36 results about "Americium" patented technology

The invention discloses a method for identifying coal gangue based on gamma rays. Double-energy gamma rays consisting of low-energy americium 241 gamma rays and moderate-energy cesium 137 gamma rays are adopted; and when the double-energy gamma rays penetrate through a coal gangue mixture, the attenuation degree of the low-energy americium rays is only related with coal gangue mixing degree, thickness, bulk density and the like of the coal gangue mixture, and the attenuation degree of the moderate-energy cesium gamma rays is only related with the bulk density and the weight of the mixture. The gangue rate of the coal gangue mixture is calculated according to the attenuation degrees of the two kinds of rays. A double-energy gamma ray source and a gamma ray probe are respectively arranged on two sides of a chute of a scraper conveyor in the back of a top coal caving support; a data analyzer is used for calculating the gangue rate of the coal gangue mixture according to the intensity information of the gamma rays detected by the probe and comparing the gangue rate with a set gangue rate threshold value; and when the gangue rate reaches or exceeds the set gangue rate threshold value, a coal discharge window is closed through an electro-hydraulic control system to stop discharging coal, so that automatic control of coal discharge is realized.

A system and method for the extraction of americium from radioactive waste solutions. The method includes the transfer of highly oxidized americium from an acidic aqueous feed solution through an immobilized liquid membrane to an organic receiving solvent, for example tributyl phosphate. The immobilized liquid membrane includes porous support and separating layers loaded with tributyl phosphate. The extracted solution is subsequently stripped of americium and recycled at the immobilized liquid membrane as neat tributyl phosphate for the continuous extraction of americium. The sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source, and the remaining constituent elements in the aqueous feed solution can be stored in glassified waste forms substantially free of americium.

A process which allows separation of americium present in an acid aqueous phase or in an organic phase from the other metal elements also found in this phase, by complexation of the americium with a water-soluble ethylenediamine derivative; and a process for selective recovery of americium from an acid aqueous phase containing, in addition to americium, other metal elements, which comprises the application of this separation process.

The invention relates to a nitride nuclear fuel characterized in that the nitride fuel is a pellet of a material with a single-phase solid solution of elements comprising at least a nitride of americium (Am), and that the material has a density of around 90% of the theoretical density. The invention further relates to a method for producing the said nuclear fuel by using the steps: mixing of starting powders, sintering of the powders into a dense pellet and a subsequent heat treatment.

The method shows how to violate predictions of quantum mechanics for matter. For light, the method has been disclosed in patent application Photon Violation Spectroscopy. The methods are different in specifying different methods for light and matter. For matter, the method typically uses the single 5.5 MeV alpha (He++) emitted from the radioisotope Americium-241 in spontaneous decay, a thin gold foil beam splitter, and two surface barrier alpha detectors. The detectors deliver a characteristic electrical pulse with amplitude proportional to matter wave energy. A circuit reads the coincidence rate and singles rates of pulses from the two detectors. Quantum mechanics predicts that the particle would go one way or the other at the beam splitter, and coincident detections of pulses characteristic of such a particle would occur only at an easily calculated chance rate. However, the method at hand shows such characteristic pulses occur in coincidence at a rate greatly exceeding chance. By exceeding chance the method demonstrates surpassing a binding energy threshold and predictions of quantum mechanics. The degree above chance is a new measure in fundamental physics and is usable as a material science probe of the beam splitter. An apparatus specially designed to test for the absence of true coincidences and then perform a beam splitting test to show split-beam coincidences becomes useful in applying the method of Particle Violation Spectroscopy to material science. Fundamental discoveries in physics have been made with this method; therefore it is a method of discovery in physics.

