96 results about "Uranium hexafluoride" patented technology

The invention relates to a process for preparing a castable powder of uranium dioxide UO2, for use in the manufacture of MOX fuel.This process comprises the following stages:1) to prepare an aqueous suspension of a powder of UO2 obtained by dry process from uranium hexafluoride, said suspension comprising 50 to 80% by weight of UO2 and at least one additive chosen among deflocculation agents, organic binders, hydrogen peroxide H2O2 and a powder of U3O8, in such a quantity that the viscosity of the suspension does not exceed 250 mPa.sec, and2) to atomise this suspension and dry it in a hot gas, at a temperature of 150 to 300° C., to obtain a castable powder of UO2 with an average particle size of 20 to 100 mum.

A method of declogging at least one filter of a plant for manufacturing uranium oxide from uranium hexafluoride, including separating, from the wall of the filter, uranium oxyfluoride particles deposited, by a stream of inert gas such as nitrogen, injected into the filter, in a counter-currentwise direction to the flow of hydrofluoric acid.

The invention discloses a recovering method for uranium and fluorine in uranium hexafluoride alkali absorption solution waste liquid. According to the recovering method, the uranium hexafluoride alkali absorption solution waste liquid is firstly acidized through acid, the pH is equal to 2.7-5.5 through adjustment, then macromolecule adsorbent (polystyrene-cyclohexyl amino maleic acid) is added to finish UO22+ adsorption, solid-liquid separation is performed, so that desorption is performed on obtained solids through acid with concentration larger than 1M to recover UO22+, Na2CO3 is added into the remaining solution to perform a next cycle of uranium hexafluoride elution, and after five or more cycles, the solution is rich in F- high in concentration. NaC1 is added into the solution, so that the solution is saturated, and NaF is precipitated due to the fact that the solubility is reduced, NaF solids are obtained through filtering and recovering, CaC12, or CaO or Ca (OH)2 solids or saturated solutions are then added into the solution, so that the residual F- forms CaF2 sediment, and fluoride removal is performed deeply. Finally, the uranium and the large amount of fluoride in the uranium hexafluoride alkali absorption solution waste liquid are recovered. Uranium in the radioactive waste liquid is recovered, zero drainage of the waste liquid is achieved, the drainage of the waste liquid reaches the national standard, and meanwhile the problem that in a traditional fluoride removal technology, a large amount of calcium fluoride is difficult to filter is solved.

The invention belongs to the field of uranium conversion, and particularly relates to a membrane purification treatment process for uranium-containing process waste liquid generated in a production process of uranic fluoride. The membrane purification treatment process comprises the following steps: (1) establishing a uranium-containing waste liquid treatment purification system; (2) carrying out pretreatment; (3) carrying out ultrafiltration; (4) carrying out nanofiltration; and (5) carrying out reverse osmosis. By establishment of the membrane purification treatment system for the uranium-containing waste liquid, after the uranium-containing waste liquid of which the uranium concentration is about 200-800 mg/L in a uranium purification and conversion system is separated by a nanofiltration unit (under the operation pressure of 0.3 MPa) and a reverse osmosis unit (under the operation pressure of 1.0 MPa), the concentration of uranium in permeate can be reduced to be 0.05 mg/L or below, and therefore the permeate reaches the emission standard. In addition, the membrane treatment technology has the advantages of low energy consumption, simple system, less secondary wastes and the like.

An overpack for receiving a steel cylinder, such as a stainless steel cylinder, containing uranium hexafluoride includes a semi-cylindrical top portion having an arcuate main body and opposing first and second semi-circular end members, and includes a semi-cylindrical bottom portion having an arcuate main body and opposing first and second semi-circular end members, with the first end members of the top and bottom portion being aligned, and the second end members of the top and bottom portion being aligned, and the overpack is disposed in a cradle.

A process for manufacturing nuclear fuel pellets through sintering of a material containing uranium dioxide obtained from a powder from a power originating from a process for a conversion of uranium hexafluoride.

