232 results about "Thorium" patented technology

The invention relates to a crown ether covalent organic framework material as well as a preparation method and application thereof. The preparation method comprises the following steps: dissolving a crown ether monomer and an amino monomer in an o-dichlorobenzene solvent, adding a catalyst acetic acid aqueous solution, and uniformly dispersing to form a reaction solution; the method comprises the following steps: freezing a reaction solution through a liquid nitrogen bath, vacuumizing, filling nitrogen, carrying out a heating reaction, and carrying out post-treatment operations such as washing and drying on a reaction product, so as to obtain the crown ether covalent organic framework material BADT-DB18C6-COF. The crown ether covalent organic framework material is formed by condensation of a crown ether aldehyde monomer and an amino monomer, the crown ether monomer is 4, 4 ', 4' ', 4 '''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde, and the crown ether monomer is 4, 4 ', 4' ', 4'''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde. The amino monomer is 4 ', 5'-bis (4-aminophenyl)-[1, 1 ': 2', 1 ''-terphenyl]-4, 4''-diamine. The crown ether covalent organic framework material BADT-DB18C6-COF prepared by the invention is mainly used as an adsorbent, is used for adsorbing and recovering uranium, thorium and americium ions from a solution, shows good adsorption capacity on radionuclides U-238, Th-232 and Am-243, and is particularly suitable for the field of spent fuel post-treatment in the nuclear industry.

The invention discloses an actinide purification method and device, and belongs to the technical field of medical radioactive isotope separation and purification. The method comprises the following steps: passing a thorium target solution containing actinide through a cation exchange column, and leaching with inorganic acid to obtain a first effluent containing Ac < 3 + >; enabling the first effluent containing the Ac < 3 + > to pass through a TK221 resin extraction column to obtain a second effluent containing the Ac < 3 + >; the second effluent containing Ac3 + passes through a TRU resin extraction column, and actinide is obtained. The TK221 resin also contains phosphine oxide and a small amount of long alkyl chain alcohol on the basis of diglycolamide, and shows better performance on purification of Ac < 3 + >. Meanwhile, the elution acidity in the TRU resin separation process is properly improved, and the outflow speed of Ac, La and Ce can be delayed, so that the effect of separating Ac from La and Ce is achieved, and the yield of 225Ac is improved while the concentration of metal impurities is reduced.

The invention provides a hafnium-based adsorption material, a preparation method thereof and a method for removing thorium ions in nuclear waste liquid. The hafnium-based adsorption material comprises an MOF with hafnium ions as a metal center and ellagic acid as an organic ligand, and in an X-ray diffraction pattern of the MOF under CuK alpha radiation, at 5 degrees lt; 2 [theta] lt; and in the range of 20 degrees, four diffraction peaks at the positions of 5.0 degrees + / -0.3 degrees, 8.6 degrees + / -0.3 degrees, 13.2 degrees + / -0.3 degrees and 17.2 degrees + / -0.3 degrees exist. The hafnium-based adsorption material disclosed by the invention has improved adsorption capacity on thorium ions under the condition of low pH value.

According to the high-precision while-drilling natural gamma-ray spectrum imager, a high-precision gamma-ray spectrum detector is selected as a sensor module, a high-energy-resolution energy spectrum diagram is obtained through energy correction processing, a stable window method is selected for energy spectrum drift correction of instrument temperature change, and the energy spectrum drift correction step is simplified; the energy spectrum drift channel number and the energy window moving range are determined by searching a K peak, a U peak and a Th peak so as to ensure the accuracy of energy window division, the high-precision potassium, uranium and thorium element content measured in real time in each sector in the stratum is obtained after spectrum unfolding, data processing, interpolation imaging and other operations are carried out according to the measurement result of each sector, and the accuracy of energy window division is improved. The real-time azimuth imaging graph of each sector is displayed, parameters such as the organic carbon content and the shale type can be accurately judged according to the imaging graphs, more accurate and more comprehensive information is provided for geosteering, the drilling success rate is greatly increased, and the operation risk is reduced.