The invention provides compound SBS modified asphalt and a preparation method thereof. The method comprises the following steps: metering petroleum asphalt, rapidly heating the petroleum asphalt to 170-180 DEG C and sending the petroleum asphalt to a mixer; metering and adding SBS polymers, rubber powder, coal tar, dimethylamino pyridine and butyl americium rubbers, and after the substances are mixed and fully swollen in the mixer, fully grinding the mixture in a colloid grinder to reach the required disperse mixed melt state; sending the ground mixture to a development tank, adding a stabilizing agent dibenzoic acid dibutyl tin and developing the mixture for 1h under stirring at 180 DEG C and low speed, thus obtaining the qualified product. The asphalt is prepared by mixing 42.3% of petroleum asphalt, 35% of rubber powder, 15% of coal tar, 2% of SBS polymers, 3% of dimethylamino pyridine, 2.5% of butyl americium rubbers and 0.2% of dibenzoic acid dibutyl tin. The rubber powder can be added at high proportion and the high / low temperature performance value and the storage stability of the finished product are improved.

The invention discloses a fast separation method of activation products of americium, uranium and gallium. The fast separation method comprises preparation before separation, chromatographic separation and process control. The separation method comprises making a radioactive solution of fission product-rich activation products of americium, uranium and gallium flow through a P204 or LN chromatographic column, an Al2O3 chromatographic column, a TBP chromatographic column and a DN chromatographic column at a certain flowing rate so that americium, uranium and gallium adsorption is realized, orderly making dilute acid and peracid solutions go through a DN chromatographic layer-P204 or LN chromatographic layer series column and a P204 or LN chromatographic layer-TBP chromatographic layer series column so that americium and uranium readsorption separation is realized and making a desorption solution go through chromatographic column groups corresponding to americium, uranium and gallium, wherein the eluants respectively are radioactivity measurement solutions of activation products of americium, uranium and gallium. The fast separation method has an americium recovery rate higher than 75%, a uranium and gallium recovery rate higher than 90%, a fission product decontamination factor higher than 104 and separation process time of 3h.

The invention discloses a determination method of plutonium age in a trace plutonium sample. The method comprises the following steps of: adding plutonium and the isotopic diluents of daughter americium thereof, namely, Pu and Am, in the sample; adding sodium nitrite for price adjustment; and then loading a TEVA resin column, using a chemical separation method to respectively obtain a Pu component and an Am component, and using a multi-receiving inductive coupling plasma mass spectrometry to measure the Pu component and the Am component respectively to obtain the values of Pu / Pu and Am / Am . According to the amount of the added diluents, the value of Pu / Am is calculated and substituted in the age-calculating formula to obtain the age of plutonium. The deviation between the age measured by the method and the reference value is within 5%.

The invention relates to the ROANEX method, which extracts actinides from used nuclear fuel in a single purification cycle. The used nuclear fuel contains actinides, U, Am, Pu, Np. and Cm, and fission products, Cs, Sr and Tc. The fission products are separated first from the used nuclear fuel. The actinides are reduced to their lowest oxidation states and then oxidized to their highest oxidations states. Uranium, Pu and Np move to an organic phase solution and Am and Cm move to a nitrate solution. Uranium, Pu, and Np are stripped from the organic phase solution, and then treated with an oxalic acid to form a precipitate. Americium and Cm are treated with a potassium carbonate solution and Am precipitates. Actinides Am, U, Pu, and Np precipitates are heated in an oven and then blended together to form a mixed oxide fuel of UO2, PuO2, NpO2 and AmO2.

The invention provides rubber compound modified asphalt and a preparation method thereof. The method comprises the following steps: petroleum asphalt is conveyed by an asphalt pump, metered by a mass flowmeter and heated to 170-180 DEG C by a rapid heating heat exchange system; and the heated hot asphalt, rubber powder and additives are added to a mixer after being proportionally metered to be stirred for mixing and swelling, and the reaction time is 40min. The rubber compound modified asphalt is prepared by mixing 39.3% of petroleum asphalt, 35% of rubber powder, 15% of coal tar, 5% of cement, 3% of dimethylamino pyridine, 2.5% of butyl americium rubbers and 0.2% of dibenzoic acid dibutyl tin. Scrap tire rubber powder is added at high proportion; therefore, the rubber compound modified asphalt has great environmental protection significance, and at the same time, the wax content in the finished asphalt is reduced, the applicable scope of temperature is expanded to minus 40-90 DEG C and the 180 DEG C rotary viscosity value is kept between 1.5Pa.s and 2.5Pa.s, thus improving the construction workability.