A system and method are disclosed for determining the enrichment of ²³⁵U in Uranium Hexafluoride (UF6) utilizing synthesized X-rays which are directed at a container test zone containing a sample of UF6. A detector placed behind the container test zone then detects and counts the X-rays which pass through the container and the UF6. In order to determine the portion of the attenuation due to the UF6 gas alone, this count rate may then be compared to a calibration count rate of X-rays passing through a calibration test zone which contains a vacuum, the test zone having experienced substantially similar environmental conditions as the actual test zone. Alternatively, X-rays of two differing energy levels may be alternately directed at the container, where either the container or the UF6 has a high sensitivity to the difference in the energy levels, and the other having a low sensitivity.

The invention provides a method and equipment for determining the background of a uranium hexafluoride gas uranium abundance measuring device. The method includes the flow step, the sealing step, the measurement step and the calculation step. In the flow step, UF6 gas of the same kind is made to flow into or flow out of a measuring container. In the sealing step, the measuring container is sealed, and then the UF6 gas in the measuring container is made to stop flowing. In the measurement step, after the pressure of the UF6 gas in the measuring container is steady, the gamma ray section of interest gross-count rate value, the pressure value and the temperature value of the UF6 gas are measured. In the calculation step, under the condition that the flow step, the sealing step and the measurement step are repeatedly executed for n times to acquire n gamma ray section of interest gross-count rate values, n pressure values and n temperature values, and the gamma ray section of interest background counting rate value of the UF6 gas uranium abundance measuring device is calculated by the utilization of the measured n gamma ray section of interest gross-count rate values, the measured n pressure values and the measured n temperature values, wherein n is an integer which is larger than or equal to 2.

The invention belongs to the field of uranium conversion, and particularly relates to a purification treatment system of fluorine-containing and uranium-containing process tail gas generated in the uranium hexafluoride production process. According to the provided fluorinated process tail gas purification treatment system, the purification treatment on harmful ingredients in the fluorinated tail gas generated in the uranium hexafluoride production process is realized by a charcoal adsorption-alkali liquor elution technology determined through the groping for a long time; the purified process tail gas reaches the emission standard. The technical method has the advantages that the system structure is simple; the operation is simple and convenient; the control conditions are easily realized; the security coefficient is high, and the like.

The invention relates to a method for simultaneously determining Mo element, Ti element and W element in uranic fluoride, belonging to the chemical analysis field. The invention is characterized in that the method for simultaneously determining the Mo element, the Ti element and the W element through a neutral network spectrophotometry is as follows: the dosage of 4 to 8 mol/L hydrochloric acids is between 0.5 and 1.0 mL; the dosage of (1-3)x10<-3>mol/L salicyl fluorone is between 2.5 and 3.5 mL; the dosage of (1-3) x10<-2>mol/L cetrimide is between 3.0 and 6.0 mL; and the color development time is between 10 and 60 minutes. The method combines a neutral network and a spectrophotometry, gives full play to the advantages of the spectrophotometry and simplifies the chemical treatment process when a plurality of compositions are determined simultaneously. The method not only can provide an accurate chemometrics method for simultaneously determining a plurality of impurity elements in the uranic fluoride through the spectrophotometry but also can be promoted to the other uranium compound analysis field. The content ranges of the Mo element, the Ti element and the W element in the uranic fluoride determined by the method are respectively 0.5 to 7.0mu g/gU,0.3 to 5.0mu g/gU and 0.5 to 7.0mu g/gU.

The invention relates to nuclear material analysis by radiation methods and can be used for on-line monitoring uranium hexafluoride enrichment in isotopic dividing production. The uranium-235 gamma-radiation intensity, the temperature and the pressure of uranium hexafluoride are determined by means of averaged measurement results within identical time intervals when a current time is changed at a value equal to the time segment of the time interval. The calculated value of the uranium-235 mass fraction is referred to the current time in coincidence with the measuring time interval. The invention also discloses a control system of the method, wherein at least three descriminators and a timer are mounted in a controller, the output end of the descriminator is connected with the input end of a gamma emission detector, each output end of the descriminators is connected with the input end of single electrical impulse counter.

