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Небесная энциклопедия

Космические корабли и станции, автоматические КА и методы их проектирования, бортовые комплексы управления, системы и средства жизнеобеспечения, особенности технологии производства ракетно-космических систем

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Мониторинг СМИ

Мониторинг СМИ и социальных сетей. Сканирование интернета, новостных сайтов, специализированных контентных площадок на базе мессенджеров. Гибкие настройки фильтров и первоначальных источников.

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Поддерживает ввод нескольких поисковых фраз (по одной на строку). При поиске обеспечивает поддержку морфологии русского и английского языка
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Применить Всего найдено 1338. Отображено 100.
02-02-2012 дата публикации

Isotope production target

Номер: US20120027152A1
Автор: Madicken Munk, Stephen Todd Keller, Steven Richard Reese, Todd Stephen Palmer
Принадлежит: Oregon State, Oregon State Board of Higher Education

An isotope production target may include an outer diameter wall and an inner diameter wall. An isotope source may be located between the inner diameter wall and the outer diameter wall, and the isotope source may comprise fissile material interspersed with one or more voided regions. A central region may be located within the inner diameter wall, and the central region may be configured to house a neutron thermalization volume.

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Номер записи: 1
26-04-2012 дата публикации

Electrochemical cell and method for separating carrier-free 18f-from a solution on an electrode

Номер: US20120097549A1
Автор: Kurt Hamacher
Принадлежит: FORSCHUNGSZENTRUM JUELICH GMBH

Disclosed is an electrochemical cell and a method for separating carrier-free radionuclides from a solution on an electrode. 18 F − is precipitated in an electrochemical cell from an aqueous solution on an anode, which is diamond-coated. Subsequently, the electrochemical cell is dried and supplied with a liquid containing a transfer catalyst, the anode is preferably switched to serve as the cathode, and 18 F − is transferred to the liquid phase.

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Номер записи: 2
14-06-2012 дата публикации

Electrochemical phase transfer devices and methods

Номер: US20120145557A1
Автор: Christian Rensch, Christoph Boeld, Marko Baller, Victor Samper
Принадлежит: General Electric Co

Devices and methods for electrochemical phase transfer utilize at least one electrode formed from either glassy carbon or a carbon and polymer composite. The device includes a device housing defining an inlet port ( 42 ), an outlet port ( 44 ) and an elongate fluid passageway ( 36 ) extending therebetween. A capture electrode ( 12 ) and a counter electrode are positioned within said housing such that the fluid passageway extends between the capture and counter electrodes.

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Номер записи: 3
27-09-2012 дата публикации

Adsorbents for Radioisotopes, Preparation Method Thereof, and Radioisotope Generators Using the Same

Номер: US20120244055A1
Автор: Hyeon Young Shin, Hyon Soo Han, Jong Sub Lee, Jun Sig LEE, Kang Duk JANG, Kwang Jae Son, Soon Bog HONG, Ul Jae Park
Принадлежит: Korea Atomic Energy Research Institute KAERI

Disclosed is a novel adsorbent for use in a 99 Mo/ 99m Tc generator, which is a medical diagnostic radioisotope generator, and in a 188 W/ 188 Re generator, which is a therapeutic radioisotope generator. The adsorbent composed of sulfated alumina or alumina-sulfated zirconia exhibits adsorption capacity superior to that of conventional adsorbents, and is stable and is thus loaded in a dry state in an adsorption column so that the radioisotope 99 Mo or 188 W can be adsorbed. Thus, it is possible to miniaturize the column, and such a miniaturized column is small, convenient to use, and highly efficient, and extracts a radioisotope satisfying the requirements for pharmaceuticals, and thus can be useful for radioisotope generators extracting 99m Tc or 188 Re.

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Номер записи: 4
04-07-2013 дата публикации

Purification of Metals

Номер: US20130171046A1
Автор: Luis Antonio Miguel Marques Barbosa
Принадлежит: Mallinckrodt LLC

A solid composition comprises: MnO 2 ; and a compound represented by the general formula (I) wherein: R is a polymer; each Y is independently a hydrogen or a negative charge; Z is either hydrogen or is not present; each n is independently 1, 2, 3, 4, 5 or 6; wherein the MnO 2 is bound to the compound of formula (I) so as to coat the surface thereof. Such a composition may be used for the separation of polyvalent metal species, such as Mo, from one or more accompanying impurities.

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Номер записи: 5
01-01-2015 дата публикации

PROCESS OF GENERATING GERMANIUM

Номер: US20150003576A1
Автор: Perrigo Melissa Dianne, Pipes David Wayne, Uhland William C.
Принадлежит:

The present disclosure generally relates to a new process for generating germanium-68 from an irradiated target body. The process includes irradiation of the target body followed by various extraction techniques to generate the germanium-68. 1. A process for generating a radioisotope , the process comprising:bombarding a target body including a starting material, wherein the bombardment of the starting material produces a radioisotope within the target body;allowing the bombarded target body to decay;stripping the bombarded target body with an acidic mixture to create a stripped solution;extracting the radioisotope from the stripped solution using a non-polar solvent to remove the acidic mixture and create a non-polar solvent fraction including the radioisotope;washing the non-polar solvent fraction including the radioisotope; and,extracting the radioisotope from the non-polar solvent fraction using water.2. The process of claim 1 , wherein the radioisotope is germanium-68.3. The process of claim 1 , wherein the starting material is an alloy comprising gallium.4. The process of claim 3 , wherein the alloy includes a metal selected from the group consisting of nickel claim 3 , indium claim 3 , tin claim 3 , iron claim 3 , ruthenium claim 3 , osmium claim 3 , chromium claim 3 , rhenium claim 3 , molybdenum claim 3 , tungsten claim 3 , manganese claim 3 , cobalt claim 3 , rhodium and combinations thereof.5. The process of claim 3 , wherein the alloy includes from about 10% to about 80% gallium claim 3 , by weight of the alloy.6. The process of claim 3 , wherein the alloy includes gallium and nickel.7. The process of claim 6 , wherein the alloy includes from about 60% to about 75% gallium and from about 25% to about 40% nickel claim 6 , by weight of the alloy.8. The process of claim 1 , wherein the acidic mixture includes hydrochloric acid and nitric acid.9. The process of claim 1 , wherein the acidic mixture includes copper (II) nitrate trihydrate and nitric acid.10. ...

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Номер записи: 6
10-01-2019 дата публикации

Process for producing Gallium-68 through the irradiation of a solution target

Номер: US20190013108A1
Автор: ABRUNHOSA Antero, ALVES Francisco, ALVES Vítor
Принадлежит:

The present disclosure relates to a process for purifying and concentrating Ga isotope produced by the irradiation with an accelerated particle beam of a Zn target in solution. The process according to the present disclosure allows for the production of pure and concentrated Ga isotope in hydrochloric acid solution. The present disclosure also relates to a disposable cassette for performing the steps of purification and concentration of the process. 1. A process for producing and purifying Gallium radioisotope , the process comprising:irradiating a target containing a target solution comprising zinc using an accelerated particle beam;diluting the irradiated target solution with water;feeding the diluted target solution into a strong cation exchanger;washing the strong cation exchanger;eluting zinc isotopes from the strong cation exchanger with a zinc elution solution including acetone;washing the strong cation exchanger;{'sup': '68', 'eluting Gallium isotope from the strong cation exchanger with hydrochloric acid solution to obtain an eluted solution;'}feeding the eluted solution into a strong anion exchanger,washing the strong anion exchanger; and{'sup': '68', 'eluting Gallium isotope from the strong anion exchanger with hydrochloric acid solution to obtain a final solution'}wherein the irradiated target solution is diluted at least 5 volume times with water.2. The process according to claim 1 , wherein the irradiated target solution is diluted at least 10 volume times with water.3. The process according to claim 1 , further comprising:complementing the eluted solution with another hydrochloric acid solution to obtain a complemented solution, wherein the complementing is performed before feeding said the eluted solution into the strong anion exchanger.4. The process according to claim 3 , wherein the complemented solution includes a molarity in hydrochloric acid between 7 M and 10 M.5. The process according to claim 1 , wherein the accelerated particle beam is a ...

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Номер записи: 7
16-01-2020 дата публикации

SHIELDING ASSEMBLY FOR A RADIOISOTOPE DELIVERY SYSTEM HAVING MULTIPLE RADIATION DETECTORS

Номер: US20200016284A1
Автор: CHERUNDOLO Brian, DENNIS Joseph, MOYERS Clif, NUNN Adrian, SCHIMMOELLER Andrew
Принадлежит:

A shielding assembly may be used in a nuclear medicine infusion system that generates and infuse radioactive liquid into a patient undergoing a diagnostic imaging procedure. In some examples, the shielding assembly has multiple compartments each formed of a shielding material providing a barrier to radioactive radiation. For example, the shielding assembly may have a first compartment configured to receive a radioisotope generator that generates a radioactive eluate via elution, a second compartment configured to receive a beta detector, and a third compartment configured to receive a gamma detector. In some examples, the compartments are arranged to minimize background radiation emitted by the radioisotope generator and detected by the gamma detector to enhance the quality of the measurements made by the gamma detector. 1. A system comprising: a first compartment configured to receive a radioisotope generator that generates a radioactive eluate via elution;', 'a second compartment configured to receive a beta detector, and', 'a third compartment configured to receive a gamma detector., 'a shielding assembly that has a plurality of compartments each formed of a shielding material providing a barrier to radioactive radiation, comprising2. The system of claim 1 , wherein the third compartment is configured to receive an eluate-receiving container such that both the gamma detector and the eluate-receiving container can be positioned in the third compartment.3. The system of claim 2 , wherein the third compartment comprises a sidewall defining an opening through which the eluate-receiving container can be inserted.4. The system of claim 3 , wherein the gamma detector is positioned to detect gamma emissions emitted by a static portion of the radioactive eluate received by the eluate-receiving container.5. The system of claim 3 , further comprising a removable insert positioned in the opening claim 3 , wherein the removable insert defines a cavity configured to receive ...

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Номер записи: 8
21-01-2016 дата публикации

Apparatus and Method for Stripping Tritium from Molten Salt

Номер: US20160019993A1
Автор: Holcomb David E., Wilson Dane F.
Принадлежит:

A method of stripping tritium from flowing stream of molten salt includes providing a tritium-separating membrane structure having a porous support, a nanoporous structural metal-ion diffusion barrier layer, and a gas-tight, nonporous palladium-bearing separative layer, directing the flowing stream of molten salt into contact with the palladium-bearing layer so that tritium contained within the molten salt is transported through the tritium-separating membrane structure, and contacting a sweep gas with the porous support for collecting the tritium.

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Номер записи: 9
03-02-2022 дата публикации

Method for producing lead-212 from an aqueous solution comprising thorium-228 and daughters thereof

Номер: US20220037046A1
Автор: Julien Torgue, Remy Dureau
Принадлежит: Orano Med SAS

A method for producing lead-212 of very high radiological purity from an aqueous solution comprising thorium-228 and daughters thereof. Manufacture of radiopharmaceuticals based on lead-212, which are useful in nuclear medicine and, in particular, in targeted alpha radiation therapy for the treatment of cancers.

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Номер записи: 10
18-01-2018 дата публикации

Production of n-13 ammonia radionuclide

Номер: US20180019034A1
Автор: Francis Y. Tsang
Принадлежит: GLOBAL MEDICAL ISOTOPE SYSTEMS LLC

A method of producing 13 N-ammonia for use in medical imaging is provided, which includes irradiating 14 N (having a natural abundance of 99.64%) with a collimated bremsstrahlung radiation (gamma-ray beam) obtained by directing high-energy electrons onto a high-Z converter. The 14 N to be irradiated may be in the form of liquid ammonia ( 14 NH 3 ) or ammonia gas to directly produce 13 N-ammonia ( 13 NH 3 ) or in the form of liquid nitrogen to indirectly produce 13 N-ammonia through conversion of the irradiated liquid nitrogen (N 2 ) via known conversion processes to 13 N-ammonia. The photons have an energy level above the threshold of the 14 N(γ,n) 13 N reaction (about 10.5 MeV).