A process which allows separation of americium present in an acid aqueous phase or in an organic phase from the other metal elements also found in this phase, by complexation of the americium with a water-soluble ethylenediamine derivative; and a process for selective recovery of americium from an acid aqueous phase containing, in addition to americium, other metal elements, which comprises the application of this separation process.

The invention discloses a determination method of plutonium age in a trace plutonium sample. In the method, a sample of plutonium is taken and measured. A scintillator is added to measure Beta counting rate for obtaining the amount of Pu. In another similar sample of plutonium, the isotopic diluent of americium, namely, Am, is added to conduct price adjustment, a TEVA resin column is added, nitrate is used for leaching and collecting the Am component, Alpha spectrum is used for measuring the value of Am / Am . The amount of Am is calculated and substituted into the age-calculating formula to obtain the age of plutonium. The method has simple chemical separation process and high recovery rate. The sample analysis time is short, the instrument operation is simple, and the method accuracy is high.

The invention belongs to the technical field of waste metal treatment, and particularly relates to a method for separating metal waste containing 241Am. The invention provides the method for separating the metal waste containing the 241Am. The separation method comprises the steps that 1, dry smelting is carried out; 2, Ag electrolytic refining is carried out; and 3, Au electrolytic refining is carried out. According to the technical scheme provided by the invention, thorough separation of four elements can be realized through treatment, three precious metals in the four elements meet radioactivity decontrol requirements through detection, the radionuclide 241Am exists in the forms of americium metaborate waste residues, concentrated americium salt and the like, and then packaging is performed for treatment and storage. The three precious metals meeting the radioactivity decontrol requirements can be reutilized. After the waste is treated, waste minimization and resource utilization maximization are achieved, social benefits and economic benefits are maximized, and precious metal leftover materials containing the 241Am generated in a source piece manufacturing process are effectively separated.

The invention provides a method for determining the age of a plutonium-containing particle. The method comprising the following steps: 1) dissolving a plutonium-containing particle sample to obtain aplutonium-containing solution; 2) adding a diluent to the plutonium-containing solution, separating and purifying to obtain a plutonium component and an americium component; 3) measuring a parameter,and calculating the quality of 241Pu; 4) measuring a parameter, and calculating the quality of 241Am; 5) measuring an activity ratio; 6) calculating the age of the plutonium-containing particle by a formula. According to the invention, the method for determining the age of a plutonium-containing particle is established, the problems that the high-temperature-generated plutonium oxide particles aredifficult to be effectively dissolved, the measuring method relies on the same matrix standard sample, and the existing mass spectrometry cannot accurately determine the plutonium-containing particlesample with low 240Pu abundance, are solved, the adaptability of the method is improved, and the method can be applied to the age determination of various plutonium-containing particles.

The invention relates to the technical field of nuclear fuel aftertreatment and particularly discloses a device for measuring americium and cesium in aftertreatment feed liquid. A NaI detector is arranged in an area defined by central holes of a lead sheath A, a lead sheath B and a lead sheath D, and a connecting line of the NaI detector penetrates out of a slot in the lead sheath B to be connected with a multi-channel spectrometer; the multi-channel spectrometer is connected with a computer through a data line; and gamma ray signals are converted into electric signals by the NaI detector, theelectric signals are screened by the multi-channel spectrometer and then transmitted into the computer, and automatic sample measuring is realized. The device can effectively shield interference of environmental gamma rays in a measuring system, the multi-channel spectrometer is utilized, thus the gamma radioactivity concentrations of cesium-137 and americium-241 can be measured simultaneously, and the measuring result is directly read by software.