The invention discloses a method for removing metal impurities through electrolytic coupling vacuum distillation of tin benzene sulfonate. The method comprises the following steps: step 1, preparing a liquid; step 2, preparing an electrolyte; step 3, electrolysis; 4, copper removal and refining; and step 5, vacuum distillation. According to the method, the crude tin is prepared into the electrolyte, the benzene sulfonate is added for electrolysis, the precipitation potential of lead and tin is changed, the impurity lead element with high difficulty is removed, other impurity elements in the crude tin can also be removed, the lead, copper and other impurity elements are further removed by combining copper removal refining and vacuum distillation, and the 5N or 6N standard high-purity tin is obtained. The content of the lead element and the content of the bismuth element are both smaller than 1 ppm, the content of the uranium element and the content of the thorium element are both smaller than 1 ppb, the requirement of a low-alpha-ray-dose material is met, the problem that radioactive alpha-particle impurity elements in crude tin are difficult to remove through an electrolytic method is effectively solved, the use requirement that the high-purity tin serves as a welding flux is met, and the quality of a semiconductor chip is greatly improved.

The invention discloses a process for removing uranium and thorium from hydrochloric acid system zirconium and hafnium waste acid. The process comprises the following steps: S1, adding a substance containing nitrate radicals into the zirconium-hafnium waste acid of a hydrochloric acid system to obtain zirconium-hafnium waste acid of a nitrate acid mixed acid system; s2, extracting the zirconium and hafnium waste acid of the mixed acid system obtained in the step S1 by adopting a TBP extracting agent to obtain an organic phase 1 containing uranium, thorium, zirconium and RE and a water phase 1 mainly containing RE; s3, the organic phase 1 obtained in the step S2 is subjected to reverse extraction through a nitric acid solution, and an organic phase 2 containing uranium and thorium and a water phase 2 mainly containing zirconium and RE are obtained; and S4, the organic phase 2 obtained in the step S3 is subjected to reverse extraction through sulfuric acid, and a blank organic phase 3 and a water phase 3 containing uranium and thorium are obtained. The process provided by the invention can realize removal of radioactive uranium and thorium in the hydrochloric acid system zirconium-hafnium waste acid. The process is high in uranium and thorium removal rate, low in zirconium and rare earth loss rate and beneficial to further comprehensive utilization of valuable elements in the waste acid.

The invention discloses carboxyl-rich chitosan gel and preparation thereof and separation and enrichment of radioactive element thorium, and the method comprises the following steps: mixing and dissolving chitosan, water and glacial acetic acid, adding 1, 2, 3, 4-butanetetracarboxylic acid, and uniformly mixing; heating at 80 DEG C to completely volatilize moisture, and heating to 100 DEG C for reaction to obtain a solid product; taking out a solid product, soaking the solid product in deionized water, and performing freeze drying treatment to obtain the carboxyl-rich chitosan gel, namely the CCS-BTCA xerogel. The problems that an existing adsorbent is poor in stability and low in adsorption activity are solved, the prepared material has the high adsorption rate, the optimal adsorption effect can be achieved at the temperature of 298 K and the pH value of 3, the adsorption capacity is 228.9 mg / g, and the maximum theoretical adsorption capacity can reach 279.4 mg / g.

The invention relates to the technical field of nuclear fuel cladding, in particular to a diffusion impervious layer based on an accident fault-tolerant fuel cladding Cr coating and a preparation method of the diffusion impervious layer, the diffusion impervious layer is composed of any one of thorium Th, niobium Nb, molybdenum Mo and a SiCf / SiC composite coating, and a pure Cr coating is arranged outside the diffusion impervious layer. According to the nuclear power cladding tube, the diffusion barrier layer is arranged between Zr and Cr, so that the Zr matrix and the Cr coating are separated, and diffusion between the Zr matrix and the Cr coating is slowed down, so that the oxidation speed of the Cr coating is reduced, the water loss accident resisting capacity of the cladding is improved, and then the nuclear power cladding tube is safer under the LCOA working condition; due to the fact that the diffusion barrier layer can effectively restrain mutual diffusion between the Cr coating and the zirconium alloy matrix, the high-temperature service performance of the coating is improved, the high-temperature oxidation resistance of the accident fault-tolerant fuel cladding Cr coating is further improved, and the survival time of a fuel cladding pipe under the LOCA accident can be prolonged.