The invention discloses a measuring container of an online detecting device for UF6 gas uranium abundance, wherein the measuring container is of a cylindrical structure, and comprises a shell, an inlet pipe, an outlet pipe and a guide plate; the guide plate divides the container into a continuous flowing channel in the shell; and the inlet pipe is located at an inlet end, and the outlet pipe is located at an outlet end. The measuring container provided by the invention is capable of online and real-time detecting gas uranium abundance, and has the characteristics of simple structure, high probing efficiency and no gas retention.

A device and method for measuring the relative amounts of ²³⁵U and ²³⁸U in the gaseous phase of UF₆ using tunable diode laser spectroscopy (TDLS). The invention is a faster, more accurate method for making such measurement. The improved method involves using methane (CH₄) or acetylene (C₂H₂) as a calibration gas, and using a Fabry-Perot interferometer to calibrate the laser frequency, to determine the relative amounts of ²³⁵U and ²³⁸U in UF₆ gas within ±0.27% accuracy, for example 7.00%±0.27%. The sample time required for such measurement is about one second.

The invention provides a device for converting solid uranium hexafluoride into gas uranium hexafluoride. A receiving and supplying container comprises a hollow container body and a uranium hexafluoride tank arranged in the hollow container, wherein the uranium hexafluoride tank is connected with a receiving channel and a supplying channel separately; a cooling device is used for reducing a temperature in the uranium hexafluoride tank to a first preset temperature, and then a control device is used for making the receiving channel communicated to receive a supplied solid uranium hexafluoride material; a heating device is used for elevating the temperature in the uranium hexafluoride tank to a second preset temperature to make the solid uranium hexafluoride in the uranium hexafluoride tank gasify, and the control device is used for making the supplying channel communicated to supply the gas uranium hexafluoride outwards. According to the device for converting the solid uranium hexafluoride into the gas uranium hexafluoride, through cooperation of the cooling device, the heating device, the uranium hexafluoride tank and the control device, a process of receiving the material by the 1-L uranium hexafluoride container in a cooled state and then supplying the gas uranium hexafluoride outwards is completed, operation of shortening a complex tedious technology line and completing the technology line in one piece of equipment is achieved, and the operability is good.

The invention belonging to the technical field of nuclear fuel circulation, and particularly relates to a method for preparing triuranium octaoxide from uranium hexafluoride. The method comprises the following three steps: step 1, preparation of a UO2F2 hydrolysate; step 2, preparation of UO2F2 powder; step 3, roasting for preparation of U3O8. According to the method, the process of preparing U3O8 through conventional UF6 reduction is shortened, and the addition of ammonia water in the process is avoided, so ammonia nitrogen wastewater generated in the process is reduced, a preparation process is simple, cost is low, and industrialization is easy to realize. According to the method, a large amount of fluorine in UF6 is converted into HF acid through water, the HF acid reacts with potassium hydroxide to finally prepare a byproduct potassium hydrogen fluoride, and the byproduct potassium hydrogen fluoride can be returned to an electrolytic fluorine production unit in a uranium conversion system so as to realize recycling. The U3O8 prepared by the method is more uniform in granularity, good in powder flowability, low in radioactivity, low in fluorine content and convenient for long-term stable storage.

A process for recovering residual uranium from emptied uranium hexafluoride shipping cylinder during cleaning, including rinsing a uranium hexafluoride shipping cylinder with hydrofluoric acid to dissolve a heel of uranium hexafluoride therein to form a mixture of sediment, precipitates and a uranium solution; separating the uranium solution from the sediment and precipitates; mixing sodium hydroxide with the uranium solution to precipitate sodium diuranate; separating the solid sodium diuranate from the sodium fluoride solution formed; re-dissolving the sodium diuranate in sodium carbonate solution to form uranyl carbonate complex solution; and adjusting the pH of uranyl carbonate complex solution further to precipitate uranyl peroxide with the addition of hydrogen peroxide. Sodium fluoride solution produced is further treated to remove fluoride by percolating it through a calcite limestone bed to form calcium fluoride solid.