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Номер записи: 11
17-01-2019 дата публикации

PROCESS OF GENERATING GERMANIUM

Номер: US20190019591A1
Автор: Perrigo Melissa Dianne, Pipes David Wayne, Uhland William C.
Принадлежит:

The present disclosure generally relates to a new process for generating germanium-68 from an irradiated target body. The process includes irradiation of the target body followed by various extraction techniques to generate the germanium-68. 1. A process for generating a radioisotope , the process comprising:bombarding a target body including a starting material, wherein the bombardment of the starting material produces a radioisotope within the target body;{'sub': '3', 'allowing the bombarded target body to decay; stripping the bombarded target body with an acidic mixture to create a stripped solution, wherein the acidic mixture includes (a) copper (II) nitrate trihydrate and nitric acid, or (b) 3 M to 6 M hydrochloric acid (HCl) and 6 M to 15 M nitric acid (HNO);'}extracting the radioisotope from the stripped solution using a non-polar solvent to remove the acidic mixture and create a non-polar solvent fraction including the radioisotope;washing the non-polar solvent fraction including the radioisotope; and,extracting the radioisotope from the non-polar solvent fraction using water.2. The process of claim 1 , wherein the radioisotope is germanium-68.3. The process of claim 1 , wherein the starting material is an alloy comprising gallium.4. The process of claim 3 , wherein the alloy includes a metal selected from the group consisting of nickel claim 3 , indium claim 3 , tin claim 3 , iron claim 3 , ruthenium claim 3 , osmium claim 3 , chromium claim 3 , rhenium claim 3 , molybdenum claim 3 , tungsten claim 3 , manganese claim 3 , cobalt claim 3 , rhodium and combinations thereof.5. The process of claim 3 , wherein the alloy includes from about 10% to about 80% gallium claim 3 , by weight of the alloy.6. The process of claim 3 , wherein the alloy includes gallium and nickel.7. The process of claim 6 , wherein the alloy includes from about 60% to about 75% gallium and from about 25% to about 40% nickel claim 6 , by weight of the alloy.8. The process of claim 1 , ...

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Номер записи: 12
16-01-2020 дата публикации

Solution Target for Cyclotron Production of Radiometals

Номер: US20200020457A1
Автор: Byrne John, DeGrado Timothy R., Pandey Mukesh K.
Принадлежит:

Methods of producing and isolating Ga, Zr, Cu, Zn, Y, Cu, Tc, Ti, N, Mn, or Sc and solution targets for use in the methods are disclosed. The methods of producing Ga, Zr, Cu, Zn, Y, Cu, Tc, Ti, N, Mn, or Sc include irradiating a closed target system with a proton beam. The system can include a solution target. The methods of producing isolated Ga, Zr, Cu, Zn, Y, Cu, Tc, Ti, Mn, or Sc further include isolating Ga, Zr, Cu, Zn, Y, Cu, Tc, Ti, Mn, or Sc by ion exchange chromatography. An example target includes a target body including a target cavity for receiving the target material; a housing defining a passageway for directing a particle beam at the target cavity; a target window for covering an opening of the target cavity; and a coolant gas flow path disposed in the passageway upstream of the target window. 1. A method for synthesizing a resin for trapping an analyte , the method comprising:(a) providing a cationic exchange resin having carboxylate groups;(b) activating carboxylate groups of the cationic exchange resin with an activating agent; and(c) reacting activated carboxylate groups of the cationic exchange resin with a hydroxylamine salt in the presence of a base to produce the resin.2. The method of wherein:the activating agent comprises an alkyl ester of chloroformic acid.3. The method of wherein:the activating agent comprises methyl chloroformate.4. The method of wherein:the base comprises an amine.5. The method of wherein:the base is a tertiary amine.6. The method of wherein:the base is triethylamine.7. The method of wherein:the hydroxylamine salt is hydroxylamine hydrochloride.8. The method of wherein:step (b) is undertaken in the presence of a solvent for the activating agent.9. The method of wherein:the solvent comprises a halogenated alkane.10. The method of wherein:the solvent is dichloromethane.11. The method of further comprising:removing the solvent under vacuum.12. The method of wherein:the analyte is a radionuclide.13. The method of wherein:{' ...

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Номер записи: 13
24-01-2019 дата публикации

CONVEYANCE SYSTEM FOR OPERATION IN RADIOACTIVE ENVIRONMENT

Номер: US20190023494A1
Автор: "DHooge Michael J.", Graves Kevin B., Petrofsky Bryan S., Verma Sumit
Принадлежит:

A system for manufacturing radionuclide generators includes an enclosure defining a radioactive environment. The enclosure includes radiation shielding to prevent radiation within the radioactive environment from moving to an exterior of the enclosure. The system also includes a conveyance system having a forward track and first carriages positioned on and movable along the forward track for conveying racks in a first direction. The conveyance system also includes a first walking beam mechanism magnetically coupled to the first carriages to move the first carriages. The conveyance system further includes a return track and second carriages positioned on and movable along the return track for conveying racks in a second direction opposite the first direction. The forward track and the return track form a loop. 1. A conveyance system for operation in an enclosed radioactive environment , the conveyance system comprising:a track;carriages positioned on and moveable along the track for conveying racks along the track;a walking beam mechanism magnetically coupled to the carriages to move the carriages along the track;a lift mechanism for lifting the racks off the carriages, the lift mechanism moveable between an extended position and a retracted position;a first sensor located to track the position of the carriages and racks along the track; anda second sensor located to detect whether the lift mechanism is in the extended position or the retracted position.2. The conveyance system of claim 1 , wherein the first sensor is located below the track and the second sensor is located adjacent the lift mechanism.3. The conveyance system of claim 1 , wherein the second sensor comprises a sensor mechanism including at least one of a mechanical switch and a magnetically-actuated electrical contact.4. The conveyance system of claim 3 , wherein the second sensor comprises a housing defining an interior that contains wiring connected to the sensor mechanism claim 3 , the system ...

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Номер записи: 14
25-01-2018 дата публикации

RADIOISOTOPE RECOVERY

Номер: US20180025801A1
Автор: Alexander Richard, Archibald Stephen James, Brown Nathan Joel, Haswell Stephen John, He Ping, Pamme Nicole, Tarn Mark Duncan
Принадлежит:

The present invention relates to a method and an apparatus for separating and recovering a radioisotope from a solution. More particularly, certain embodiments of the invention relate to a method for recovering a radioisotope from a solution by electro-trapping and release using a microfluidic cell (). The radioisotope may subsequently be used in the preparation of radiopharmaceuticals. 1. A method for separating and recovering a radioisotope from an aqueous solution comprising the radioisotope , the method comprising:using a microfluidic device comprising a chamber;flowing the aqueous solution to the chamber, the chamber comprising a first electrode and a second electrode;generating a first electric field between the first and second electrodes, thereby trapping the radioisotope on the first electrode;flowing an organic-based solution to the chamber comprising the first and the second electrodes; andgenerating a second electric field between the first and the second electrodes;wherein the second electric field has an opposing polarity to the first electric field, thereby releasing the radioactive isotope from the first electrode into the organic-based solution; andwherein the first electrode is formed from a carbon rod or section thereof.2. The method of claim 1 , further comprising one or more of the features selected from:flowing the aqueous solution at a flow rate of at least 0.1 mL/min;flowing the organic-based solution at a flow rate of at least 0.05 mL/min;applying a voltage of no greater than 30 V across the first and second electrodes to generate the first electric field; andapplying a voltage of no greater than 10 V across the first and second electrodes to generate the second electric field.35-. (canceled)6. The method of claim 1 , wherein the chamber has a volume of no greater than approximately 50 μL.7. The method of claim 1 , wherein the first electrode has a flat surface comprising a plurality of recesses and/or the first electrode has a polished ...

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Номер записи: 15
10-02-2022 дата публикации

Processes and systems for producing and/or purifying gallium-68

Номер: US20220044835A1
Автор: Aiman H. ALNAHWI, Brigitte Guerin, Samia Ait-Mohand, Sebastien Tremblay
Принадлежит: SOCPRA Sciences et Genie SEC

The present disclosure relates processes and systems for producing and/or purifying 68Ga from an irradiated substrate of 68Zn. In some embodiments, the process rely on the use two cation-exchange chromatography columns to separate 68Ga from 68Zn and other radionuclides and metallic impurities. The process achieves a high overall yield of 68Ga and a high effective molar activity while being implementable in a time compatible with the short half-life of 68Ga. In additional embodiments, the process is implemented by an automated system.

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Номер записи: 16
10-02-2022 дата публикации

TECHNETIUM 99M ISOLATION SYSTEM AND TECHNETIUM 99M ISOLATION METHOD

Номер: US20220044836A1
Автор: TAKAHASHI Naruto
Принадлежит:

An initial introduction control part introduces an aqueous solution containing molybdenum 99 and technetium 99m, and an organic solvent being capable of dissolving the technetium 99m into an extraction tank. A micro-mixing control part micro-mixes the aqueous solution and the organic solvent by heating and stirring a mixed solution of the aqueous solution and the organic solvent introduced into the extraction tank with a heater, while applying ultrasonic to the mixed solution. A separation control part separates the mixed solution micro-mixed into two phases of aqueous solution and an organic solvent. A taking-out introduction control part passes the organic solvent separated into two phases through an adsorption column be capable of adsorbing molybdenum 99 and introduces the organic solvent into an evaporation elution tank. An evaporation control part evaporates the organic solvent and leaves residue by reducing pressure inside the evaporation elution tank and heating the organic solvent introduced into the evaporation elution tank with a heater, while applying ultrasonic to the organic solvent. An elution control part introduces physiological saline solution into the residue and elutes technetium 99m into the physiological saline solution from the residue. 1. A technetium 99m isolation system comprising:an initial introduction control part introducing an aqueous solution containing molybdenum 99 and technetium 99m, and an organic solvent being capable of dissolving the technetium 99m into an extraction tank;a micro-mixing control part micro-mixing the aqueous solution and the organic solvent by heating and stirring a mixed solution of the aqueous solution and the organic solvent introduced into the extraction tank with a heater, while applying ultrasonic to the mixed solution;a separation control part separating the mixed solution micro-mixed into two phases of aqueous solution and an organic solvent;a taking-out introduction control part passing the organic ...

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Номер записи: 17
23-01-2020 дата публикации

TARGET IRRADIATION SYSTEMS FOR THE PRODUCTION OF RADIOISOTOPES

Номер: US20200027618A1
Автор: Fisher Benjamin D., Onderwater Thomas G.
Принадлежит:

A target irradiation system including an irradiated target removal system having a body defining a central bore, an elevator received within the central bore, and a docking surface for placing the irradiated target removal system in fluid communication with a vessel penetration of a reactor. A target canister slidably receives the radioisotope target therein, and the elevator is configured to receive the target canister. The elevator is lowered into the reactor when irradiating the radioisotope target, and the irradiated target removal system forms a portion of a pressure boundary of the reactor during target irradiation. 1. A target irradiation system for irradiating a radioisotope target in a vessel penetration of a fission reactor , comprising:an irradiated target removal system including a body defining a central bore, an elevator that is configured to be selectively received within the central bore, and a docking surface that is configured to selectively place the irradiated target removal system in fluid communication with the vessel penetration; anda target canister including a body defining a target bore that is configured to slidably receive the radioisotope target therein, and a cap configured to attach to the body of the target canister, thereby providing a water-tight seal for the target bore,wherein the elevator is configured to receive the target canister thereon, the elevator is lowered into the vessel penetration of the reactor when irradiating the radioisotope target, and the irradiated target removal system forms a portion of a pressure boundary of the reactor when in fluid communication with the vessel penetration.2. The target irradiation system of claim 1 , wherein the fission reactor is a heavy-water moderated fission reactor and the vessel penetration is an adjuster port.3. The target irradiation system of claim 1 , wherein the irradiated target removal system further comprises a winch and pulley assembly connected to the elevator by a cable.4 ...

