realized, the purpose of quantitative analysis of target nuclide in high radioactivity residue is achieved, to ensure the accurate determination and safe and efficient processing of target nuclide in residue.The present application will soluble nuclide be transferred into liquid phase as far as possible, the source item intensity and volume of high-level solid residue are reduced to the greatest extent, the burden of final radioactive waste disposal is reduced, and it is suitable for insoluble residue component analysis in nuclear fuel reprocessing plant.

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.

This invention discloses a method for recovering hydrocarbons from tight rock cores. The recovery method includes the following steps: Pyrolysis of the oil-bearing rock core after crushing to obtain the relative content of total hydrocarbons S1+S2 in the core; dissolving the oil-bearing rock core with chloroform, separating the resulting crude oil to obtain saturated hydrocarbons, aromatic hydrocarbons, non-hydrocarbons, and asphaltenes, thereby obtaining the proportion of saturated hydrocarbons in the total hydrocarbons in the core; performing full-hydrocarbon gas chromatography analysis on the dissolved crude oil to quantitatively analyze n-alkanes and isoalkanes, analyze the relative peak areas of all alkane compounds in the full-hydrocarbon chromatogram, and identify plutonium and phytane; taking crude oil from the corresponding production well of the core for full-hydrocarbon gas chromatography analysis to quantitatively analyze n-alkanes and isoalkanes, analyze the relative peak areas of all alkane compounds in the full-hydrocarbon chromatogram, and identify plutonium and phytane; obtaining the total hydrocarbon content in the core as: (S1+S2) / (1-h).

The present application relates to a kind of plutonium-containing material quality analysis method, by the following steps: the calculation model of measurement object is established, the detection efficiency when loading different material amount is simulated by MCNP, the detection efficiency library of material spontaneous fission neutron under the calculation model is calculated;The detection efficiency of neutron counting module corresponding to different C value is calculated and verified, and the detection efficiency ε of corresponding neutron counting module is judged according to the C value when loading material;The total number of neutrons S of neutron counting module, the detection efficiency ε of neutron counting module is used to calculate the material quality m loaded in mixing funnel.The method provided by the present application can be used to quickly monitor the material content in mixing funnel using neutron analysis method under the condition that the process is basically stable, obtain the material content in mixing funnel, to realize real-time measurement of material quantity under special conditions, and the measurement deviation is less than 10%, meet the normal operation and safety requirement of process production, can realize normal filling and transfer storage of material.

The application discloses a bubble cap and a purification device suitable for purifying secondary steam generated by evaporation and concentration of a plutonium-containing material liquid. By reducing the size of the bubble cap and simultaneously ventilating the lower edge of the bubble cap outer cylinder, the gas flow area and the purification coefficient of the bubble cap tower can be improved. The bubble cap comprises a riser and a bubble cap outer cylinder. The riser is arranged on a tower tray body. The bubble cap outer cylinder is arranged on the riser. A plurality of tooth gaps are arranged on the side wall of the bubble cap outer cylinder. A plurality of triangular grooves are arranged on the lower edge of the bubble cap outer cylinder, so that the lower edge of the bubble cap outer cylinder is sawtooth-shaped. The outer diameter of the bubble cap outer cylinder is D, and 40mm≤D≤50mm. The inner diameter of the riser is d, and 30mm≤d≤35mm. The height of the triangular groove is H, and 8mm≤H≤12mm.

The application discloses an evaporator and method for plutonium-containing feed liquid evaporation concentration and oxalic acid destruction, comprising a heating chamber and a separation chamber, wherein: the number of the heating chamber is multiple, multiple heating chambers are symmetrically distributed around the separation chamber, the upper part of each heating chamber is communicated with the separation chamber through an upper circulation pipe, the lower part of each heating chamber is communicated with the separation chamber through a lower circulation pipe, and a communication pipe is arranged between each heating chamber; a feed liquid inlet is arranged on the lower circulation pipe, and a feed liquid outlet is arranged at the bottom of the lower circulation pipe; the separation chamber is provided with a secondary steam outlet and a leaching liquid inlet, a purification tower plate is arranged in the separation chamber, and a demister is arranged on the purification tower plate. The application can perform plutonium-containing feed liquid evaporation concentration, synchronously complete oxalic acid destruction, purify secondary steam, meet the critical safety requirement, and improve the working flux.

