57 results about "Plutonium" patented technology

The invention discloses a corrosion test method of simulated feed liquid for an evaporator of radioactive feed liquid, which comprises the following steps of: 1) for samples of at least two materials, simulating plutonium ions in the radioactive feed liquid by using the simulated feed liquid in a fast screening test and using vanadium ions in the simulated feed liquid to simulate plutonium ions in the radioactive feed liquid; selecting a first material of which the corrosion performance is suitable for an evaporator operation environment; 2) carrying out a corrosion test on the first material, using the simulated material liquid in the corrosion test, and selecting a second material; and (3) a corrosion body test is conducted on the second material, specifically, the second material is processed into a simulation evaporator, in the corrosion body test, the simulation feed liquid is used for running in the simulation evaporator to simulate running of the radioactive feed liquid in the evaporator, and a simulation evaporator product is selected. According to the method, the corrosion resistance of the plutonium-containing evaporator material is reasonably and effectively determined, the method plays a guiding role in research and operation of the plutonium-containing evaporator, and the method is good in simulation performance and high in operability.

The present invention relates to a method for separating trace plutonium from a large amount of uranium by using a solid phase leaching resin, and belongs to the field of spent fuel post-processing. A spent fuel post-processing simulation liquid containing uranium and plutonium is prepared, and the spent fuel post-processing simulation liquid is pumped into a leaching resin column at a first preset flow rate to perform a dynamic test, and the uranium and plutonium concentration in the aqueous phase at the outlet is analyzed at intervals to draw a penetration curve; an eluent is pumped into the leaching resin column after adsorption equilibrium at a second preset flow rate, and the eluent is used to elute plutonium. The method disclosed in the present invention is used to efficiently separate and purify trace plutonium in a large amount of uranium by using an extraction chromatography method having the characteristics of selectivity of solvent extraction and high efficiency of chromatographic separation, and a solid phase leaching resin with a wide range of acidity applications and high selectivity for Pu is used to achieve efficient separation of trace plutonium in a large amount of uranium, and the separation process does not require acid adjustment, and the operation is simple, the method is simple, the operation is simple, and the practicability is strong, which has a strong promotion advantage.

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.

The invention relates to the field of nuclear technologies, in particular to an oxide-based uranium-plutonium nuclear fuel for an atomic power station with thermal neutron reactors. The technical result of the invention is the development of a REMIX-type fuel composition formulation making it possible, in the nuclear fuel cycle, to involve fissile nuclear materials regenerated from spent nuclear fuel with the simultaneous use of enriched natural uranium, and also moreover providing a solution to the problems of separately using previously isolated plutonium (including weapons-grade plutonium), of reducing the quantity of spent nuclear fuel by recycling fissile materials from spent nuclear fuel during the lifespan of a nuclear reactor (up to 5 recyclings), of economizing on natural uranium for manufacturing fuel and, as a result, of making the nuclear fuel cycle self-sustainable. The invention makes it possible to use REMIX fuel for solving an entire range of problems which a closed nuclear fuel cycle may face, taking into account the number of reactors, availability of natural uranium and the amount of stored spent nuclear fuel.

A unique integrated system for recycling spent nuclear fuel (SNF) from a reactor includes a stripping tank to separate fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound. A fluorination tank is connected to the stripping tank. A fluorinating agent is injected into the fluorination tank and reacts with the uranium compound to convert the uranium compound into UF6. An electrolytic extraction tank is connected to the fluorination tank to remove plutonium and actinides by electrolytic extraction. (Fig. 8)

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,

An integrated system for recycling used nuclear fuel (UNF) emerging from a reactor has a decladding vessel separating fuel pellets from nuclear fuel rods via oxidation to produce a uranium compound. A

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 invention provides a method, device and electronic device for determining a surface adsorption model of a plutonium alloy. The method constructs a crystal structure model of plutonium based on the theoretical value of the lattice constant, replaces the plutonium atoms in the crystal structure model with pre-acquired doped atoms to construct an initial plutonium alloy structure model; optimizes the structure of the initial plutonium alloy structure model to obtain a target plutonium alloy structure model in a stable state, and then places the pre-acquired adsorption molecules in the target plutonium alloy structure model in a stable state to construct an initial plutonium alloy surface adsorption model, and optimizes the structure of the initial plutonium alloy surface adsorption model to obtain a target plutonium alloy surface adsorption model that can stably adsorb, solving the technical problem in the prior art that a surface adsorption model cannot be constructed for plutonium alloy materials and meeting the application requirements for plutonium alloy materials in actual application scenarios.