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Номер записи: 18
23-01-2020 дата публикации

Compact Radioisotope Generator

Номер: US20200027619A1
Автор: Buchanan Charles Russell, Hamill James J., Siegel Stefan B.
Принадлежит:

Disclosed are a method and apparatus for making a radioisotope and a composition of matter including the radioisotope. The radioisotope is made by exposing a material to neutrons from a portable neutron source. 1. A composition of matter comprising a radioisotope , the radioisotope made according to a method comprising:obtaining a solution comprising a particular isotope dissolved in the solution;placing the solution into a container;exposing the solution to neutrons from a portable neutron source by completely surrounding the portable neutron source with at least the particular isotope, the particular isotope reacting with the neutrons and transforming into the radioisotope having a short half-life; andextracting the radioisotope from the solution.2. The composition of claim 1 , wherein the radioisotope is dispersed throughout a solid material.3. The composition of claim 2 , wherein the solid material has a geometrical shape.4. The composition of claim 1 , wherein the radioisotope that is made is Cu.5. The composition of claim 1 , wherein the container is located within a medical patient examination facility to expedite use after preparation. This is a divisional of U.S. patent application Ser. No. 12/887,933, filed Sep. 22, 2010, the contents of which are incorporated by reference.This application is directed toward production and use of radioactive isotopes, or radioisotopes.Radioactive isotopes have many beneficial uses. As one example, positron-emitting copper isotopes, such as copper-64 (Cu) and copper-60 (Cu) have a number of uses in clinical and pre-clinical nuclear medicine. These uses include, but are not limited to, the labeling of compounds and the creation of phantom objects suitable for localization and coregistration of multimodality imaging systems, such as those which combine magnetic resonance and positron-emission (MR-PET) imaging. In some instances these radioisotopes are used for oncology imaging and oncological therapy.The production of ...

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Номер записи: 19
23-01-2020 дата публикации

Compact Radioisotope Generator

Номер: US20200027620A1
Автор: Charles Russell Buchanan, James J. Hamill, Stefan B. Siegel
Принадлежит: Siemens Medical Solutions USA Inc

Disclosed are a method and apparatus for making a radioisotope and a composition of matter including the radioisotope. The radioisotope is made by exposing a material to neutrons from a portable neutron source.

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Номер записи: 20
28-01-2021 дата публикации

THE APPARATUS OF PRODUCING NUCLIDE USING FLUID TARGET

Номер: US20210027904A1
Автор: KANG Joo Hyun, Lee Kyo Chul, Lim Sang Moo, OH Se Young
Принадлежит:

The disclosure provide an apparatus for producing a nuclide by using a liquid target which can perform the nuclear reaction process and can discharge the radioactive gas such as Radon within the vial. As described above, an apparatus for producing a nuclide by using a liquid target according to the present disclosure can minimize quantitative loss of a reactant by performing the nuclear reaction process using a target of a liquefied state and reusing a liquefied target on which the nuclear reaction process has not been performed, and can improve safety by enabling the radioactive gas generated to be disposed. 1. An apparatus for producing a nuclide by using a liquid target , the apparatus including:a chamber provided with a reaction space which is configured to accommodate a liquid reactant;a vial which is configured to temporarily accommodate the liquid reactant before a nuclear reaction process and a liquid product after the nuclear reaction process;a syringe pump which is driven to suck a liquid material accommodated in the vial before the nuclear reaction process and enable the sucked liquid material to be supplied to the chamber; andan exhaust unit which is formed such that a radioactive gas within the vial can be discharged.2. The apparatus of claim 1 , wherein the liquid reactant comprises liquefied radium (Ra-226) claim 1 , the liquid product comprises liquefied radium (Ra-226) and liquefied actinium (Ac-225) claim 1 , and the radioactive gas is radon (Rn).3. The apparatus of claim 2 , additionally including a first shielding box which accommodates the vial in an inner space thereof to prevent leakage of the radioactive gas to the outside when the radioactive gas is flown out of the vial.4. The apparatus of claim 3 , additionally including a second shielding box which is configured to accommodate the syringe pump and the first shielding box in an inner space thereof and enable leakage of the radioactive gas to the outside to be prevented.5. The apparatus of ...

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Номер записи: 21
28-01-2021 дата публикации

SYSTEMS, APPARATUS AND METHODS FOR SEPARATING ACTINIUM, RADIUM, AND THORIUM

Номер: US20210027905A1
Автор: ROBERTSON Andrew Kyle Henderson, Schaffer Paul, YANG Hua, Zeisler Stefan
Принадлежит:

A method of separating actinium and/or radium from proton-irradiated thorium metal. The thorium metal is irradiated to produce isotopes including thorium, actinium and/or radium. The resultant product is dissolved in solution and a selective precipitant is used to precipitate a bulk portion of the thorium. The precipitated thorium can be recovered. Chromatography is carried out on the remaining solution to remove residual thorium and to separate the actinium from the radium. 1. A method of separating thorium from actinium and/or radium , the method comprising:placing the thorium and the actinium and/or radium in a weak acid solution;adding a selective precipitant to the weak acid solution and precipitating a bulk portion of the dissolved thorium under precipitation conditions while leaving the actinium and/or radium in the solution; andfiltering to separate the precipitated bulk portion of the thorium from the actinium and/or radium in the solution.2. A method of separating actinium and/or radium from thorium , the method comprising the steps of:placing the thorium and the actinium and/or radium in a weak acid to yield a first solution;adding a selective precipitant and precipitating a bulk portion of the thorium under precipitation conditions while retaining the actinium and/or radium and a residual portion of the thorium in a second solution;filtering to separate the precipitated bulk portion of the thorium from the second solution; andconducting chromatographic purification of the second solution to separate the actinium and/or radium from the residual thorium.3. A method as defined in claim 2 , wherein the thorium and the actinium and/or radium are produced by irradiating thorium metal claim 2 , and wherein the irradiated thorium metal is dissolved in the weak acid to yield the first solution.4. A method of producing thorium radioisotopes claim 2 , the method comprising:irradiating thorium metal to produce thorium radioisotopes;dissolving the irradiated thorium ...

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Номер записи: 22
28-01-2021 дата публикации

METHOD OF PRODUCING ACTINIUM BY LIQUEFIED RADIUM

Номер: US20210027906A1
Автор: KANG Joo Hyun, Lee Kyo Chul, Lim Sang Moo, OH Se Young
Принадлежит:

A method of producing actinium by using liquefied radium can minimize loss of Ra-226 according to the state change of Ac-225 by producing Ac-225 using Ra-226 of a liquefied state, moving the produced Ac-225 in a liquefied state after Ac-225 is produced, and separating and reusing Ac-225, thereby enabling a nuclear reaction process of Ac-225 to be performed. Further, a method of producing actinium by using liquefied radium according to the present disclosure has an effect of enabling safety to be improved by including a radon collection unit which is capable of discharging and isolating radon produced from Ra-226, thereby preventing radiation exposure due to radon. 1. A method of producing actinium by using liquefied radium , the method comprising step of;a step of moving the liquefied radium to load the liquefied radium into a reaction space inside a chamber;a step of producing actinium through a nuclear reaction process by irradiating a particle beam to the liquefied radium of the reaction space inside the chamber; andan unloading step of moving a product comprising the liquefied radium and actinium to the outside of the chamber.2. The method of claim 1 , further comprising step of separating actinium from the product.3. The method of claim 2 , further comprising a reloading step of transferring pure liquefied radium obtained by separating actinium from the product to the reaction space of the chamber.4. The method of claim 2 , further comprising a radon discharge step of discharging radon generated from radium while performing the loading step or the unloading step.5. The method of claim 4 , wherein the radon discharge step comprises condensing radon to discard radon.6. The method of claim 4 , wherein the radon discharge step comprises diluting radon with external air to discharge the diluted radon.7. The method of claim 2 , wherein the loading step comprises moving a preset amount of radium to the reaction space.8. The method of claim 7 , wherein the loading step ...

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Номер записи: 23
30-01-2020 дата публикации

Radioisotope delivery system with multiple detectors to detect gamma and beta emissions

Номер: US20200030522A1
Автор: Adrian Nunn, Andrew SCHIMMOELLER, Brian CHERUNDOLO, Clif MOYERS
Принадлежит: Bracco Diagnostics Inc

A nuclear medicine infusion system (10) may be used to generate and infuse radioactive liquid into a patient undergoing a diagnostic imaging procedure. In some examples, the infusion system includes a frame (30) that carries a radioisotope generator (52) that generates radioactive eluate via elution. The frame may also carry a beta detector (58) and a gamma detector (60). The beta detector can be positioned to measure beta emissions emitted from the radioactive eluate supplied by the generator. The gamma detector can be positioned to measure gamma emissions emitted from a portion of the radioactive eluate to evaluate a safety of the radioactive eluate delivered by the infusion system.

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Номер записи: 24
30-01-2020 дата публикации

SYSTEMS AND TECHNIQUES FOR GENERATING, INFUSING, AND CONTROLLING RADIOISOTOPE DELIVERY

Номер: US20200030523A1
Автор: BOLAS Tarah, CHERUNDOLO Brian, MOYERS Clif, NUNN Adrian, SCHIMMOELLER Andrew
Принадлежит:

An infusion system may include a strontium-rubidium radioisotope generator that generates a radioactive eluate via elution, a beta detector, a gamma detector, and a controller. The beta detector and the gamma detector may be positioned to measure beta emissions and gamma emissions, respectively, emitted from the radioactive eluate. In some examples, the controller is configured to determine an activity of rubidium in the radioactive eluate based on the beta emissions measured by the beta detector and determine an activity of strontium in the radioactive eluate based on the gamma emissions measured by the gamma detector. 1. An infusion system comprising:a frame that carries a beta detector, a gamma detector, and a controller communicatively coupled to the beta detector and the gamma detector,wherein the frame is further configured to receive a strontium-rubidium radioisotope generator that generates a radioactive eluate via elution,the beta detector is positioned to measure beta emissions emitted from the radioactive eluate,the gamma detector is positioned to measure gamma emissions emitted from the radioactive eluate, andthe controller is configured to determine an activity of rubidium in the radioactive eluate based on the beta emissions measured by the beta detector and determine an activity of strontium in the radioactive eluate based on the gamma emissions measured by the gamma detector.2. The infusion system of claim 1 , wherein the gamma detector is positioned to measure the gamma emissions emitted from a static portion of the radioactive eluate.3. The infusion system of claim 1 , wherein the controller is configured to control the infusion system to prevent a patient infusion procedure if the determined activity of strontium exceeds an allowable limit.4. The infusion system of claim 1 , further comprising an infusion tubing line configured to receive the radioactive eluate claim 1 , either directly or indirectly claim 1 , from the strontium-rubidium ...

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Номер записи: 25
02-02-2017 дата публикации

TECHNIQUES FOR ON-DEMAND PRODUCTION OF MEDICAL ISOTOPES SUCH AS MO-99/TC-99M AND RADIOACTIVE IODINE ISOTOPES INCLUDING I-131

Номер: US20170032860A1
Автор: Tsang Francis Yu-Hei
Принадлежит:

A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron's probability of causing fission (n, f) reactions in U-238. The U-238 can be deployed as one or more layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material. The gaseous fission products can be withdrawn from the irradiation chamber on a continuous basis, and the radioactive iodine isotopes (including I-131) extracted. 120-. (canceled)21. A method for producing radioisotopes comprising:introducing non-enriched uranium (“NEU”) material into a an irradiation chamber, the irradiation chamber having one or more walls formed of neutron-reflecting material; at least some neutrons from the irradiating are reflected from at least one of the one or more walls, thereby increasing the path length over which those neutrons are in the NEU material, and', 'the increased path length increases the probability that those neutrons in the NEU material will cause fast fission reactions; and, 'irradiating the NEU material with neutrons having energies above 800 keV to cause fast fission reactions to occur in the NEU material and generate fission products, whereinextracting the fission products from the NEU material.22. The method of wherein one of the fission products extracted comprises at least one of molybdenum-99 (Mo-99) and technetium-99m (Tc-99m).23. The method of wherein one of the fission products extracted comprises at least one of iodine 131 (I-131) and iodine 132 (I-132).24. The method of wherein the NEU material in the irradiation chamber occupies a single spatially contiguous region.25. The method of wherein the NEU material in the irradiation chamber occupies multiple spatially disjoint regions.26. The method of wherein the one or more ...

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Номер записи: 26
04-02-2016 дата публикации

Production of carbon-11 using a liquid target

Номер: US20160035448A1
Автор: Peter Andras Zavodszky, Tiberiu Mircea Siclovan
Принадлежит: General Electric Co

The present disclosure relates to the generation of radioisotopes, includes 11-carbon, from liquid targets. In certain embodiments, a liquid hydrazine target is employed which, when irradiated, such as with a charged particle beam, generates 11-carbon in a form that may be recovered and used in downstream processes, such as the generation of radiopharmaceuticals.