The present invention relates to the technical field of filament cathode materials, and in particular to a tungsten alloy wire, a preparation method and use thereof, and a tungsten carbide alloy filament. The tungsten alloy filament comprises the following components: the tungsten alloy comprises the following components: 0.0005-0.3wt% carbon element, 0.25-2.6wt% M element, 0.05-0.5wt% oxygen element, the remainder being tungsten element and inevitable impurities, and the components do not contain thorium element; the initial recrystallization temperature of the tungsten alloy wire is 48%Fc-56%Fc, and / or, the average size of its 80%Fc recrystallized grains is (1-15) μm. The tungsten alloy wire does not contain thorium element and does not have radioactive pollution problem. Its initial recrystallization temperature is between 48%Fc and 56%Fc. Its processing performance is greatly improved compared with thoriated tungsten material. Compared with thoriated tungsten wire, the wire has fewer flaw detection crack points. It is used as the matrix of tungsten carbide alloy filament. After carburization, the tungsten matrix has fine grains. The filament is not easy to break during the production and use of magnetron coils (i.e., filaments), and the production yield and quality are improved.

Systems and methods for generating radionuclides, such as radium-224. Systems herein may include a first cartridge having a first opening, a second opening, and a chamber therebetween having a first resin having affinity for thorium-228 and bismuth-212; a second cartridge having a first opening, a second opening, and a chamber therebetween having a second resin having affinity for thorium-228 and bismuth-212, a third cartridge having a first opening, a second opening, and a chamber therebetween comprising a third resin having affinity for thorium-228 and bismuth-212, a fourth cartridge having a first opening, a second opening, and a chamber therebetween having a third resin having affinity for lead-212; wherein a continuous flow path is formed from a top of the first cartridge though the second cartridge, through the third cartridge, and to a bottom of the fourth cartridge during system use.

The invention relates to low-alpha-ray aluminum oxide and a preparation method thereof, and belongs to the technical field of aluminum oxide powder preparation. Raw materials of the low-alpha-ray aluminum oxide comprise a sodium aluminate solution, the sodium aluminate solution contains thorium and uranium, and the method comprises the following steps: carrying out a first decomposition reaction on the sodium aluminate solution by using a first aluminum hydroxide seed crystal to obtain a first decomposition solution; carrying out solid-liquid separation on the first decomposition solution to obtain a liquid-phase substance; performing a second decomposition reaction on the liquid-phase substance by using a second aluminum hydroxide seed crystal to obtain aluminum hydroxide; dissolving aluminum hydroxide with sulfuric acid to obtain an aluminum sulfate solution; performing chemical reaction on the aluminum sulfate solution and ammonium sulfate to obtain aluminum ammonium sulfate; re-crystallizing the aluminum ammonium sulfate for multiple times to obtain high-purity aluminum ammonium sulfate; and roasting the high-purity aluminum ammonium sulfate to obtain the low-alpha-ray aluminum oxide. The purity of the low-alpha-ray aluminum oxide is greater than or equal to 99.99%, the thorium content is less than or equal to 5ppb, the uranium content is less than or equal to 5ppb, and the median particle size is adjustable in a range of 1-50 microns.