The invention discloses a method for measuring uranium content in uranium hexafluoride hydrolyzate. The method mainly includes: preparation of uranium standard solution; instrument debugging, software setting, and optimization of working parameters; working curve, analysis limit, and measurement range of uranium Determination; determination and other steps. The analysis is comparable to the GRAV method in precision, high accuracy, stable method, simple and fast, and short time-consuming. It takes less than 1 minute to measure a sample, and hundreds of samples can be processed at one time in one working day. It is suitable for large-scale UF6 hydrolysis. Determination of uranium content in liquid samples.

The invention belongs to the technical field of nuclear fuel cycling and particularly relates to a valve opening device for a uranium hexafluoride pressure heating tank. The valve opening device comprises a rotary box, a left connecting shaft, a cross shaft type universal coupling, a right connecting shaft, a magnetic coupling, a valve, an open valve disk, a spring and a cardan universal joint. Anoutput shaft of the rotary box is connected with the left connecting shaft, the cross shaft type universal coupling, the right connecting shaft, the magnetic coupling, the cardan universal joint andthe rotary box in sequence. The spring is arranged between the open valve disk and the cardan universal joint. The open valve disk is in centering connection with a valve rod of the valve. The valve opening device adopts the magnetic driving mode; the rotary box, the universal coupling, the universal joint, the spring and other parts are arranged on a transmission shaft. It is guaranteed that opening and closing of the angle valve on an internal storage container can be achieved without opening an end cover of the pressure heating tank.

A dry conversion reactor for converting uranium hexafluoride to uranium dioxide, the dry conversion reactor including a gas-phase reaction segment and a fluidized bed segment, wherein at least one of the gas-phase reaction segment and the fluidized bed segment is a replaceable segment. A method for operating a dry conversion reactor utilizing a uranium hexafluoride to uranium dioxide conversion process, the method including replacing at least one conversion reactor segment.

The invention belongs to the technical field of analysis, detection and metering and relates to a system and method applicable to uranium hexafluoride liquefaction and sample distribution. According to the system, a pressure gauge, a drying oven, a channel residual sample collection container and a phi3 valve F7 are respectively connected with the different interfaces of a sample storage channel;the phi3 valve F7 is connected with a phi3 valve bottom flange transition head through a bend connecting piece; a phi3 valve F8 is installed on the phi3 valve bottom flange transition head; the phi3 valve F8 is connected with two sample separate taking openings through sampling connecting pipes; and plugging heads are arranged at the sample separate taking openings respectively. With the system and method of the invention adopted, sample distribution efficiency is improved, the stability of a sample distribution system is enhanced, the labor intensity of operators is reduced, irradiation dosesare reduced, and the security protection and environmental protection risks of posts are controllable.

The invention belongs to a uranium hexafluoride system in the field of uranium enrichment, and particularly relates to a tail spring device applied to a drum vacuum valve in the system. The tail spring device comprises a worm shaft, a nut assembly, a spring assembly and a sleeve assembly, wherein the sleeve assembly is connected to an external system through bolts; when a valve does switch-on and switch-off actions, a motor drives the worm shaft to rotate, the nut assembly linearly slides to a spring base to compress a spring through the action of a U-shaped groove of the sleeve assembly, the spring transmits force borne by the spring to the base, and the force is exerted on the worm shaft through a half gasket; counterforce generated through spring compression hinders rotation of the worm shaft, the friction force of the worm shaft is increased, and then the moment transmitted by the worm shaft is decreased. According to the tail spring device, the compressing force of the spring is borne by the worm shaft, the sleeve assembly does not bear the elastic force any more, therefore, the service life of a sleeve is prolonged, and the reliability of the valve is improved.

The invention relates to the use of a compound of formula KMgF₃ to trap metals present in the form of fluorides and/or of oxyfluorides in a gaseous or liquid phase.

It also relates to a compound of formula KMgF₃ which has a surface specific area at least equal to 30 m²/g and at most equal to 150 m²/g and also to its methods of preparation.

The invention notably finds application in the nuclear industry, in which it can advantageously be used to purify uranium hexafluoride (UF₆) present in a gaseous or liquid stream, with regard to metal impurities which are also present in this stream.