The present application relates to a kind of spent fuel assembly nuclear material weighing method, system and readable storage medium, the nuclear material weighing method includes: obtaining the initial nominal enrichment of fuel assembly, the type of spent fuel assembly, the burnup value of spent fuel assembly, irradiation duration, decay duration;It is input to the fuel assembly decay heat model established in advance, and obtains the nuclear material mass table exported by fuel assembly decay heat model;Receive the specific nuclide input by user, and from the nuclear material mass table, the mass of specific nuclide generated by unit mass of metal uranium nuclear reaction is inquired;The initial mass of metal uranium of fuel assembly is obtained, and the mass of specific nuclide generated by unit mass of metal uranium nuclear reaction is combined, the mass of uranium-235 in spent fuel assembly, the mass of plutonium-239, the mass of uranium, the mass of plutonium are calculated.Calculation of the mass of nuclear material at any time point from just shutdown time to spent fuel assembly external handling can be realized by the technical scheme, and human error can be reduced.

The application relates to a preparation method of a solid-phase adsorption resin and application of the solid-phase adsorption resin in uranium purification, and the preparation method comprises the following steps: (1) taking triethylamine, carbodiimide hydrochloride and 1-hydroxyphenyltriazole as catalysts, carrying out a substitution reaction of nitrilotriacetic acid and dialkylamine in dichloromethane solvent, drying a reaction product, and removing impurities through reduced pressure distillation to obtain a polyamide ligand product; and (2) loading the polyamide ligand onto a P120 carrier through a vacuum impregnation method to obtain the solid-phase adsorption resin. The solid-phase adsorption resin is suitable for deep purification of trace plutonium in a large amount of uranium, is suitable for a wide concentration range of uranium, plutonium and acid, has high purification depth, and is simple to operate, and is superior to existing related field separation materials.

The application discloses a corrosion test method for simulating a radioactive feed liquid of an evaporator, and comprises the following steps: 1) selecting a first material with corrosion performance suitable for an operating environment of the evaporator by using a simulated feed liquid in a quick screening test, wherein vanadium ions are used in the simulated feed liquid to simulate plutonium ions in the radioactive feed liquid; 2) selecting a second material by performing a corrosion test on the first material, wherein the simulated feed liquid is used in the corrosion test; 3) selecting a simulated evaporator product by performing a corrosion body test on the second material, wherein the second material is processed into the simulated evaporator, the simulated feed liquid is used in the corrosion body test to simulate operation of the radioactive feed liquid in the evaporator in the simulated evaporator. The method is reasonable and effective in determining the corrosion resistance of a material of a plutonium-containing evaporator, plays a guiding role in research and operation of the plutonium-containing evaporator, and is good in simulation and strong in operability.

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 be used for efficiently extracting tetravalent plutonium in a spent fuel post-treatment process, and is wide in applicable nitric acid concentration and high in extraction efficiency.

1. The name of the design product: soil sampler (plutonium 210). 2. The use of the design product: for soil sampling. 3. The design points of the design product: in shape. 4. The picture or photo that best shows the design points: perspective view 1.

The application belongs to the technical field of spent fuel reprocessing, and particularly relates to a plutonium reduction back extraction method. The method comprises the following steps: controlling the initial concentration of plutonium in 2AF according to the flow ratio and the concentration multiple of plutonium; controlling the liquid flow rate to be 160-200 L / h when 2AF liquid passes through an oxidation column; controlling the liquid flow rate to be 160-200 L / h when 2AF liquid passes through a degassing column; and completely removing the bottom liquid in the 2AF feeding tank every two or three batches, and completely removing the waste coal oil in the oil removal tank by one-time passing through a pulse extraction column. The method can reduce the content of nitrous acid in the plutonium second-cycle organic phase, avoid the accumulation of nitrous acid in the 2AF liquid and the organic phase, reduce the thermal effect of plutonium accumulation, and is suitable for long-term continuous operation of the system. Through repeated experiments, the system can be stably operated, and the column temperature and the hydrodynamic parameters of the 2BX column are stable.