Disclosed is a full-wave band spectral scanner. The full-wave band spectral scanner mainly comprises a light source device, a photoseparation device, a primary reflecting device, a standard device, a secondary reflecting device, a dimming control device, a third reflecting device, a sample feeding device, a fourth reflecting device, a signal amplifying device, a signal correcting device, a signal analyzing device and a signal recording device. An outer cover of the light source device is made of hafnium and rutherfordium alloy. A ten-color glass filter disc is arranged in the light source device, the surface of the filter disc is covered with a meta-bromic acid pentaniobate composite filter film with the thickness being 20 microns, and a near ultraviolet optical filter and a far infrared optical filter are arranged in the light source device. A rhodium and osmium nitrate hexahydrate nano composite filter film is arranged in the near ultraviolet optical filter, a high energy light-emitting electrode is arranged in the light source device, and the electrode is made of americium and berkelium oxide alloy. An octahedral prism and a double-oblique crystal grating are arranged in the photoseparation device, and the octahedral prism and the double-oblique crystal grating are alternately arrayed in a series connection mode. A parabolic mirror, a hyperbolic mirror and an ellipsoidal lens are arranged in the primary reflecting device.

The invention discloses a solution reactor used for the plutonium combustion and the neptunium-237 or the americium-241 transmutation. The solution reactor adopts plutonium or plutonium and americium in the spent fuel of the nuclear power plant as the fuel, and adopts neptunium-237 or americium-241 as the target. By adopting the solution reactor, the minor actinides which have long half life and high toxicity and can not be stored for a long time, such as the neptunium-237 or the americium-241, can be changed into nuclides with short half life and low toxicity through transmutation. Because the solution reactor adopts the spent fuel and can generate electricity at the low temperature, the cost-effectiveness and the security are high; the neutron provided by the plutonium combustion can extract plutonium-238 from the neptunium-237 solution target as the isotope battery, thereby the inexpensive energy is provided; the americium-242m extracted from the americium-241 solution target can be widely applied to the aerospace industry and the medical industry.

The invention discloses a nuclear reactor long-life neutron source device and a using method, the device comprises a neutron source assembly capable of installing a long-life neutron source in a sealing mode, and the neutron source assembly is connected with a reactor control rod driving mechanism. The neutron source assembly is inserted into or pulled out of a reactor core through the reactor control rod driving mechanism; the long-life neutron source can be an americium-beryllium source or a plutonium-beryllium source. According to the invention, the consumption of the neutron source material during the operation of the reactor can be obviously reduced, and the long-life characteristic of the neutron source material is maintained, so that the neutron source material can be repeatedly used in one or more units, and the uncertainty of the project cost and project progress of primary neutron source material purchase of each newly built nuclear power plant unit in China is avoided.

The novel lipophilic metal extractants of the class dialkyldiaza-tetraalkyloctanediamide (DADA) useful to selectively separate trivalent americium (sup. 241 Am) from trivalent lanthanides are represented by the formula 1:Wherein R is a C1 to C5 normal alkyl and R′ is a C4 to C8 normal and branched alkyl group. The compounds are synthesized at high yield and purity by the reaction of corresponding N,N′-dialkylethylenediamine and N,N-dialkyl-2-chloroacetamide. The separation is achieved by utilizing the soft-soft interaction between the trivalent actinides and ‘N’ atoms of the extractant. Both soft donor “n” and hard donor ‘O’ sites are incorporated in the molecule for better extraction of trivalent actinides over trivalent lanthanides. Thus, this molecule can be used as selective extractant to separate trivalent actinides from trivalent lanthanides.

The invention discloses a luminous pill of a metal halide white lamp, which contains sodium iodide, thallium iodide, indium iodide, iridium iodide and americium iodide. Compared with the prior art, the prepared metal halide white lamp is characterized by good color rendering, the color rendering index being larger than 80, and good photochromic consistency.

