371results about "Specific isotope recovery" patented technology

The invention provides a solid target system for simultaneously producing Ge-68 and PET nuclides under a high-power beam, and the system comprises an intermediate-energy cyclotron which is used for transmitting an intermediate-energy high-intensity proton beam; the cascade target is arranged beside the intermediate-energy cyclotron and comprises a shell, and a first aluminum target body, a second aluminum target body, a Ga4Ni alloy capsule target and a metal sealing target which are sequentially arranged along the emission direction of the intermediate-energy high-current proton beam, and the shell is filled with cooling water; the target film is erected between the intermediate energy cyclotron and the first aluminum target body; the helium cooling unit is erected between the target film and the first aluminum target body; and the control unit is used for controlling to start the helium cooling unit and the medium-energy cyclotron, so that the medium-energy high-intensity proton beam sequentially passes through the target film, the first aluminum target body, the shell and the cooling water, then enters the Ga4Ni alloy capsule target to produce Ge-68 nuclide, and then enters the metal sealing target to produce PET nuclide. The method has the beneficial effects that the melting of the target material can be reduced, the probability of direct chemical reaction of molten Ga and Nb is reduced, and the types of nuclide production are increased.

A molten salt reactor system may produce fission products, such as molybdenum-99, which may be extracted by an extraction system. The extraction system may be configured to sparge the molten fuel salt with an inert gas to dislodge gaseous fission products from the molten fuel salt, so that the gaseous fission products flow through an off-gas outlet into a gas transfer assembly. The example extraction system may also include introducing a gaseous halogenating agent to prevent the fission products from depositing on a pipe surface of the gas transfer assembly. The gaseous fission products may flow from the gas transfer assembly to a carboy containing an aqueous solution, to process and / or filter the fission products. The resulting solution containing processed fission products in the carboy may be further distilled into final, usable products.

The present invention relates to assemblies and method for obtaining a container comprising 212Pb on the walls obtained from a 212Pb precursor isotope source. The invention provides an improved system and method for producing 212Pb in high purity without the need for processing, with high yields, and which safely and efficiently can be transported to the locations where it is to be used.

The present invention is directed towards methods for isolating molybdenum compounds from a mixture of fission products. The mixture of fission products may be extracted from a molten salt reactor system. Utilizing a phase transfer agent, the molybdenum compounds may be extracted from an aqueous solution into an organic solution, thereby isolating the molybdenum compound from the mixture of fission products. Molybdate may then be isolated from the resulting organic solution and provided to a generator to facilitate transformation into technitum-99m.

The present invention relates to assemblies and method for obtaining a container comprising 212Pb on the walls obtained from a 212Pb precursor isotope source. The invention provides an improved system and method for producing 212Pb in high purity without the need for processing, with high yields, and which safely and efficiently can be transported to the locations where it is to be used.

A mixed-spectrum molten-salt reactor has a primary containment vessel. A reactor core is housed in the primary containment vessel, the reactor core has an atypical geometry that holds a fuel salt. A blanket region is formed within the primary containment vessel and surrounding the reactor core, the blanket region comprises a blanket salt.

The present invention discloses a cyclotron metal composite target for preparing targeted radionuclide drugs, which comprises an aluminum substrate whose shape matches the shape of the target position of the accelerator. A bismuth layer is attached to or embedded in the surface of the aluminum substrate, and the thickness of the bismuth layer is 10-20 μm. The present invention provides a cyclotron metal composite target for preparing targeted radionuclide drugs, which can reduce or even avoid the melting of the target, improve the heat conduction efficiency of the target, and avoid the pollution of the equipment and its products.

Disclosed herein are methods and systems for the use of extraction chromatography resins to concentrate Ac from dilute solutions that may contain Th isotopes and / or other isotopes, such as Ra. As disclosed herein, at least some extraction chromatography resins strongly retain trivalent actinides between 4 and 6 M HNO3 that can be recovered in good yields with small volumes of dilute acids. These small volumes of acid eluents result in significant concentration of the solution of the trivalent / certain tetravalent ions, such that they can be further processed without requiring lengthy evaporation steps that are utilized in standard Ac concentration methods.