The present application relates to a kind of from REMIX spent fuel without separation recovery uranium, plutonium, belong to spent fuel reprocessing technical field, the method comprises the following steps: S1, the REMIX spent fuel assembly structure of to be handled is disassembled and sheared, and REMIX spent fuel pellet is obtained;S2, the spent fuel pellet is dissolved, and feed liquid is obtained;S3, the element of splitting in the feed liquid is decontaminated using co-extraction, co-reverse extraction circulation process, so that splitting purification meets product requirement;S4, after the feed liquid treated in step S3 is treated using tail end co-precipitation calcination process, and uranium-plutonium co-precipitation is obtained, and uranium-plutonium mixed oxide is obtained.The method provided by the present application can meet the requirement of 99.9% of uranium-plutonium yield under the condition of uranium-plutonium not separation.

The invention provides a method for separating radioactive elements in a fast reactor spent fuel, which comprises the following steps: a mixing step: adding a first extractant into a mixture of the fast reactor spent fuel and a pressurized water reactor spent fuel to obtain a first mixture, the concentration of plutonium element in the fast reactor spent fuel is greater than that of plutonium element in the pressurized water reactor spent fuel, and the concentration of the pressurized water reactor spent fuel is greater than that of the first mixture; the fast reactor spent fuel and the pressurized water reactor spent fuel both contain uranium element, and the concentration of plutonium element in the first mixture is less than 30 g / L; the first mixture, a second extraction agent and a reducing agent are mixed to obtain a second mixture, the second mixture comprises an organic extraction phase and an inorganic extraction phase, and the reducing agent is used for reducing the chemical valence of the plutonium element from + 4 valence to + 3 valence; and a separation step: separating a plutonium element product from the inorganic extraction phase and separating a uranium element product from the organic extraction phase, and the separation method can realize separation of radioactive elements from the fast reactor spent fuel.

This invention discloses a plutonium-containing material measuring device, relating to the technical field of radioactive element measuring devices. The device includes a measuring chamber, a lower cover assembly, and an automatic feeding structure. A gamma detector is installed above the measuring chamber, and a neutron detector is installed inside. The automatic feeding structure includes a horizontal transmission mechanism and a synchronous lifting mechanism. The synchronous lifting mechanism is a symmetrical structure driven by a single power source. The lower cover assembly, which carries the material to be measured, is placed on the horizontal transmission mechanism and can be conveyed horizontally to the lifting position. The synchronous lifting mechanism then conveys the lower cover assembly along with the material to be measured vertically into the upper measuring chamber, and the lower cover assembly seals the bottom of the measuring chamber, thus forming a closed measuring space. The device provided by this invention can simultaneously perform neutron and gamma measurements on materials, reducing measurement time, minimizing transport time, reducing radiation hazards to personnel, reducing equipment size, and increasing measurement accuracy.

This invention discloses an apparatus for oxidizing Pu(III) in a liquid feed to Pu(IV), comprising: an oxidant supply unit; a liquid feed unit; an oxidation unit connected to both the oxidant supply unit and the liquid feed unit, including a continuous oxidation reactor and a storage tank, wherein the continuous oxidation reactor is located within the storage tank and the material overflows from the continuous oxidation reactor into the storage tank; and a tail gas absorption unit connected to the storage tank. The liquid feed unit, oxidation unit, and tail gas absorption unit are located in a glove box or a heated chamber. This invention achieves long-term, continuous, stable, and safe operation of the oxidation of Pu(III)-containing liquid feed, as well as miniaturization, less plutonium mass loss, effective control of the critical mass of plutonium, and control of heat generation per unit time within the equipment, effective gas-liquid separation, and easy discharge of the generated gas.