The invention relates to an aminophosphonate functionalized carbon material as well as a preparation method and application thereof. The amino phosphonate functionalized carbon material comprises a carbon material matrix and a structure which is covalently connected with the carbon material matrix and is shown as a formula I. The amino phosphonate functionalized carbon material can be applied to a capacitive deionization technology to selectively adsorb thorium from a mixed solution containing thorium and rare earth, and has high thorium adsorption capacity. # imgabs0 #

The invention belongs to the field of medical isotope separation and preparation, and particularly relates to a method for separating and purifying < 212 > Pb from a thorium solution, which comprises the following steps: 1) leacheate is introduced into a Pb selective adsorption column to pretreat the Pb selective adsorption column, and then < 232 > Th salt solution and leacheate are introduced in sequence; 2) continuously introducing a desorption solution into the Pb selective adsorption column, enabling the desorption solution passing through the Pb selective adsorption column to flow into the Th selective adsorption column, and collecting the outflowing desorption solution; a desorption solution only containing 212Pb is obtained; according to the method, the 212Pb in the high-concentration thorium solution can be separated and purified at the same time. While < 232 > Th impurity entrainment and pollution in a < 212 > Pb product are remarkably reduced, compared with a single-column process, the separation time is basically not increased. The separated 212Pb product has high radionuclide purity and chemical purity at the same time, and can meet related pharmaceutical requirements.

The present application belongs to a kind of medical radioisotope preparation field, specifically related to a kind of method for extracting lead-212 and bismuth-212 in thorium-232 decay chain, solution containing thorium-232 decay chain material is passed through anion exchange resin column, the solution containing thorium-232 decay chain material includes halogen ion, and the concentration of hydrogen ion in the solution containing thorium-232 decay chain material is greater than 0.01mol / L, lead-212 and bismuth-212 are absorbed by anion exchange resin column, and other nuclides directly pass through anion exchange resin column and are not absorbed, lead-212 and bismuth-212 are desorbed from anion exchange resin column by elution reagent, and lead-212 and bismuth-212 are separated and extracted;The present application can extract lead-212 and bismuth-212 simultaneously, and the extraction speed is fast, the purity is high, and the product is not easy to cause organic pollution.

The invention discloses a method for recovering uranium, thorium and rare earth from associated tantalum-niobium ore, which comprises the following steps: grinding the associated tantalum-niobium ore to a preset particle size, carrying out heating reaction on the ground associated tantalum-niobium ore and a dilute sulfuric acid solution to obtain a leaching solution containing uranium, thorium and rare earth, and then carrying out solid-liquid separation to obtain a leaching solution clear solution and a filter cake; adding a sodium hydroxide solution into the leachate clear liquid, controlling the pH value, then separating and extracting uranium by adopting 201 * 7 strongly basic anion exchange resin to obtain uranium saturated resin and adsorption tail liquid, leaching the uranium saturated resin to obtain uranium leaching qualified liquid, and precipitating the uranium leaching qualified liquid through sodium hydroxide to obtain sodium diuranate; carrying out extraction treatment on the adsorption tail liquid to obtain a saturated organic phase and raffinate; carrying out back extraction treatment on the saturated organic phase to obtain thorium carbonate back water, and carrying out sulfuric acid precipitation to obtain thorium sulfate; and adding a sodium carbonate solution into the raffinate for neutralization and precipitation, and filtering to obtain the rare earth carbonate. The method aims at improving the metal extraction rate of the associated tantalum-niobium ore.

A tungsten alloy wire, a preparation method therefor and a use thereof, and a tungsten carbide alloy filament. The tungsten alloy wire comprises the following components: 0.0005-0.3 wt% of a carbon element, 0.25-2.6 wt% of an M element, 0.05-0.5 wt% of an oxygen element and the balance being a tungsten element and inevitable impurities, and does not comprise a thorium element; and the initial recrystallization temperature of the tungsten alloy wire ranges from 48 %Fc to 56 %Fc, and / or the average size of 80 %Fc recrystallization grains is 1-15 μm. The tungsten alloy wire does not comprise the thorium element, does not cause radioactive contamination, has the initial recrystallization temperature ranging from 48 %Fc to 56 %Fc, and has the processing performance greatly improved compared with that of a thorium-containing tungsten wire; compared with a thorium-containing tungsten wire, the tungsten alloy wire has less flaw detection crack points, when the tungsten alloy wire is used as a substrate of a tungsten carbide alloy filament, grains of the carbonized substrate are fine, the filament is not easy to break in the production and use of a magnetron coil, and the production yield and quality are improved.