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Номер записи: 27
11-02-2016 дата публикации

Production of copper-67 from an enriched zinc-68 target

Номер: US20160040267A1
Автор: Jon Stoner, Tim Gardner
Принадлежит: Idaho State University

An apparatus including a heating element and a sublimation vessel disposed adjacent the heating element such that the heating element heats a portion thereof. A collection vessel is removably disposed within the sublimation vessel and is open on an end thereof. A crucible is configured to sealingly position a solid mixture against the collection vessel.

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Номер записи: 28
09-02-2017 дата публикации

METHODS AND DEVICES FOR ISOLATING LEAD 203

Номер: US20170040074A1
Автор: Olewine Keith R.
Принадлежит: Lantheus Medical Imaging, Inc.

Methods for isolating Pb and/or Pb isotopes from various sources are provided. Compositions comprising Pb and/or Pb isotopes free of certain amounts of various contaminants are also provided. 1. A method , comprisingcontacting a chelating resin that comprises iminodiacetic acid with a solution comprising Pb, andeluting the Pb bound to the chelating resin with a heated sodium hydroxide solution, wherein the heated sodium hydroxide solution is at a temperature suitable for the selective elution of Pb.2. The method of claim 1 , wherein the temperature of the heated sodium hydroxide solution is about 85-95° C.3. A method claim 1 , comprisingcontacting a chelating resin that comprises iminodiacetic acid with a solution comprising Pb, andeluting the Pb bound to the chelating resin with heated sodium hydroxide solution, wherein the temperature of the sodium hydroxide is at about 90° C.4. A method claim 1 , comprisingeluting Pb from a chelating resin using a sodium hydroxide solution at a temperature of about 85-95° C.or of about 85° C. or higher.5. A composition claim 1 , comprising{'sup': '203', 'Pb and less than 0.1 μg/mCi Ni and/or'}less than 0.1 μg/mCi Cu and/orless than 0.5 μg/mCi Zn and/orless than 0.25 μg/mCi Fe and/orless than 0.05 μg/mCi Tl.6. The composition of claim 5 , further comprising sodium hydroxide. This application claims the benefit of U.S. Provisional Application No. 61/979,957 filed on Apr. 15, 2014, the entire contents of which are incorporated by reference herein.Radioactive isotopes of many metallic elements have potential uses in the diagnosis and treatment of disease. The lead-203 isotope (Pb), for example, which has a half-life of about 52 hours and decays by electron capture, has excellent promise in medical diagnostics. As a result, recent advances in radioimmunotherapy and peptide targeted radiotherapy have created a great demand for Pb.Pb is an important isotope in certain medical applications. For example, because of its relatively short ...

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Номер записи: 29
07-02-2019 дата публикации

Fuel channel isotope irradiation at full operating power

Номер: US20190043630A1
Автор: Benjamin D. Fisher, Thomas G. Onderwater
Принадлежит: BWXT Isotope Technology Group Inc

A method of a method of irradiating a target material in a heavy water reactor for the production of an isotope, including the steps of providing a target comprised of a material suitable for producing the isotope by way of a neutron capture event, placing the target in a primary fluid side of the heavy water reactor, and irradiating the target.

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Номер записи: 30
07-02-2019 дата публикации

METHOD FOR PREPARING RADIOACTIVE SUBSTANCE THROUGH MUON IRRADIATION, AND SUBSTANCE PREPARED USING SAID METHOD

Номер: US20190043631A1
Автор: MATSUZAKI Teiichiro, SAKURAI Hiroyoshi
Принадлежит:

In order to prepare a useful radioactive substance from radionuclides included in high-level radioactive waste and the like, an embodiment of the present invention provides a method for preparing a radioactive substance including a muon irradiation step for obtaining a first radionuclide by causing negative muons to be incident onto a radioactive target nuclide and triggering a nuclear muon capture reaction. The prepared radioactive substance includes at least one of the first radionuclide and a second radionuclide that is at least one type of a descendant nuclide obtained from the first radionuclide through radioactive decay. An embodiment of the present invention also provides the radioactive substance. 1. A method for producing a radioactive substance comprising a muon irradiation step for obtaining a first radionuclide through a muon nuclear capture reaction by irradiating a target nuclide which is a radionuclide with negative muons ,wherein the radioactive substance to be produced comprises at least one of the first radionuclide and a second radionuclide, the second radionuclide being a descendant nuclide obtained from the first radionuclide via radioactive decay.2. The method for producing a radioactive substance according to further comprising preparing a target raw material containing the target nuclide to be irradiated with negative muon prior to the muon irradiation step claim 1 ,wherein the target nuclide in the target raw material is any of radionuclides in long-lived fission products (LLFPs) contained in a spent nuclear fuel or a substance separated from a spent nuclear fuel.3. The method for producing a radioactive substance according to claim 2 , wherein the target nuclide is Tc claim 2 , the first radionuclide is Mo claim 2 , and the second radionuclide is Tc.4. The method for producing a radioactive substance according to claim 1 ,{'sup': 99', '99', '99m, 'wherein the target nuclide is Tc, the first radionuclide is Mo, and the second radionuclide is ...

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Номер записи: 31
07-02-2019 дата публикации

Light-Nuclei Element Synthesis

Номер: US20190043632A1
Автор: Mathew M. Zuckerman
Принадлежит: Nex-Gen Solar Technologies LLC

A system and method for the synthesis of light-nuclei elements (LNEs), including the battery element Lithium, in high-purity form. The method eliminates the need for high-energy proton collision in Cosmic Rays to produce Nitrogen-15. LNEs are produced by placing a mixture with carbon, nitrogen, and oxygen (CNO) source material in a strong, fixed magnetic field, then introducing instability to the CNO's stable isotopes through high-frequency radio waves tuned to the nuclear magnetic resonance (NMR) frequency of a target material in the mixture to produce a LNE product material, and then separating the LNE product material from other materials within the mixture by enhancing gravity separation based on the opposite signs of respective dipole magnetic moments (DMM) to cause attraction of the product material, such as Lithium, to the South magnetic pole away from another product material, such as Beryllium, that is attracted to the North magnetic pole.

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Номер записи: 32
18-02-2021 дата публикации

NEW METHOD AND APPARATUS FOR THE PRODUCTION OF HIGH PURITY RADIONUCLIDES

Номер: US20210050126A1
Автор: Andreoletti Gilbert, Bourdet Patrick, Dureau Rémy, Maquaire Patrick, Torgue Julien
Принадлежит:

An apparatus is for the automated production of a daughter radionuclide from a parent radionuclide using a generator comprising a solid medium onto which the parent nuclide is fixed and whereby the daughter nuclide is formed by radioactive decay of the parent nuclide. The apparatus includes a fluid circuit including a chromatography column having a head port and a tail port, at least one connection port for connecting the generator to the fluid circuit, at least one inlet port for connecting fluid sources to the fluid circuit and at least one valve controlled by an electronic control unit for selectively connecting the chromatography column, the connection port and the at least one inlet port in various configurations. The various configurations include a first elution configuration for circulating an A′ solution exiting the generator and containing the daughter radionuclide, through the chromatography column from the head port to the tail port for loading the chromatography column with the daughter radionuclide; a first washing configuration for circulating an A washing solution from a solution inlet through the chromatography column from the head port to the tail port; and a second washing configuration for circulating an A′ washing solution from a solution inlet through the chromatography column from the tail port to the head port. 1. An apparatus for the automated production of a daughter radionuclide from a parent radionuclide using a generator comprising a solid medium onto which the parent nuclide is fixed and whereby the daughter nuclide is formed by radioactive decay of the parent nuclide , the apparatus comprising a fluid circuit comprising:a chromatography column having a head port and a tail port;at least one connection port for connecting the generator to the fluid circuit;at least one inlet port for connecting fluid sources to the fluid circuit; and [{'b': '1', 'a first elution configuration for circulating an A′ solution exiting the generator and ...

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Номер записи: 33
25-02-2016 дата публикации

Method for preparation of alpha sources of polonium using sulfide micro-precipitation

Номер: US20160055928A1
Автор: Nicolas Guerin, Xiongxin Dai
Принадлежит: Atomic Energy of Canada Ltd AECL

A method for preparing alpha sources of polonium. A sample of polonium is provided in a solution. A controlled amount of sulfide and a controlled amount of a metal capable of forming an insoluble sulfide salt in the solution are introduced into the solution, in order to co-precipitate polonium from the solution. The precipitates are filtered out.

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Номер записи: 34
21-02-2019 дата публикации

METHODS FOR PRODUCING Cu-67 RADIOISOTOPE WITH USE OF A CERAMIC CAPSULE FOR MEDICAL APPLICATIONS

Номер: US20190057791A1
Автор: EHST David A., Willit James L.
Принадлежит: UCHICAGO ARGONNE, LLC

The present invention provides a target unit and a sublimation apparatus for use in a method for producing Cu67 radioisotope suitable for use in medical applications. The method comprises irradiating a metallic zinc-68 (Zn68) target within a sealed ceramic capsule with a high energy gamma ray beam. After irradiation, the Cu67 is isolated from the Zn68 by any suitable method (e.g. chemical and or physical separation). In a preferred embodiment, the Cu67 is isolated by sublimation of the zinc in a ceramic sublimation tube to afford a copper residue containing Cu67. The Cu67 can be further purified by chemical means. 1. A target unit for producing Cu67 radioisotope comprising:a cage body releasably coupled to a screw-on cap; anda ceramic capsule containing a solid Zn68 target ingot and having one open end and one closed end and defining an interior chamber for the target;wherein the ceramic capsule is releasably sealed within the target body between the cage body and the screw-on cap to form a substantially water-tight seal over the open end of the capsule; the interior of the ceramic capsule is in intimate physical contact with the solid Zn68 target ingot; and the Zn68 of the target is substantially free of traces of residual oxygen that interfere with contact of the Zn68 to the capsule.2. The target unit of claim 1 , wherein the cage body and the screw-on cap are composed of aluminum.3. The target unit of claim 1 , wherein the cage body and the screw-on cap are composed of different alloys of aluminum to minimize the possibility of thread galling.4. The target unit of claim 1 , wherein the cage body and the screw-on cap are each composed of different alloys of aluminum selected from the group consisting of 6061 Al and 2024 Al.5. The target unit of claim 1 , wherein the cage body includes apertures to allow cooling water to contact the capsule during irradiation thereof to prevent melting or partial melting of the zinc target ingot during the irradiation.6. The target ...

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Номер записи: 35
01-03-2018 дата публикации

PRODUCTION OF MOLYBDENUM-99 USING ELECTRON BEAMS

Номер: US20180061516A1
Автор: de JONG Mark, Diamond William, LIN Linda, NAGARKAL Vinay, REGIER Christopher, ULLRICH Douglas
Принадлежит: Canadian Light Source Inc.

An apparatus for producing Mo from a plurality of Mo targets through a photo-nuclear reaction on the Mo targets. The apparatus comprises: (i) an electron linear accelerator component; (ii) an energy converter component capable of receiving the electron beam and producing therefrom a shower of bremsstrahlung photons; (iii) a target irradiation component for receiving the shower of bremsstrahlung photons for irradiation of a target holder mounted and positioned therein. The target holder houses a plurality of Mo target discs. The apparatus additionally comprises (iv) a target holder transfer and recovery component for receiving, manipulating and conveying the target holder by remote control; (v) a first cooling system sealingly engaged with the energy converter component for circulation of a coolant fluid therethrough; and (vi) a second cooling system sealingly engaged with the target irradiation component for circulation of a coolant fluid therethrough. 1. An apparatus for producing molybdenum-99 (Mo) from a plurality of molybdenum-100 (Mo) targets through a photo-nuclear reaction on the Mo targets , the apparatus comprising:a linear accelerator component capable of producing an electron beam;a converter component capable of receiving the electron beam and producing therefrom a shower of bremsstrahlung photons;{'sup': '100', 'a target irradiation component for receiving the shower of bremsstrahlung photons, the target irradiation component having a chamber for receiving, demountingly engaging, and positioning therein a target holder housing a plurality of Mo target discs;'}a remote controlled grapple assembly transportable along and within the apparatus, the grapple assembly demountably engageable with an end of the target holder; anda cooling system sealingly engaged with the converter component for circulation of a coolant fluid therethrough.2. The apparatus according to claim 1 , wherein the linear accelerator component has at least 10 kW of power to about 100 kW ...

