12 results about "Plutonium" patented technology

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.

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 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 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 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 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.

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.

A method for recovering plutonium from destructive analytical waste liquid of complex systems includes the following steps: S1: containing C2O4 2‑ Waste liquid containing SO4 was added to H2O2 solution. 2‑ and PO4 3‑ The waste liquid was treated with Fe(NO3)3 solution; the two solutions were combined and the acidity was adjusted to (7-8) mol / L; S2: Sodium nitrite was added to oxidize plutonium to plutonium(IV), and then plutonium was extracted from the aqueous phase at a certain ratio and acidity, followed by back-extraction at a certain ratio to obtain a plutonium solution; S3: The plutonium solution obtained in S2 was adjusted in valence and acidity, and plutonium was recovered by column separation. This invention can solve the problem of plutonium recovery from destructive analytical waste liquid containing complex anions during the nuclear fuel cycle, effectively reducing the loss of nuclear materials. The collected waste liquid was treated using a waste liquid pretreatment-extraction / back-extraction-column separation process, and the final plutonium recovery rate was greater than 90%.