The invention relates to a method for machine learning aided design of a structural alloy for a fourth-generation thorium-based molten salt reactor, and belongs to the technical field of artificial intelligence auxiliary material design. The method comprises the steps that S1, an alloy component machine learning model for phase composition prediction is established, specifically, based on known alloy sample data, the sample data comprise alloy element content and corresponding phase composition information, and the machine learning model used for predicting alloy phase composition is established; s2, screening an initial alloy component space through phase composition constraint conditions; s3, secondary screening of alloy components based on thermodynamic simulation; s4, alloy sample preparation and solid solution homogenization treatment; s5, carrying out a mechanical property test; and S6, experimental result feedback and machine learning model closed-loop updating are carried out. According to the method, the artificial intelligence technology and the traditional alloy design theory are organically fused, an efficient and reliable new path is provided for research and development of key structural materials for the fourth-generation thorium-based molten salt reactor, and the important scientific research value and the engineering application prospect are achieved.

This invention discloses a Th / Ra generator containing natural thorium and its production. 224 The Ra method. The generator of this invention includes a cation exchange resin column and bottles I to VI; wherein bottle I is filled with the stock solution; bottle III is filled with deionized water; bottle IV is filled with 4.0–6.0 M nitric acid; bottles II, V, and VI are empty; bottle V is a specially designed collection bottle with an extraction resin column at its opening for collection. 224 Ra. The method of the present invention includes the following steps: 1) using a mixture containing 228 1) A natural thorium mixture of Th and Ra was concentrated to obtain a concentrate, and then the pH of the system was adjusted to serve as the packing stock solution for the Th / Ra generator; 2) The stock solution was passed through a cation exchange resin column to obtain a cation exchange resin column loaded with thorium. 224 Ra, and collect the eluent from the original solution for later use; 3) Elute the cation exchange resin column with inorganic acid. 224 Ra, then the eluent is collected using a special collection bottle to obtain... 224 Ra. This invention enables the separation, purification, and recycling of [product name]. 224 Ra, to overcome 228 Th / 224 The Ra generator has a stringent loading process to avoid the radioactive risks associated with handling tons of natural thorium.

The application relates to the technical field of quartz crystal purification, and particularly discloses a method for eliminating radioactive elements in quartz crystals, which comprises the following steps: adding quartz sand into a composite cleaning solution, stirring and cleaning, then performing solid-liquid separation and water washing to obtain pretreated quartz sand; performing heat treatment on the pretreated quartz sand in an oxygen-containing atmosphere at 850-1000 DEG C for 2-4 hours to obtain heat-treated quartz sand; rapidly cooling the heat-treated quartz sand in cooling water at 10-30 DEG C for 15-30 seconds, then crushing and screening to obtain crushed quartz sand; adding the crushed quartz sand into a selective acid cleaning solution under stirring to perform an acid cleaning reaction, then performing solid-liquid separation, washing and drying after the reaction to obtain low-radioactivity quartz sand. The composite cleaning solution and the selective acid cleaning solution are used in cooperation to efficiently remove radioactive elements such as uranium and thorium in the quartz crystals, and the physical and chemical properties of the quartz crystals can meet the requirements of high-end semiconductor and optical applications.

Provided are: high-purity alumina particles which have improved purity by reducing the amount of impurities remaining in alumina particles derived from metallic aluminum and also by reducing the amount of metallic aluminum; and a method for producing the high-purity alumina particles. The aluminum oxide particles derived from metallic aluminum have a uranium element content of not more than 5 ppb, a thorium element content of not more than 5 ppb, a metallic aluminum content of not more than 5 ppm, a sodium content of not more than 10 ppm, and an iron content of not more than 10 ppm, and a metallic aluminum powder is introduced into a flame to be gasified and oxidized, and is spheroidized by surface tension during cooling to form spherical aluminum oxide particles. Spherical alumina particles are introduced into an alkaline solution to dissolve metallic aluminum remaining in the particles, and the spherical alumina particles are dried.