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Номер записи: 36
28-02-2019 дата публикации

INTEGRATED STRONTIUM-RUBIDIUM RADIOISOTOPE INFUSION SYSTEMS

Номер: US20190060558A1
Автор: Fontaine Aaron M., Gelbach Janet L., Hidem Stephen E., Hunter Kathryn M., McDonald Patrick M., Swenson Rolf E., ZODDA Julius P.
Принадлежит:

Methods for setting up, maintaining and operating a radiopharmaceutical infusion system, that includes a radioisotope generator, are facilitated by a computer of the system. The computer may include pre-programmed instructions and a computer interface, for interaction with a user of the system, for example, in order to track contained volumes of eluant and/or eluate, and/or to track time from completion of an elution performed by the system, and/or to calculate one or more system and/or injection parameters for quality control, and/or to perform purges of the system, and/or to facilitate diagnostic imaging. 1. A method of building an infusion system to deliver a rubidium radioactive eluate comprising: the first opening is configured for a strontium-rubidium radioisotope generator to be inserted into and removed from the first shielding compartment,', 'the second opening is configured for a waste bottle to be inserted into and removed from the second shielding compartment, and', 'the shielded well is configured to receive an eluate reservoir, the eluate reservoir being configured to receive a sample of the rubidium radioactive eluate;, 'installing on a cart a first shielding compartment having a first opening, a second shielding compartment having a second opening, and a shielded well, wherein collect the sample of the rubidium radioactive eluate in the eluate reservoir by pumping saline from a saline reservoir into the strontium-rubidium radioisotope generator via a saline tubing line thereby generating the rubidium radioactive eluate that is discharged through an eluate tubing line,', 'determine a strontium breakthrough test result on the sample of the rubidium radioactive eluate collected in the eluate reservoir in the shielded well on-board the cart while the eluate reservoir remains in the shielded well on-board the cart,', 'not allow a patient infusion if the strontium breakthrough test result is greater than or equal to an allowed limit,', 'track a volume of ...

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Номер записи: 37
28-02-2019 дата публикации

Radioisotope Production

Номер: US20190066859A1
Автор: BIJLSMA Sipke Jacob, De Jager Pieter Willem Herman, Derksen Antonius Theodorus Anna Maria, Frijns Olav Waldemar Vladimir, KASTELIJN Aukje, LANSBERGEN Robert Gabriël Maria, NIKIPELOV Andrey Alexandrovich, Ten Kate Nicolaas, Verspay Jacobus Johannus Leonardus Hendricus
Принадлежит: ASML Netherlands B.V.

A radioisotope production apparatus (RI) comprising an electron source arranged to provide an electron beam (E). The electron source comprises an electron injector () and an electron accelerator (). The radioisotope production apparatus (RI) further comprises a target support structure configured to hold a target () and a beam splitter () arranged to direct the a first portion of the electron beam along a first path towards a first side of the target () and to direct a second portion of the electron beam along a second path towards a second side of the target (). 175-. (canceled)76. A radioisotope production apparatus comprising:an electron source arranged to provide an electron beam, the electron source comprising an electron injector and an electron accelerator;a target support structure configured to hold a target; anda beam splitter arranged to direct a first portion of the electron beam along a first path towards a first side of the target and to direct a second portion of the electron beam along a second path towards a second side of the target,wherein the electron beam comprises pulses, andwherein the beam splitter is arranged to direct substantially half of the pulses along the first path and half of the pulses along the second path.77. The radioisotope production apparatus of claim 76 , wherein the beam splitter comprises a deflector.78. The radioisotope production apparatus of claim 76 , wherein:the target comprises an electron target and a photon target; andthe electron target is arranged to receive at least one of the first and second portions of the electron beam and to emit photons towards the photon target.79. The radioisotope production apparatus of claim 78 , wherein the electron target comprises a first part arranged to receive the first portion of the electron beam and a second part arranged to receive the second portion of the electron beam.80. The radioisotope production apparatus of claim 79 , wherein the first and second parts of the electron ...

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Номер записи: 38
28-02-2019 дата публикации

68Ge/68Ga Generator

Номер: US20190066860A1
Автор: Harfensteller Mark, Jernström Jussi, Meckel Marian, Totskiy Yury, Zhernosekov Konstantin
Принадлежит:

A Ge/Ga generator for a continuous production of a Ga daughter nuclide, wherein the Ge parent nuclide thereof is specifically adsorbed to an inorganic support material and wherein said Ge parent nuclide continuously decays to Ga by electron capture at a half-life of 270.82 d, wherein the inorganic support material is at least one oxide of a metal being selected from the group consisting of: Vanadium, Niobium and Tantalum. The use of at least one oxide of a metal being selected from the group consisting of: Vanadium, Niobium and Tantalum as an inorganic support material for the manufacture of a Ge/Ga generator for pharmaceutical purposes. With the inorganic support material of the present invention, it is possible to load Ge/Ga generators with up to 8000 MBq of Ge (corresponding to 80 μg Germanium). 118-. (canceled)19. A Ge/Ga generator for a continuous production of a Ga daughter nuclide , wherein the Ge parent nuclide thereof is specifically adsorbed to an inorganic support material and wherein said Ge parent nuclide continuously decays to Ga by electron capture at a half-life of 270.82 d ,characterized in thatthe inorganic support material is at least one oxide of a metal being selected from the group consisting of Vanadium, Niobium and Tantalum.20. The generator according to claim 19 , characterized in that the oxide is an oxide having the general formula (1):{'br': None, 'sub': 2', '5, 'MO\u2003\u2003(1),'}wherein M is selected from the group consisting of Vanadium, Niobium and Tantalum.21. The generator according to claim 19 , characterized in that the oxide is tantalum pentaoxide (TaO).22. The generator according to claim 21 , characterized in that said TaOis present in its alpha- and/or beta-crystalline form.23. The generator according to claim 19 , characterized in that the oxide is obtainable by hydrolyzing a metal halogenide of the general formula (2):{'br': None, 'sub': '5', 'MX\u2003\u2003(2),'}wherein M is selected from the group consisting of Vanadium, ...

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Номер записи: 39
27-02-2020 дата публикации

RADIOISOTOPE TARGET STATION

Номер: US20200066418A1
Автор: Bailey James L., BROWN Michael Alexander, Chemerisov Sergey A., EHST David A., Grudzinski James J., Jr. Jerry A., Kmak Ronald T., Nolen, Rotsch David A., Smith Nicholas A.
Принадлежит: UCHICAGO ARGONNE, LLC

A system for producing and harvesting radioisotopes is provided, the system having a converter housing defining a first beam window; a converter carrier and cartridge in slidable communication with the converter housing; a target housing positioned downstream from the converter housing, the target housing defining a second beam window; and a target carrier in slidable communication with the target housing. 1. A system for producing radioisotopes , the system comprising:a) a converter housing having an upstream end and a downstream end, the upstream end defining a first window;b) a converter carrier in slidable communication with the converter housing;c) a target housing positioned downstream from the converter housing, the target housing defining a second window; andd) a target carrier removably received by the target housing, wherein the first window, the converter housing, the converter carrier, the target housing and the target carrier define a tunnel adapted to receive a particle beam.2. The system as recited in wherein the target carrier communicates with an upwardly facing surface of the target housing.3. The system as recited in wherein a coolant fluid is in thermal communication with the converter housing and the target housing.4. The system as recited in further comprising a target capsule adapted to be received by the target carrier.5. The system as recited in wherein the target capsule is adapted to receive target isotope having a weight of between 1 mg and 100 claim 4 ,000 mg.6. The system as recited in wherein the first window has a convex topography relative to the upstream end.7. The system as recited in wherein the first window has a flat topography relative to the upstream end.8. The system as recited in wherein the particle beam comprises an incident electron beam having an energy ranging from 0 MeV to 100 MeV.9. The system as recited in wherein the particle beam comprises an incident electron beam having a beam power ranging from 0 kW to 100 kW.10 ...

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Номер записи: 40
17-03-2016 дата публикации

System And Method For Generating Molybdenum-99 And Metastable Technetium-99, And Other Isotopes

Номер: US20160078971A1
Автор: Clayton James E.
Принадлежит:

Accelerator based systems are disclosed for the generation of isotopes, such as molybdenum-98 (“99Mo”) and metastable technetium-99 (“99mTc”) from molybdenum-98 (“98Mo”). Multilayer targets are disclosed for use in the system and other systems to generate 99mTc and 98Mo, and other isotopes. In one example a multilayer target comprises a first, inner target of 98Mo surrounded, at least in part, by a separate, second outer layer of 98Mo. In another example, a first target layer of molybdenum-100 is surrounded, at least in part, by a second target layer of 98Mo. In another example, a first inner target comprises a Bremsstrahlung target material surrounded, at least in part, by a second target layer of molybdenum-100, surrounded, at least in part, by a third target layer of 98Mo. 1. A system for generating isotopes , comprising:an accelerator;a source of charged particles coupled to the accelerator to inject charged particles into the accelerator;a target comprising:a first, inner target material, comprising a first isotope of a first material; anda second, outer target material comprising a second isotope of a second material, the second outer target material at least partially surrounding the first, inner target material, the second, outer target material defining a passage for accelerated charged particles to the first, inner target material.2. The system of claim 1 , wherein the first material and the second material are the same and the first isotope and the second isotope are different isotopes of the first material.3. The system of claim 2 , wherein the first claim 2 , inner target material and the second claim 2 , outer target material are separated by a gap.4. The system of claim 1 , wherein the first isotope and the second isotope each comprise molybdenum-98.5. The system of claim 2 , wherein the first isotope comprises molybdenum-100 and the second isotope comprises molybdenum-98.6. The system of claim 1 , wherein the target further comprises a layer of ...

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Номер записи: 41
24-03-2022 дата публикации

COMPACT ASSEMBLY FOR PRODUCTION OF MEDICAL ISOTOPES VIA PHOTONUCLEAR REACTIONS

Номер: US20220093283A1
Автор: Brown Michael A., Chemerisov Sergey D., Henning Walter F., Jr. Jerry A., Nolen, Rotsch David A., Song Jeongseog
Принадлежит: UCHICAGO ARGONNE, LLC

The invention provides a method for generating medical isotopes, the method comprising contacting a primary radiation beam with a converter for a time sufficient to produce a secondary beam of gamma particles, and contacting the beam of gamma particles to a target, where the cross section dimension of the beam of gamma particles is similar to the cross section dimension of the target. Both the converter and target are small in diameter and very closely spaced. Also provided is a system for producing medical isotopes, the device comprising a housing having a first upstream end and a second downstream end, a radiotransparent channel (collimator) with a first upstream end and a downstream end, wherein the upstream end is adapted to receive a radiation beam, a target positioned downstream of the downstream end of the channel and coaxially aligned with the channel, wherein the target has a cross section that is similar to the cross section of the channel. 1. A system for producing medical isotopes , the system comprising:a) a housing having a first upstream end and a second downstream end;b) a radiotransparent channel with a first upstream end and a downstream end, wherein the upstream end is adapted to receive a radiation beam; andc) a target positioned downstream of the downstream end of the channel and coaxially aligned with the channel, wherein the target has a cross section that is similar to the cross section of the channel.211. The system as recited in claim further comprising a heat sink in thermal communication with the channel and target.311. The system as recited in claim further comprising a converter disposed between the channel and the target , wherein the converter and front face of the target are protected from degradation by air by being bathed in a vacuum or an inert fluid.412. The system as recited in claim wherein the heat sink is a fluid and the housing defines a first fluid conduit and a second fluid conduit.514. The system as recited in claim ...

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Номер записи: 42
21-03-2019 дата публикации

HIGH PURITY THERAPEUTIC BONE AGENTS

Номер: US20190083661A1
Автор: Frank R. Keith, Simon Jaime, WILSON David A.
Принадлежит: IsoTherapeutics Group, LLC

This invention relates to radioactive, bone-seeking, pharmaceutical methods, compositions and formulations that have a lower impurity profile, a longer shelf life, improved availability and are less expensive to prepare. The compositions of this invention can be conveniently prepared in a timely manner resulting in improved availability and delivery of the drugs to patients. 1. A method for the treatment of a Patient comprising administration to the Patient having bone pain , one or more calcific tumors , or in need of a bone marrow suppressing procedure , a pharmaceutically-acceptable , formulation of a chelate composition comprising a Clinically Relevant Dosage of the composition that is therapeutically effective , said composition either:i) has said formulation as a chelate composition consisting essentially of Sm-153 and DOTMP or a physiologically-acceptable salt thereof, orii) has said formulation comprising a kit containing as two separate components, the DOTMP chelant and the Sm-153 isotope, which components are mixed to form the chelate composition at the appropriate time prior to use,wherein the Sm-153 isotope possesses an extended Expiration Date of greater than or equal to about 5 days, and wherein the Sm-153 dosage is at least 35 mCi;with the proviso that the chelate composition is prepared by a process comprising the steps of:{'sup': '2', 'a) irradiating Sm-152 in a lower flux portion of the nuclear reactor having less than 8.5×1013 neutron/cm-sec to form low specific activity Sm-153, wherein the isotope composition after the irradiation contains mainly Sm-152 and Sm-153 with reduced trace impurity of Eu-154 less than 0.093 μCi Eu-154/mCi Sm-153 after 5 days of decay, thereby providing an extended Expiration Date greater than or equal to about 5 days;'}b) taking the prepared isotope mixture from step a) and either using it in step c) or allowing it to decay and then using it in step c) which decay further lowers the specific activity of the Sm-153 ...