The novel lipophilic metal extractants of the class dialkyldiaza-tetraalkyloctanediamide (DADA) useful to selectively separate trivalent americium (sup. 241 Am) from trivalent lanthanides are represented by the formula 1:Wherein R is a C1 to C5 normal alkyl and R′ is a C4 to C8 normal and branched alkyl group. The compounds are synthesized at high yield and purity by the reaction of corresponding N,N′-dialkylethylenediamine and N,N-dialkyl-2-chloroacetamide. The separation is achieved by utilizing the soft-soft interaction between the trivalent actinides and ‘N’ atoms of the extractant. Both soft donor “n” and hard donor ‘O’ sites are incorporated in the molecule for better extraction of trivalent actinides over trivalent lanthanides. Thus, this molecule can be used as selective extractant to separate trivalent actinides from trivalent lanthanides.

The proposed invention relates to processes of extraction and concentration of radio nuclides and can be used in radiochemical technologies when processing liquid radioactive wastes.A method for extraction of americium from liquid radioactive wastes and its separation from rare-earth elements comprises simultaneous extraction of americium and rare-earth elements from radioactive nitrate solution with neutral solution of organic extracting agent in polar fluorinated organic solvent, washing of saturated with metals organic phase, selective re-extraction of americium. N,N,N′,N′-tetraalkyl-amide of diglycolic acid is used as an extracting agent and solution containing 5-20 g / L of complexon, 5-60 g / L of nitrogen-containing organic acid and 60-240 g / L of salting-out agent is used as a solution for re-extraction of americium.Technical effect is the extraction of americium from acidic liquid radioactive solutions and its separation from all rare-earth elements in a single extraction cycle.

A process is for oxidizing americium(III) to americium(VI) includes providing a aqueous acidic composition comprising americium(III) and a mineral acid and exposing the composition to ozone and silver ion under conditions suitable for oxidation of the americium(III) to americium(VI). Nitric, acid is a suitable mineral acid for the process. Extraction of the americium from the silver is possible using organic phosphonate extractant.

A high-sensitivity atomic spectral-line analysis instrument is characterized by mainly comprising a high-voltage discharging light-beam apparatus, a spectrum collection apparatus, a sample vaporization apparatus, a sample introduction apparatus, a light-beam filtering apparatus, a spectral-line separating apparatus, a spectrum development apparatus, a spectra-line conversion apparatus and a spectra-line display and recording apparatus; the high-voltage discharging light-beam apparatus comprises a high-voltage discharging light-beam electrode, and the electrode material is an americium einsteinium oxide alloy material; the spectrum collection apparatus comprises an nonagonal-pyramid-type focusing reflection lens, and the surface of the lens is plated with an ammonium magnesium octachromate-berkelium niobium bromide nanometer composite light-transmitting film with the thickness of 3.8 mu m; the sample vaporization apparatus comprises a high-temperature vaporization electrode, and the electrode material is a technetium oxide material; and the spectra-line separation apparatus comprises an nonagonal-prism type spectrum separation lens, the surface of the lens is plated with a spectra-line filtering membrane with the thickness of 10 mu m, and the filtering membrane is a neodymium ytterbium tetradecaphosphate composite membrane.

The invention discloses a destructible americium-beryllium neutron source and belongs to the technical field of nuclear application. According to the invention, a device structure is designed so thatwhen an americium-beryllium neutron source is needed to be destroyed, a source core of a radioactive source is soaked by a strong alkali solution, and metal beryllium in the source core is dissolved;then AmO2 in the source core is separated from the metal beryllium by using a filter screen, so that the alpha particles emitted from 241Am are prevented from bombarding the metal beryllium, a 9Be(alpha,n)12C reaction is stopped, and then the production of neutrons is stopped. According to the invention, a scrapped americium-beryllium neutron source can be conveniently destroyed, so that neutronsare no longer ejected, the harm to human bodies is reduced, and the treatment cost of the discarded americium-beryllium neutron source is reduced; meanwhile, the remaining AmO2 can be used as a low-energy gamma source, so that the repeated use of radioactive waste materials is realized.