A method that includes accelerating ions toward a lattice of carbon fibers and capturing the ions in the lattice of carbon fibers.

Systems and methods for generating radionuclides, such as radium-224, are disclosed herein. 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, 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, and a fourth cartridge having a first opening, a second opening, and a chamber therebetween having a fourth 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.

A radioisotope target station, including a target housing. The target housing includes a target material, and a target backing material. The radioisotope target station includes a loader mechanism comprising at least one jaw configured to secure the target housing, a loading position comprising a magazine, wherein the magazine is configured to hold two or more target housings, an irradiation position comprising a beam, where the beam is configured to irradiate the target housing orthogonally, a cooling fluid source fluidly coupled to the loader mechanism, where the cooling fluid source is configured to cool the target housing as the target housing is irradiated, and an ejection position comprising an ejector and an ejection chute, where the ejector is configured to detach the target housing from the loader mechanism and the ejection chute is configured to receive the target housing when detached.

A target for heavy ion or electron bombardment is contemplated as is a method of its preparation. In that method, a) liquid target solution of a target metal salt dissolved in water is deposited into the bottom of a metal target capsule having an open top, and b) heated within the target capsule to form a crystalline material. Steps a) and b) are repeated until the crystalline material contains a target amount of target salt. The target capsule is then heated until the weight of the target capsule remains constant. To mitigate the formation of salt creep during the preparation, the liquid target solution can contain about 2 to about 20 percent v / v of a C2-C4 polyol, or the inside surface of the target capsule is coated with a hydrophobic film that provides a contact angle with water of about 70 to about 1300 prior to liquid deposition.

A converterless method for converting 226Ra to 225Ac by direct electron beam bombardment with high energy electron beam having an average effective beam power of about 20 to about 250 kW at a beam energy of about 25 to about 100 MeV to impact the 226Ra is disclosed. Contrary to other electron-based bombardment methods, the present method is carried out without a converter that slows the electrons and transforms their kinetic energy into Bremsstrahlung photons, but rather into the production of virtual photons that transmute the target 226Ra to 225Ra that decays to 225Ac.

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 present application relates to a kind of methods for separating actinium from irradiated thorium target, belong to medical radioisotope manufacturing technical field, solve the problem that thorium element and lanthanide element impurity content in final actinium product in prior art is higher.The present application discloses a kind of methods for separating actinium from irradiated thorium target, first, irradiated thorium target is dissolved, then using cation exchange resin column, UTEVA or TK200 resin extraction column, DGA resin extraction column, LN resin extraction column four-stage series separation purification process, obtain final actinium product.Actinium is efficiently and selectively recovered, and the thorium element and lanthanide element impurity content in final actinium product is significantly reduced.

Techniques are described for producing actinium-225 by arranging a converter material at least partially around a target material, where the target material includes radium-226. The converter material comprises a compound of lithium and hydrogen at least partially enriched with lithium-6 and deuterium, and is configured to convert thermal neutrons to fast neutrons, which are then incident on the radium-226. If the fast neutrons have a kinetic energy of at least 6.4 MeV, they will initiate a (n,2n) reaction that converts radium-226 into radium-225, which then decays into actinium-225. The process can desirably be performed within an environment rich in thermal neutrons, such as within a fission reactor.

An automatic doctor blade (1) for scraping a radioactive target (101) to be used for a radiopharmaceutical, the automatic doctor blade comprising: a target holder (2) for holding a radioactive target; a scraping unit (3) comprising a scraping member for scraping the radioactive target located in the target holder; a receiver (5) for receiving material scraped from the radioactive target; a motor (6) for moving the scraping unit and the target holder relative to each other; and a control unit for controlling the motor. A surface (102) of the radioactive target to be scraped is arranged at an angle between 45 degrees and 315 degrees, the angle of 0 degrees corresponding to the radioactive target being oriented such that the surface of the radioactive target to be scraped is perpendicular to the vertical direction and facing upward, the angle of the target corresponding to a clockwise rotation of the target from 0 degrees.