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Номер записи: 43
12-03-2020 дата публикации

Sytem and method for collecting and isolating radiosotopes

Номер: US20200082956A1
Автор: Amanda J. Youker, Artem Gelis, David Rotsch, Jerry A. Nolen, JR., Michael A. Brown
Принадлежит: UChicago Argonne LLC

A method for obtaining 225AC from 225Ra having the steps of assembling a column having an inorganic stationary phase; priming the column to immobilize 226Ra 225Ra and natural decay products therefrom; immobilizing the 226Ra, 225Ra, 224Ra, and natural decay products therefrom onto a stationary phase within the column; and eluting the column containing the 225Ra with an aqueous sulfate solution to obtain a milking effluent that contains 225AC. Also provided is a method for obtaining pure 225AC from its isotope parents, the method comprising assembling a column having a stationary phase comprising an inorganic material; priming the column with the isotope parents to immobilize 225Ac, and natural decay products of 225AC; immobilizing the 225Ac, and natural decay products therefrom onto the stationary phase within the column 226Ra, 225Ra, 224Ra; and eluting the column containing the 225AC to obtain an effluent that contains the isotope parents.

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Номер записи: 44
30-03-2017 дата публикации

PRODUCTION OF 43SC RADIONUCLIDE AND RADIOPHARMACEUTICALS THEREOF FOR USE IN POSITRON EMISSION TOMOGRAPHY

Номер: US20170087260A1
Автор: BUNKA MARUTA, TUERLER ANDREAS, VAN DER MEULEN NICHOLAS
Принадлежит:

The radionuclide Sc is produced at commercially significant yields and at specific activities and radionuclidic purities which are suitable for use in radiodiagnostic agents including imaging agents. In a method, a solid target having an isotopically enriched target layer prepared on an inert substrate is positioned in a specially configured target holder and irradiated with a charged-particle beam of protons or deuterons. The beam is generated using an accelerator such as a biomedical cyclotron at energies ranging from 3 to about 22 MeV. The method includes the use of three different nuclear reactions: a) irradiation of enriched Ca targets with protons to generate the radionuclide Scin the nuclear reaction Ca (p,n)Sc, b) irradiation of enriched Ca targets with deuterons to generate the radionuclide Sc in the nuclear reaction Ca(d,n)Sc, and c) irradiation of enriched Ti targets with protons to generate the radionuclide Sc in the nuclear reaction Ti (p,a)Sc. 17-. (canceled)8. A method for generating Sc , wherein one of the following method steps is applied:{'sup': 43', '43', '43, 'performing a nuclear reaction of Ca(p,n)Sc using enriched Ca at proton beam energies of 5 to 24 MeV;'}{'sup': 42', '43', '42, 'performing a nuclear reaction of Ca(d,n)Sc using enriched Ca and deuteron beam energies of 3 to 12 Mev; or'}{'sup': 46', '43', '46, 'performing a nuclear reaction of Ti(p,α)Sc using enriched Ti and proton beam energies of 10 to 24 MeV.'}9. The method according to claim 8 , which further comprises:{'sup': 43', '43', '43', '43, 'sub': 3', '3', '2', '2, 'irradiating the enriched Ca target in form of CaCO, Ca (NO), CaFor CaO powders of Ca metal having a Ca content of 50% or higher with the proton beam thereby turning the Ca content into the Sc;'}{'sup': 43', '43, 'dissolving an irradiated enriched Ca target in acidic solution and passing a resulting solution through a first column loaded with DGA resin in order to absorb Sc ions;'}{'sup': 43', '43, 'eluting absorbed Sc ...

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Номер записи: 45
05-04-2018 дата публикации

INTEGRATED STRONTIUM-RUBIDIUM RADIOISOTOPE INFUSION SYSTEMS

Номер: US20180093035A1
Автор: Fontaine Aaron M., Gelbach Janet L., Hidem Stephen E., Hunter Kathryn M., McDonald Patrick M., Swenson Rolf E., ZODDA Julius P.
Принадлежит:

Methods for setting up, maintaining and operating a radiopharmaceutical infusion system, that includes a radioisotope generator, are facilitated by a computer of the system. The computer may include pre-programmed instructions and a computer interface, for interaction with a user of the system, for example, in order to track contained volumes of eluant and/or eluate, and/or to track time from completion of an elution performed by the system, and/or to calculate one or more system and/or injection parameters for quality control, and/or to perform purges of the system, and/or to facilitate diagnostic imaging. 1a first shielding compartment having a first opening facing vertically upwardly through which a strontium-rubidium radioisotope generator can be inserted into and removed from the first shielding compartment;a first door configured to provide access to the first shielding compartment and to close over the first opening;a second shielding compartment having a second opening facing vertically upwardly through which the waste bottle can be inserted into and removed from the second shielding compartment;wherein the first opening is located at a lower elevation than the second opening;a radioactivity detector positioned to measure radioactivity of a rubidium radioactive eluate flowing through an eluate tubing line in fluid communication with an outlet tubing port of the strontium-rubidium radioisotope generator;a computer configured to receive an input from a user for controlling operation of the infusion system;a shielded well configured to receive an eluate reservoir, wherein the eluate reservoir is configured to receive a test sample; and pump saline from a saline reservoir into the strontium-rubidium radioisotope generator through an inlet tubing port of the strontium-rubidium radioisotope generator thereby generating the rubidium radioactive eluate that is discharged through the outlet tubing port,', 'fill the eluate reservoir in the shielded well with the test ...

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Номер записи: 46
12-05-2022 дата публикации

PURIFICATION OF ACTINIUM

Номер: US20220145423A1
Автор: CARDINAELS Thomas, MAERTENS Dominic, VAN HECKE Karen
Принадлежит:

A method for purifying Ac from a mixture includes Ac and at least one element selected from Ra, Pb, Po, Bi and La. The method includes the steps of: (a) performing a first separation using a first extraction chromatographic column based on a first resin (either a diglycolamide resin or a dialkylphosphoric acid resin) and a first matrix solution; and (b) performing a second separation using a second extraction chromatographic column based on a second resin (respectively either a dialkylphosphoric acid resin or a diglycolamide resin). 116.-. (canceled)17. A method for purifying Ac from a mixture comprising Ac and at least one element selected from Ra , Pb , Po , Bi and La , the method comprising: a1. loading the mixture on a first extraction chromatographic column based on a first resin and a first matrix solution,', 'a2. washing the mixture loaded on the first extraction chromatographic column with a first washing solution, and', 'a3. eluting the mixture loaded on the first extraction chromatographic column with a first eluent to obtain a first eluate; and, 'a. performing a first separation, comprising'} b1. loading the first eluate on a second extraction chromatographic column based on a second resin and a second matrix solution,', 'b2. washing the first eluate loaded on the second extraction chromatographic column with a second washing solution, and', 'b3. eluting the first eluate loaded on the second extraction chromatographic column with a second eluent to obtain a second eluate containing the purified Ac;, 'b. performing a second separation, comprising'}wherein eitherthe first resin is a diglycolamide resin, the first matrix solution has a pH between −0.8 and 0, the first washing solution has a pH between −0.8 and 0 and the first eluent has a pH between 1 and 4, andthe second resin is a dialkyl phosphoric acid resin, the second matrix solution has a pH between 1 and 4, the second washing solution has a pH between 1 and 4 and the second eluent has a pH between −0 ...

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Номер записи: 47
19-03-2020 дата публикации

Method for Enhanced Nuclear Reactions

Номер: US20200090822A1
Автор: Lo Shui Yin
Принадлежит:

A method for enhanced nuclear reactions includes the steps of providing a first target which is a thin film with solid deuterium; providing a deuteron beam having an extreme ultraviolet laser and a first infrared laser to apply on the first target to ionize the deuterium to form positive charged deuterons and electrons; providing a second infrared laser to the first target to accelerate the electrons and the positively charged deuterons; separating the accelerated electrons and the accelerated positively charged deuterons under a magnetic field; providing the accelerated electrons to move in a circular motion and the accelerated positive charger deuterons to move to form a cluster of accelerated positively charged deuterons, and breaking the cluster of accelerated positive charged deuterons into small pieces of positively charged deuterons. 1. A method for enhanced nuclear reactions , comprising the steps of:providing a first target which is a thin film with solid deuterium;providing an extreme ultraviolet laser and first infrared lasers to apply on the first target to ionize the deuterium to form positive charged deuterons and electrons;providing a second infrared laser to the first target to accelerate the electrons and the positively charged deuterons;separating the accelerated electrons and the accelerated positively charged deuterons under a magnetic field;providing the accelerated electrons to move in a circular motion and the accelerated positive charger deuterons to move to form a cluster of accelerated positively charged deuterons; andbreaking the cluster of accelerated positive charged deuterons into small pieces of positively charged deuterons.2. The method for enhanced nuclear reactions claim 1 , as recited in claim 1 , wherein the first target is selected from a group consisting of lithium 7 and boron 11.3. The method for enhanced nuclear reactions claim 1 , as recited in claim 1 , wherein the nuclear reaction is a deuteron-deuterons reaction.4. The ...

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Номер записи: 48
19-03-2020 дата публикации

PNEUMATICALLY OPERATED TARGET IRRADIATION SYSTEMS FOR THE PRODUCTION OF RADIOISOTOPES

Номер: US20200090824A1
Автор: Alboino Mark A., Fisher Benjamin D., Long Andrew F., Onderwater Thomas G.
Принадлежит:

A target irradiation system for irradiating a radioisotope target in a vessel penetration of a fission reactor, including a target elevator assembly including a body portion defining a central bore and an open bottom end, a center tube that is disposed within the central bore of the body portion, a target basket that is slidably receivable within the center tube, and a winch that is connected to the target basket by a cable, wherein the target basket is configured to receive the radioisotope target therein and be lowered into the vessel penetration of the reactor when irradiating the radioisotope target. 1. A target irradiation system for irradiating a radioisotope target in a vessel penetration of a fission reactor , comprising:a target elevator assembly including a body portion defining a central bore and an open bottom end, a center tube that is disposed within the central bore of the body portion, a target basket that is slidably receivable within the center tube, and a winch that is connected to the target basket by a cable, the target elevator assembly being affixed to the vessel penetration of the reactor; anda target passage that is in fluid communication with the target elevator assembly,wherein the target basket is configured to receive the radioisotope target therein via the target passage and be lowered into the vessel penetration of the reactor when irradiating the radioisotope target, and the target elevator assembly forms a portion of the pressure boundary of the reactor when in fluid communication with the reactor.2. The target irradiation system of claim 1 , wherein the fission reactor is a heavy-water moderated fission reactor and the vessel penetration is an adjuster assembly port.3. The target irradiation system of claim 2 , wherein the radioisotope target is comprised of natural molybdenum.4. The target irradiation system of claim 1 , wherein the target elevator assembly is affixed to the vessel penetration so that a portion of the body portion ...

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Номер записи: 49
12-05-2022 дата публикации

Methods of producing enriched scandium-47, and related systems and apparatuses

Номер: US20220148752A1
Автор: Mathew S. Snow, Mathew T. Kinlaw
Принадлежит: Battelle Energy Alliance Llc

A method of producing enriched 47 Sc comprises irradiating a V structure comprising 51 V with at least one incident photon beam having an endpoint energy within a range of from about 14 MeV to about 44 MeV to convert at least some of the 51 V to 47 Sc and form a 47 Sc-containing structure. The 47 Sc of the 47 Sc-containing structure is separated from additional components of the 47 Sc-containing structure using a chromatography process. Systems and apparatuses for producing enriched 47 Sc are also described.