The application relates to a thorium and spallation element separation method, which comprises the following steps: injecting a thorium-based solution into column 1 to obtain a first effluent, washing column 1 with a first inorganic acid solution and a second inorganic acid solution in sequence to obtain a first washing solution and a first thorium solution respectively; injecting the first thorium solution into column 2 to obtain a second effluent, washing column 2 with a third inorganic acid solution and a second inorganic acid solution in sequence to obtain a second washing solution and a second thorium solution; mixing the first thorium solution and the second thorium solution, injecting into column 3 to obtain a third effluent, and washing column 3 with a fourth inorganic acid solution to obtain a third thorium solution. The method does not introduce other foreign reagents and is not polluted by foreign impurity ions, and the high ion potential of Th is utilized to preferentially adsorb on a cation column to achieve the purpose of separation from other spallation elements.

The invention relates to the technical field of filament cathode materials, in particular to a tungsten carbide alloy filament, a preparation method thereof and a heating device. The tungsten carbide alloy filament comprises a base body and a carbonization layer attached to the outer surface of the base body. The grain width of the carbonized layer is 1.0 [mu] m to 15.0 [mu] m; the components of the matrix and the carbonized layer comprise a tungsten element, a carbon element, an oxygen element and an M element; the M element is selected from one or a combination of more of La, Y, Sc, Nd, Sm, Lu, Ce, Gd, Tb, Dy, Ho, Pr, Er, Tm, Yb, Eu, Hf and Zr; wherein the component of the tungsten carbide alloy filament does not contain a thorium element. The emission stability of the tungsten alloy lamp filament is improved, the stable emission capability of a traditional thorium tungsten lamp filament can be achieved under the condition that no thorium element is added, the problem that an existing tungsten lamp filament without thorium is poor in stable emission capability is solved, and the problem of radioactive contamination of the traditional thorium tungsten lamp filament is solved.

The present invention provides an N,N-dihydrocarbyl amide carboxylic acid, preparation method therefor and use thereof. The N,N-dihydrocarbyl amide carboxylic acid can be used as an extractant for enriching rare earth elements from raw materials containing low-concentration rare earth elements, separating and purifying yttrium element from mixed rare earth raw material, and separating elements such as aluminum, iron, radioactive thorium, radioactive uranium and actinide from mixed rare earth raw material, etc. The compound can be synthesized in a simple and cost-efficient way. As an extractant, it has good chemical stability and can withstand strong acid and strong alkali without decomposition.

The invention belongs to the technical field of hydrometallurgy, and in particular relates to a method for removing thorium from a lutetium-containing liquid. The invention provides a method for removing thorium from a lutetium-containing liquid, comprising the following steps: extracting the lutetium-containing liquid with a metal extract to obtain an organic phase; the volume ratio of the metal extract to the lutetium-containing liquid is 5-10:1; mixing the organic phase with a hydrochloric acid solution for back extraction to obtain an aqueous phase; the molar concentration of the hydrochloric acid solution is 6.3-6.7 mol / L, and the volume ratio of the organic phase to the hydrochloric acid solution is 5-10:1; mixing the aqueous phase with an oxalic acid solution for precipitation to obtain a crude lutetium oxalate product; the mixing is carried out under stirring conditions, the stirring temperature is 58-62°C, and the stirring speed is 100-200 r / min; washing the crude lutetium oxalate product to obtain lutetium oxalate; the washing solvent is a mixed solution of ammonium oxalate solution and ethanol; the mass ratio of the crude lutetium oxalate product to the solvent is 1:3-5; and calcining the lutetium oxalate to obtain lutetium oxide.