Compositions comprising high levels of high specific activity copper-64, and process for preparing said compositions. The compositions comprise from about 2 Ci to about 15 Ci of copper-64 and have specific activities up to about 3800 mCi copper-64 per microgram of copper. The processes for preparing said compositions comprise bombarding a nickel-64 target with a low energy, high current proton beam, and purifying the copper-64 from other metals by a process comprising ion exchange chromatography or a process comprising a combination of extraction chromatography and ion exchange chromatography.

A target for heavy ion or electron bombardment is contemplated as is a method of its preparation. In that method, a ) liquid target solution of a target metal salt dissolved in water is deposited into the bottom of a metal target capsule having an open top, and b ) heated within the target capsule to form a crystalline material. Steps a ) and b) are repeated until the crystalline material contains a target amount of target salt. The target capsule is then heated until the weight of the target capsule remains constant. To mitigate the formation of salt creep during the preparation, the liquid target solution can contain about 2 to about 20 percent v / v of a C2-C4 polyol, or the inside surface of the target capsule is coated with a hydrophobic film that provides a contact angle with water of about 70 to about 130° prior to liquid deposition.

A method is provided of continuous manufacturing of Actinium-225 from a radium generator by repeating chemical purification is provided.

The present application relates to radionuclide separation and purification technical field, specifically disclose a kind of separation and purification method for preparing lead-212 and actinium-225 by photo nuclear reaction method.It includes the following steps: radium target solution is sequentially passed through lead resin column and DGA resin column, and recovery liquid is obtained;First eluent is used to elute lead resin column, and lead-212 first purification liquid is obtained;Second eluent is used to elute DGA resin column, and actinium-225 first purification liquid is obtained;Lead-212 first purification liquid is introduced into first LN resin column, and third eluent is used to elute first LN resin column, then third desorption liquid is introduced into first LN resin column, and lead-212 second purification liquid is obtained;Actinium-225 first purification liquid is introduced into second LN resin column, and fourth eluent is used to elute second LN resin column, then fourth desorption liquid is introduced into second LN resin column, and actinium-225 second purification product solution is obtained.The present application effectively avoids the cumbersome operation steps such as medium conversion, improves the separation and purification efficiency of lead-212 and actinium-225 in radium target.

A radionuclide separating system for separating a 213Bi daughter radionuclide from a 225Ac parent radionuclide, the radionuclide separating system comprising: an inlet for loading a liquid solution comprising the 225Ac parent radionuclide onto a column; the column comprising a sorbent material wherein the sorbent material is capable of interacting with the 225Ac parent radionuclide and 213Bi daughter radionuclide so as to allow selective desorption of the 225Ac parent radionuclide and / or the 213Bi daughter radionuclide at a different moment in time; and an outlet for selectively obtaining the 213Bi daughter radionuclide based on the selective desorption of the 225Ac parent radionuclide and the 213Bi daughter radionuclide, wherein the sorbent material is a carbon-based sorbent material.

To extract technetium-99m from low-specific-activity molybdenum-99 in a short time without being influenced by the volume of molybdenum solution.The method for recovering technetium-99m from low-specific-activity molybdenum-99 is a method in which the daughter nuclide technetium-99m produced through the decay of molybdenum-99 with a low specific activity contained in a high-concentration molybdenum solution containing the molybdenum-99 with a low specific activity is separated by using activated charcoal and recovered, and the method is characterized in that the activated charcoal is immersed into the molybdenum solution and selectively adsorbs a trace amount of technetium-99m in the solution, even when molybdenum-99 is present 1016 times or more abundantly in terms of the ratio of the number of atoms.

The present invention relates to assemblies and method for obtaining a container comprising 212Pb on the walls obtained from a 212Pb precursor isotope source. The invention provides an improved system and method for producing 212Pb in high purity without the need for processing, with high yields, and which safely and efficiently can be transported to the locations where it is to be used.