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Номер записи: 50
04-04-2019 дата публикации

COMPACT ASSEMBLY FOR PRODUCTION OF MEDICAL ISOTOPES VIA PHOTONUCLEAR REACTIONS

Номер: US20190103198A1
Автор: BROWN Michael Alexander, Chemerisov Sergey D., Henning Walter F., Jr. Jerry A., Nolen, Rotsch David A., Song Jeongseong
Принадлежит: UCHICAGO ARGONNE, LLC

The invention provides a method for generating medical isotopes, the method comprising contacting a primary radiation beam with a converter for a time sufficient to produce a secondary beam of gamma particles, and contacting the beam of gamma particles to a target, where the cross section dimension of the beam of gamma particles is similar to the cross section dimension of the target. Both the converter and target are small in diameter and very closely spaced. Also provided is a system for producing medical isotopes, the device comprising a housing having a first upstream end and a second downstream end, a radiotransparent channel (collimator) with a first upstream end and a downstream end, wherein the upstream end is adapted to receive a radiation beam, a target positioned downstream of the downstream end of the channel and coaxially aligned with the channel, wherein the target has a cross section that is similar to the cross section of the channel. 1. A method for generating medical isotopes , the method comprising:a) supplying a target enriched in a specific isotope; andb) impinging a radiation beam onto the target for a time sufficient to produce medical isotopes, wherein a cross section of the beam is approximately same as a cross section of the target.2. The method as recited in further comprising a converter positioned between the beam and the target.3. The method as recited in wherein the beam traverses a collimator which limits the beam diameter to be slightly smaller than the diameter of the converter.4. The method as recited in wherein a coolant is in thermal communication with the collimator claim 3 , the converter claim 3 , and the target.5. The method as recited in wherein the radiation beam is comprised of electrons with an energy between 30 MeV and 45 MeV with beam total beam power approximately 10 to 20 kW.6. The method as recited in wherein the target is 226Ra and 225Ac is the medical isotope produced.7. The method as recited in wherein only about ...

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Номер записи: 51
21-04-2016 дата публикации

PRODUCTION OF ISOTOPES USING HIGH POWER PROTON BEAMS

Номер: US20160111176A1
Автор: Gomes Itacil C., Jr. Jerry A., Nolen
Принадлежит: UCHICAGO ARGONNE, LLC

The invention provides for a method for producing isotopes using a beam of particles from an accelerator, whereby the beam is maintained at between about 70 to 2000 MeV; and contacting a thorium-containing target with the particles. The medically important isotope Ac is produced via the nuclear reaction (p,2p6n),whereby an energetic proton causes the ejection of 2 protons and 6 neutrons from a Th target nucleus. Another medically important isotope Bi is then available as a decay product. The production of highly purified At is also provided. 1. A method for producing astatine isotope , the method comprising:a. irradiating a thorium target for a time and at an energy sufficient to produce radon isotopes;b. extracting the radon isotopes from the target;c. condensing and purifying the extracted radon isotopes; andd. generating At from the purified radon isotopes.2. The method as recited in wherein the target is irradiated with protons maintained at an energy of about 100 to 400 MeV.3. The method as recited in wherein At is generated from the decay of Rn and chemically separated from radon gas.4. The method as recited in wherein the step of extracting the radon isotopes from the target comprises heating the irradiated target.5. The method as recited in wherein the radon is extracted along with co-extractants as a gas mixture and the step of purifying the radon comprises subjecting the gas mixture to a cold trap to separate the co-extractants from radon.6. The method as recited in wherein the radon remains in vapor phase.7. The method as recited in wherein the radon is continuously extracted from the target. This Utility Patent Application claims priority benefit as a Divisional of U.S. Non-Provisional Application No. 13/025,079, filed on Feb. 10, 2011, presently pending, which in turn claims priority benefit as a Non-Provisional Application of U.S. Provisional Application No. 61/303,023 filed on Feb. 10, 2010, presently expired, the entirety of both Applications ...

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Номер записи: 52
20-04-2017 дата публикации

Method to produce a high-purity Zr-89 through physical irradiation and measurement thereof

Номер: US20170110211A1
Автор: CHU Han-Hsiang, Duh Ting-Shien, LI MING-HSIN, Lin Wuu-Jyh
Принадлежит:

A method to produce a high-purity Zr-89 on a solid target through physical irradiation and measurement by selecting a target Barn value of the cross-sectional area of nuclear reaction, drawing a horizontal line to intersect at two points on the function diagram curve and drawing a vertical line downward from each of the two points intersecting at X-axis to obtain incident energy values at the two intersecting points on the X-axis, and followed by plotting an attenuation function diagram curve of penetration depth versus incident energy of Y-89(p,n)Zr-89, selecting an attenuation function diagram curve and a minimum attenuation position of the selected attenuation function diagram curve in correspondence to the incident energy in the interval of incident energy absorption range to obtain an optimal plating thickness value on the solid target. 1. A method of physical irradiation and measurement for producing a high purity Zr-89 on a solid target , comprising steps:{'b': '11', 'Step S, plotting a function diagram curve of nuclear incident energy versus reaction cross-sectional area for each of Y-89(p, n) Zr-89 and relevant radionuclide zirconium (Zr)-88, zirconium (Zr)-87, and the kinds in accordance with each of their atomic physical characteristics, and providing an equation for the function diagram curve;'}{'b': 12', '1', '2, 'Step S, selecting a target Barn value of the cross-sectional area of nuclear reaction and drawing a horizontal line to intersect at two points on the function diagram curve of nuclear incident energy versus reaction cross-sectional area, followed by drawing a vertical line downward from each of the two points on the function diagram curve and intersecting at X-axis to obtain incident energy values (E, E) at the two intersecting points on the X-axis;'}{'b': 13', '1', '2', '1', '2, 'Step S, substituting the two incident energy values (E, E) into the equation of each of the function diagram curve of nuclear incident energy versus reaction cross- ...

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Номер записи: 53
13-05-2021 дата публикации

AUTOMATIC PROCESS PLATFORM FOR THE PRODUCTION OF ASTATINE-211 [ AT 211] RADIOPHARMACEUTICALS

Номер: US20210139389A1
Автор: Aneheim Emma, Lindegren Sture
Принадлежит:

A system and method for automatic production of astatine-211 labeled molecules is described. The invention represents a significant advantage in the preparation of At-211 radiopharmaceuticals including better reproducibility, reduced production time and increased radiation safety. The invention also enables routine automatic synthesis of radiopharmaceuticals in a clinical setting, in conjunction or at short distance from a cyclotron unit capable of producing the radionuclide. 1. A process for automatic synthesis from isolation of At-211 nuclide from irradiated Bi-209 target material to the full synthesis product of At-211-labeled molecules comprising{'b': 100', '101, 'dry-distilling At-211 in a furnace system (, ) and'}{'b': '109', 'introducing At-211 into a reaction vial () comprising of a precursor molecule adapted to bind At-211, characterized in that the process comprises the steps of'}{'b': '106', 'A) condensing the dry-distilled At-211 by cooling in a cooling unit () to obtain At-211 as a dry residue,'}B) eluting the At-211 with a transfer liquid that solvate the dry residue of At-211,C) introducing At-211 for further chemical processing into said reaction vial,D) activating At-211 for further chemical processing,E) reacting activated At-211 with a precursor molecule,2125. The process of claim 1 , wherein the At-211 is obtained by scraping an irradiated bismuth target to At-211 powder () target material.3120. The process of claim 2 , wherein in the scraping of the irradiated bismuth target is performed using a scraping unit ().4. The process of claim 1 , wherein in step B) the transfer liquid is an organic solvent.5. The process of claim 1 , wherein in step C) the organic solvent is evaporated leaving a dry residue of At-211.6. The process of claim 1 , wherein in step B) the transfer liquid is an adaptive solvent oxidizing At-211.7. The process of claim 1 , wherein in step B) the transfer liquid is an adaptive solvent reducing At-211.8. The process of claim 1 ...

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Номер записи: 54
10-05-2018 дата публикации

TARGET ASSEMBLY AND ISOTOPE PRODUCTION SYSTEM

Номер: US20180130567A1
Автор: Eriksson Tomas, Norling Jonas, Pärnaste Martin
Принадлежит:

Target assembly for an isotope production system. The target assembly includes a target body having a production chamber and a beam cavity that is adjacent to the production chamber. The production chamber is configured to hold a target material. The beam cavity opens to an exterior of the target body and is configured to receive a particle beam that is incident on the production chamber. The target assembly also includes a target sheet positioned to separate the beam cavity and the production chamber. The target sheet has a side that is exposed to the production chamber such that the target sheet is in contact with the target material during isotope production. The target sheet includes graphene. 1. A target assembly for an isotope production system , the target assembly comprising:a target body having a production chamber and a beam cavity that is adjacent to the production chamber, the production chamber configured to hold a target material, the beam cavity opening to an exterior of the target body and being configured to receive a particle beam that is incident on the production chamber; anda target sheet positioned to separate the beam cavity and the production chamber, the target sheet having a side that is exposed to the production chamber such that the target sheet is in contact with the target material during isotope production, wherein the target sheet comprises graphene.2. The target assembly of claim 1 , wherein the target sheet includes a graphene layer that consists essentially of the graphene.3. The target assembly of claim 2 , wherein the target sheet also includes a chamber layer that is stacked with respect to the graphene layer claim 2 , the chamber layer being positioned between the graphene layer and the production chamber and exposed to the production chamber such that the target material is in contact with the chamber layer during isotope production.4. The target assembly of claim 3 , wherein the chamber layer is devoid of a material that ...

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Номер записи: 55
01-09-2022 дата публикации

Nuclear Reactor Assemblies, Nuclear Reactor Target Assemblies, and Nuclear Reactor Methods

Номер: US20220277865A1
Автор: Gates Robert O., Katalenich Jeffrey A., Livingston Jim, Prichard Andrew W., Reid Bruce D., Schmitt Bruce E.
Принадлежит: BATTELLE MEMORIAL INSTITUTE

Reactor target assemblies are provided that can include a housing defining a perimeter of at least one volume and Np or Am spheres within the one volume. Reactor assemblies are provided that can include a reactor vessel and a bundle of target assemblies within the reactor vessel, at least one of the target assemblies comprising a housing defining a volume with Np or Am spheres being within the volume. Irradiation methods are also provided that can include irradiating Np or Am spheres within a nuclear reactor, then removing the irradiated spheres from the reactor and treating the irradiated spheres. 1. A method for producing spheres of Am , Np , or Pu , the method comprising:preparing a nitrate solution comprising Am, Np, or Pu;gelling the Am, Np, or Pu;providing the gelled Am, Np, or Pu through a conduit with another fluid to form gelled Am, Np, or Pu spheres; andwashing and drying gelled spheres to form solid Am, Np, or Pu spheres.2. The method of further comprising forming a mixed feed solution comprising the nitrate solution to gel the Am claim 1 , Np claim 1 , or Pu.3. The method of wherein the mixed feed solution comprises HMTA.4. The method of wherein the mixed feed solution comprises urea.5. The method of further comprising changing the temperature of the mixed feed solution to a point above freezing to gel the Am claim 2 , Np claim 2 , or Pu.6. The method of further comprising providing the gelled Am claim 1 , Np claim 1 , or Pu from the conduit to a forming fluid.7. The method of further comprising removing the forming fluid when washing the gelled spheres.8. The method of further comprising washing the gelled spheres with a solvent.9. The method of wherein the solvent is trichloroethylene and/or isopropyl alcohol.10. The method of wherein the solvent comprises ammonium hydroxide.11. The method of wherein the drying comprises heating the washed spheres.12. The method of wherein claim 11 , upon drying claim 11 , the spheres are rinsed with water and dried to ...

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Номер записи: 56
02-05-2019 дата публикации

METHODS AND DEVICES FOR ISOLATING LEAD 203

Номер: US20190131025A1
Автор: Olewine Keith R.
Принадлежит: Lantheus Medical Imaging, Inc.