The present invention discloses a mixed K-edge absorption-X-ray fluorescence analysis method, which uses X-rays to irradiate a sample to be tested, so that the sample to be tested produces uranium K-edge absorption, and excites L-series characteristic fluorescence of uranium and plutonium in the sample to be tested, and obtains the absorption spectrum of uranium in the sample to be tested according to the X-rays after the uranium K-edge absorption of the sample to be tested, and obtains the uranium concentration according to the absorption spectrum and the X-ray blank spectrum; obtains the L-series fluorescence peak area relationship of uranium and plutonium in the sample to be tested according to the L-series characteristic fluorescence of uranium and plutonium in the sample to be tested, thereby obtaining the uranium / plutonium concentration ratio, and then obtaining the plutonium concentration of the sample to be tested. The mixed K-edge absorption-X-ray fluorescence analysis method of the present invention can break the limitation of the scope of application, effectively improve the accuracy of the analysis result, and effectively reduce the influence of X-ray scattering. The present invention also provides a mixed K-edge absorption-X-ray fluorescence analysis device.

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 invention specifically relates to a fast reactor core, belonging to the technical field of fast reactor cores, and includes a reactor body container and a core accommodated and supported within the reactor body container; the core includes a fuel assembly, a reflector assembly, a first shielding assembly, a storage well assembly, and a second shielding assembly that are arranged radially from the inside to the outside in sequence, as well as a control rod assembly arranged between the fuel assemblies; the fuel assembly is composed of depleted uranium fuel element rods and uranium-plutonium mixed fuel element rods, and the quantity ratio of the depleted uranium fuel element rods to the uranium-plutonium mixed fuel element rods is determined by the breeding ratio within the core; the breeding ratio within the core is 0.99 to 1.01. For the fast reactor core of the present invention, the reactivity increased by the newly bred fissile nuclides within the core is used to balance the burnup reactivity damage during the subsequent reactor operation, fundamentally eliminating the prompt supercritical safety risk brought about by excessive excess reactivity.

The invention belongs to the technical field of nuclear material balance analysis, and discloses an improved neutron multiplex calculation model for uranium and plutonium analysis, and the calculation model of a uranium increment factor M is a0 + a1M + a2M2 + a3M3 = 0; the calculation model of the plutonium multiplication factor M is a0 + a1M + a2M2 + a3M3 = 0; a novel relation calculation model of the plutonium multiplication factors M and alpha is a0 + a1N + a2N2 = 0; and the calculation model of the fission rate F is # imgabs0 # alpha, and the calculation model of the fission rate F is # imgabs1 #. A novel relational expression among related parameters is given, and an optimization processing algorithm is designed for the realization of the neutron multiplex calculation model, so that many problems of the existing model are eliminated, repeated calculation is avoided, and the calculation efficiency is greatly improved.

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.

To provide a nuclear reactor that is decreasing light water reactor spent nuclear fuel to persuade the media of mass communication and local residents, not to mention no need for a final disposal site, on the premise that inhabitants are decreasing and the population density is small and that it is unknown when and where what scale of earthquake takes place.SOLUTION: The breeding performance is enhanced with heavy gas cooling material and nuclear fuel low in plutonium enrichment, and the nuclear fuel is burnt to enhance the plutonium enrichment of the nuclear fuel. Simple processing of only removing fission products from the used nuclear fuel is only required. Although even photovoltaic power generation facilities exploded, dumped objects will increase in future when the power selling price decreases to 8 yen / kwh under a trade guidance (Chinese products may be used as fences in EU and as sunshades in Japan.). The nuclear reactor may take advantages thereof.SELECTED DRAWING: Figure 1

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 invention discloses an ion exchange device and a separation method of a super plutonium element, and belongs to the technical field of ion exchange. The ion exchange device provided by the invention comprises a liquid inlet, a liquid conveying pump, a first back pressure valve, a first safety valve, an ion exchange assembly and a first liquid outlet which are connected through a pipeline, the ion exchange assembly comprises at least two ion exchange columns which are directly connected with the first safety valve through pipelines; every two adjacent ion exchange columns are connected end to end through a pipeline; and a ball valve is arranged on each section of pipeline. The ion exchange device provided by the invention can effectively prolong the service life of the ion exchange resin for separating the super plutonium element, and has the advantages of flexible operation method, capability of considering the separation efficiency of the super plutonium element and the like. The invention also provides a method for separating the super plutonium element by adopting the ion exchange device.