A method that includes accelerating ions toward a lattice of carbon fibers and capturing the ions in the lattice of carbon fibers.

The present invention provides a radionuclide production system and a radionuclide production method with which it is possible to produce radionuclides efficiently by improving the cooling of a target material containing a raw material nuclide and reducing contamination of the target material with impurities. A radionuclide production system comprises a particle beam irradiation device that generates a particle beam, and a target that generates radionuclides through irradiation with the particle beam. In the target, a target material containing a raw material nuclide that generates radionuclides is mounted on a target cooling plate for cooling the target material. The radionuclide production method comprises: forming a target in which a target material containing a raw material nuclide is mounted on a target cooling plate for cooling the target material; and producing radionuclides by irradiating the target with a particle beam or bremsstrahlung radiation while cooling the target material using the target cooling plate.

An irradiation target for the production of radioisotopes comprises a plurality of plates having aligned central openings. First and second elongated central members extend in opposite directions through the aligned central openings of the plates. The first and second elongated central members each have a pair of wings at first and second ends of the plurality of plates, respectively. The first and second elongated central members further have a pair of tabs adjacent to the wings of the other elongated central member. Tabs of the first elongated central member are bent to engage one end face of the plurality of plates. Tabs of the second elongated central member are folded over the wings of the first elongated central member. The plates and the first and second elongated central members are formed of materials that produce molybdenum-99 (Mo-99) by way of neutron capture.

The invention discloses a method for preparing the 68 Ge method. The method comprises the following steps: S1, preparing a metal gallium nickel alloy target, and then dissolving it after irradiation; the structure of the metal gallium nickel alloy target is as follows 1) or 2): 1) at least two layers, including a metal base layer and a metal gallium nickel alloy layer; 2) at least three layers, including a metal base layer, a metal protective layer and a metal gallium nickel alloy layer; S2, passing the dissolved solution into a double series chromatographic column for separation and purification, and collecting the 68 Ge solution; S3, the obtained 68 The Ge solution was passed into 68 Ge / 68 In the Ga generator, when 68 After Ge is fully adsorbed, the eluent is used to wash the Ge at preset intervals. 68 Ge / 68 The separation column in the Ga generator is eluted, and the eluted eluent is collected to obtain 68 Ga products. The present invention 68 The separation and purification process of Ge has clear ideas, simple operation, and is easy to realize automated separation and purification, and can obtain high purity and radioactive concentration. 68 Ge.

225 Actinium is converted into liquid by proton, deuterium or gamma irradiation in the irradiation device (2). 226 The radium target is irradiated and the resulting liquid target solution in the first extractor (6) is extracted. 225 By extracting actinium, 226 It is produced from radium. 225 The liquid target solution from which actinium has been removed is irradiated again to generate more 225 The liquid target solution is preferably circulated through the irradiation device in a closed loop (4) and through the first extraction device (6) in a further closed loop (7). The advantage of such a method is that the irradiated target solution does not have to be dried and redissolved so that the produced actinium can be separated from the radium, and no further drying and redissolution steps are required to regenerate the liquid target starting from the separated radium. The radium target is therefore in particular 226 Taking into account the radon gas that is continuously produced by the decay of radium, it can be recycled in a more efficient and safe way.

A method and system (300) for separating thorium from uranium and its decay products is provided. The method includes combining a nuclear fuel feedstock (102) containing thorium and uranium with a first acid (104, 204) to form a first solution. The first solution is contacted with an ion exchange resin (308) that is selective for thorium or uranium. The thorium or uranium is at least partially removed from the first solution by binding the thorium or uranium to the ion exchange resin (308), thereby forming a second solution (106, 206). The second solution (106, 206) is combined with oxalic acid to precipitate the uranium or thorium from the second solution, forming a precipitate (110, 210). The precipitate (110, 210) is separated from the second solution (106, 206).