The application discloses a kind of uranium nuclide automatic extraction device and method, wherein, device includes dissolving target device, acid regulator, extractor, purifier, waste liquid tank, product tank, control system and shielding box, dissolving target device, acid regulator, extractor, purifier, waste liquid tank, product tank are connected by conveying pipe and valve, liquid conveying power is provided by peristaltic pump, control system is set in shielding box, and relevant components in shielding box are electrically connected.The application also describes a method for extracting a uranium nuclide solution substantially free of impurities from an irradiated thorium target material.The uranium nuclide automatic extraction device of the present application can achieve rapid and efficient extraction and purification of uranium nuclides from irradiated thorium target material.The device has the characteristics of simple structure, compactness and automation operation, which greatly improves the production efficiency and reduces the radiation dose received by the operator.

This invention relates to an aminophosphonate-functionalized carbon material, its preparation method, and its uses. The aminophosphonate-functionalized carbon material comprises a carbon matrix and a structure of Formula I covalently linked to the carbon matrix. It can be applied to capacitive deionization technology to selectively adsorb thorium from a mixed solution containing thorium and rare earth elements, exhibiting a high adsorption capacity for thorium.

Voltage power source device having a connected structure in which each one of ends of a first metal and a second metal that generate the Seebeck effect are connected by a one-end side bonding part in which uranium and thorium are supported on a carrier made of boron, and the other ends are connected by an other bonding part in which uranium and thorium are supported on a carrier made of carbon. The one-end side bonding part is raised to a first temperature through α-decay of uranium (U) and thorium (Th), and the other bonding part is raised to a second temperature that is different from the first temperature through α-decay of uranium (U) and thorium (Th), whereby a current is generated by the Seebeck effect based on the relative temperature difference between the two bonding parts. Voltage power source device and power generation method enable next-generation type energy supply.

This invention belongs to the field of uranium resource exploration technology, specifically relating to a method and apparatus for detecting gamma rays. The method includes: utilizing... 241 The equivalent peak position after the interaction of Am and NaI(Tl) crystals serves as a self-stabilizing peak, providing a stable gamma-ray reference peak and achieving automatic spectrum stabilization. A digital high-speed multichannel module acquires gamma-ray signals, achieving full acquisition of gamma-ray signals in the energy range of 30keV to 3000keV. Multiple Regions of Interest (ROIs) are set for the acquired gamma-ray signals using digital energy thresholds, obtaining CPS counts for the quantitative gamma energy range, the natural gamma energy range, and the uranium-thorium-potassium (UHT) energy range of the gamma spectrum. Using the calibration coefficients of each parameter in the standard model, and based on the CPS counts for the quantitative gamma energy range, the natural gamma energy range, and the UHT / UHT energy range of the gamma spectrum, the quantitative gamma, natural gamma, and gamma spectrum detection results are output. This invention enables a single probe to simultaneously output measured data for quantitative gamma, natural gamma, and gamma spectra, improving the efficiency of acquiring these data.

The invention discloses a preparation method of 5N-grade ultra-pure lutetium oxide, which comprises the following steps: removing thorium and uranium from a lutetium chloride solution, carrying out primary precipitation with oxalic acid, firing to obtain a primary lutetium oxide product, re-dissolving the primary lutetium oxide product with nitric acid, carrying out secondary precipitation with ammonia water, and firing to obtain the 5N-grade ultra-pure lutetium oxide. According to the method disclosed by the invention, the preparation of lutetium oxide with higher purity can be realized by combining extraction with impurity grading removal of oxalic acid-ammonia water double precipitation so as to meet quality index requirements required by application of materials with higher performance. The preparation method is high in stability, high in yield and suitable for industrial production and application.

The present invention discloses a method for extracting uranium and thorium elements, belonging to the technical field of rare earth liquid treatment. The method includes mixing an alkyl phenyl oxovaleric acid extractant with an organic solvent to obtain an organic phase; mixing the organic phase with a rare earth solution containing uranium and thorium elements for extraction, so that the rare earth remains in the aqueous phase and the uranium and thorium elements enter the organic phase. This method has fewer extraction stages, a simple process, good separation effect, and high extraction rates of uranium and thorium.