Methods for isolating Pb and/or Pb isotopes from various sources are provided. Compositions comprising Pb and/or Pb isotopes free of certain amounts of various contaminants are also provided. 14-. (canceled)5. A composition , comprising{'sup': '203', 'Pb and less than 0.1 μg/mCi Ni and/or'}less than 0.1 μg/mCi Cu and/orless than 0.5 μg/mCi Zn and/orless than 0.25 μg/mCi Fe and/orless than 0.05 μg/mCi Tl.6. The composition of claim 5 , further comprising sodium hydroxide. This application claims the benefit of U.S. Provisional Application No. 61/979,957 filed on Apr. 15, 2014, the entire contents of which are incorporated by reference herein.Radioactive isotopes of many metallic elements have potential uses in the diagnosis and treatment of disease. The lead-203 isotope (Pb), for example, which has a half-life of about 52 hours and decays by electron capture, has excellent promise in medical diagnostics. As a result, recent advances in radioimmunotherapy and peptide targeted radiotherapy have created a great demand for Pb.Pb is an important isotope in certain medical applications. For example, because of its relatively short half-life (˜52 hours) and decay scheme (279 KeV gamma energy, no beta emissions), Pb is particularly suited for imaging based diagnostics and radioimmunotherapeutic applications. With such medical applications, it is important to have a Pb source free of undesirable contaminants. However, Pb is typically generated as a byproduct of Pb and thallium-201 (Tl) production by cyclotrons. As a result, it must be isolated from the cyclotron waste stream, which contains metal contaminants, such as copper, nickel, iron, and zinc. The invention provides efficient means for doing so, based on the surprising discovery that under particular conditions Pb can be eluted almost exclusively from the cyclotron waste stream. As a result, other metal contaminants are left behind, thereby rendering the Pb in a suitable form for its further use in medical and other ...

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Номер записи: 57
23-04-2020 дата публикации

Accelerator-driven neutron activator for brachytherapy

Номер: US20200126683A1
Автор: Luca Maciocco, Stefano Buono
Принадлежит: Advanced Accelerator Applications, Advanced Accelerator Applications SA

A neutron activator for neutron activation of a material, the neutron activator being configured to produce neutrons from an interaction with a proton beam ( 7 ), the neutron activator comprising: a neutron source comprising a metallic target ( 1 ), and a Beryllium first reflector-moderator ( 4 ) peripheral to the neutron source and comprising a neutron activation area ( 10 ) configured to accommodate the neutron source and the material to be activated, the neutron activation area ( 10 ) of the first reflector-moderator ( 4 ) comprising a bore configured to accommodate the neutron source.

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Номер записи: 58
19-05-2016 дата публикации

System and method for producing technetium-99m using existing pet cyclotrons

Номер: US20160141061A1
Автор: Eric A. Burgett
Принадлежит: Individual

The present invention relates generally to a system and method for producing Technetium-99m. More specifically, the present invention relates to a novel method and device for modifying commercially-available, widely-used low energy positron emission tomography (PET) cyclotrons in order to produce Technetium-99m in a more efficient, less expensive manner that previously known.

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Номер записи: 59
18-05-2017 дата публикации

Method for purification of 225ac from irradiated 226ra-targets

Номер: US20170137916A1
Автор: Andreas TÜRLER, Ernst Huenges, Eva Kabai, Josue Manuel Moreno Bermudez, Richard Henkelmann
Принадлежит: Actinium Pharmaceuticals Inc

The present invention describes a method for purification of 225 Ac from irradiated 226 Ra-targets provided on a support comprising a leaching treatment of the 226 Ra-targets for leaching essentially for the entirety of 223 Ac and 226 Ra with nitric or hydrochloric acid, followed by a first extraction chromatography for separating 225 Ac from 226 Ra and other Ra-isotops and a second extraction chromotography for separating 223 Ac from 210 Po and 210 Pb. The finally purified 225 Ac can be used to prepare compositions useful for pharmaceutical purposes.

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Номер записи: 60
17-05-2018 дата публикации

PROCESSES FOR GENERATING GERMANIUM-68 WITH REDUCED VOLATILES

Номер: US20180137947A1
Автор: Owens Mark, Perrigo Melissa, Pipes David, Uhland William
Принадлежит:

Processes for producing germanium-68 from a gallium target body are disclosed. In some embodiments, germanium-68 and gallium are precipitated to remove metal impurities. Germanium-68 and gallium are re-dissolved and loaded onto an ion exchange column to separate germanium-68 from gallium. 1. A process for generating germanium-68 , the process comprising:bombarding a target body including gallium, wherein the bombardment of the target body produces germanium-68 within the target body;stripping the bombarded target body with an acidic mixture to create a stripped solution comprising gallium and geramanium-68; andprecipitating gallium and germanium-68 from a precipitation solution to separate gallium and geramnium-68 from impurity metals in the precipitation solution.2. The method as set forth in further comprising contacting geramanium-68 and gallium with an ion exchange resin to separate germanium-68 from the gallium.3. The process as set forth in or wherein the stripped solution comprises nickel and copper claim 1 , nickel and copper not precipitating with gallium and germanium-68 to separate nickel and copper from gallium and geramnium-68.4. The process as set forth in or wherein gallium and germanium-68 are precipitated by adding ammonium hydroxide to the stripped solution claim 1 , germanium-68 forming a hydroxide that is precipitated from the precipitation solution and gallium forming a hydroxide that is precipitated from the precipitation solution.5. The process as set forth in or wherein the acidic mixture comprises nitric acid.6. The process as set forth in or wherein the acidic mixture comprises copper nitrate.7. The process as set forth in or wherein the acidic mixture does not contain a halide.8. The process as set forth in or wherein germanium-68 and gallium are dissolved in an ion exchange feed solution before contacting germanium-68 and gallium with the cation exchange resin claim 1 , the ion exchange feed solution not comprising a halide.9. The process ...

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Номер записи: 61
14-08-2014 дата публикации

PRODUCTION OF ACTINIUM-227 AND THORIUM-228 FROM RADIUM-226 TO SUPPLY ALPHA-EMITTING ISOTOPES RADIUM-223, THORIUM-227, RADIUM-224, BISMUTH-212

Номер: US20140226774A1
Автор: Holden Charles S., Schenter Robert E.
Принадлежит: THORENCO MEDICAL ISOTOPES LLC

An actinium-227 production device having a plurality of metallic or ceramic caplets, each enclosing a radium-226 compound in redundantly nested sealed cylinders. The radium-226 compound is compacted into a disk and diluted with heat transporting ceramic materials. A thermal neutron shield including spectrum shaping materials to protect actinium-227 produced from exposure to thermal neutrons is included along with a strong neutron absorber to shape the neutron spectrum such that radium-226 nuclei are exposed to neutrons in the higher epithermal energy groups upon entry into the target with an energy of between 20 eV and 1 KeV. 1. An actinium-227 production device , comprising:a plurality of metallic or ceramic caplets enclosing radium-226-containing material enclosed redundantly by nested sealed cylinders, wherein said radium-226-containing material is radium-226 carbonate or another radium-226 compound compacted into a disk and diluted with heat transporting ceramic materials;a thermal neutron shield including spectrum shaping materials to protect actinium-227 produced from exposure to thermal neutrons; anda strong neutron absorber;wherein the thermal neutron absorber and the strong neutron absorber shape the neutron spectrum such that said radium-226-containing material is exposed to neutrons in the higher epithermal energy groups between 20 eV and 1 KeV.2. The actinium-227 production device of claim 1 , wherein said caplets have a cross-sectional geometric shape selected from the group consisting of disk-shaped claim 1 , hexagonal claim 1 , octagonal claim 1 , square claim 1 , and rectangular.3. The actinium-227 production device of claim 1 , further including an exterior jacket including a radon trapping matrix.4. The actinium-227 production device of claim 3 , wherein said radon trapping matrix is fabricated from a silver exchanged zeolite.5. The actinium-227 production device of claim 3 , wherein said radon trapping matrix is fabricated from metallurgical grade ...

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Номер записи: 62
26-05-2016 дата публикации

PRODUCTIONS OF RADIOISOTOPES

Номер: US20160148712A1
Автор: SANDQUIST GARY M.
Принадлежит:

The present disclosure generally relates to methods and structures for the production of radioisotopes from the thermal neutron irradiation of selected natural isotopes. The methods, structures and operations are applicable to the production of any radioisotope that may be produced from neutron irradiation. 1. A structure for production of Mo-99 via neutron capture comprising:a structural support;a containment fixture configured to maintain a high vacuum and provide electrical isolation of internal components;an irradiation target configured to provide an enhanced neutron irradiation environment of appropriate neutron energy spectrum; anda high surface area target derived from a selection of various configurations, chemical, physical and/or isotopic compositions, placement locations, or designs.2. The structure for production of Mo-99 via neutron capture of claim 1 , wherein the containment fixture is designed claim 1 , configured and operated to sequester molybdenum radioisotopes produced.3. The structure for production of Mo-99 via neutron capture of claim 1 , wherein the high surface area target comprises a composition of shapes claim 1 , sizes claim 1 , and spatial distributions derived from molybdenum.4. The structure for production of Mo-99 via neutron capture of claim 1 , wherein the irradiation target further comprises supporting structures.5. A structure for production of Mo-99 via neutron capture comprising:a high surface area target configured to be inserted into an electrically insulating sleeve;a fixture comprising an electrical conducting layer that electrically couples the high surface area target, wherein the fixture is configured to have an electrically insulating sleeve positioned inside the fixture; andan inner metal sleeve.6. The structure for production of Mo-99 via neutron capture of claim 5 , wherein the high surface area target comprises a material derived from selected molybdenum isotopes.7. The structure for production of Mo-99 via neutron ...

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Номер записи: 63
24-05-2018 дата публикации

PURIFICATION PROCESS

Номер: US20180142326A1
Автор: Barbosa Luis A.M.M.
Принадлежит:

A process for purifying Mo-99 from an acidic solution obtained by dissolving an irradiated solid target comprising uranium in an acidic medium, or from an acidic solution comprising uranium and which has previously been irradiated in a nuclear reactor, or from an acidic solution comprising uranium and which has been used as reactor fuel in a homogeneous reactor, the process comprising contacting the acidic solution with an adsorbent comprising a zirconium oxide, zirconium hydroxide, zirconium alkoxide, zirconium halide and/or zirconium oxide halide, and eluting the Mo-99 from the adsorbent using a solution of a strong base, the eluate then being subjected to a subsequent purification process involving an alkaline-based Mo-99 chromatographic recovery step on an anion exchange material. Also provided is apparatus for carrying out the process. 1. An apparatus for carrying out a process for purifying Mo-99 , the apparatus comprising:a column or vessel containing an adsorbent comprising a zirconium oxide, zirconium hydroxide, zirconium alkoxide, zirconium halide and/or zirconium oxide halide;a source of a solution of a strong base, the source of strong base solution being arranged in fluid communication with the column or vessel containing the first adsorbent; anda column or vessel containing an anion exchange material and arranged in downstream fluid communication with the column or vessel containing the adsorbent.2. An apparatus of claim 1 , wherein the adsorbent further comprises a titanium oxide and/or silicon oxide.3. An apparatus of claim 1 , wherein the zirconium compound is present at a concentration of from 5 to 70 mol % of the adsorbent.4. An apparatus of claim 1 , wherein the adsorbent is in the form of pellets.5. The apparatus of claim 1 , the apparatus further comprisinga source of a solution of an acid, the source of acid solution being arranged in fluid communication with the column or vessel containing the anion exchange material; and{'sub': 2', '2, 'a ...

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Номер записи: 64
09-05-2019 дата публикации

Targetry coupled separations

Номер: US20190139665A1
Автор: Joshua C. Walter, Ken Czerwinski
Принадлежит: TerraPower LLC

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

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Номер записи: 65
21-08-2014 дата публикации

Extraction Process

Номер: US20140234186A1
Автор: Barbosa Luis A.M.M.
Принадлежит:

A process for extracting Cs-137 from i) an acidic solution obtained by dissolving an irradiated solid target comprising uranium, ii) an acidic solution comprising uranium which has previously been irradiated in a nuclear reactor, or iii) an acidic solution comprising uranium which has been used as reactor fuel in a homogeneous reactor, the acidic solution i), ii) or iii) having been treated to harvest Mo-99, wherein the process comprises contacting the treated acidic solution with an adsorbent comprising ammonium molybdophosphate (AMP). In an embodiment, the AMP is combined with an organic or inorganic polymeric support, for example AMP synthesised within hollow aluminosilicate microspheres (AMP-C). 1. A process for extracting Cs-137 from i) an acidic solution obtained by dissolving an irradiated solid target comprising uranium , ii) an acidic solution comprising uranium which has previously been irradiated in a nuclear reactor , or iii) an acidic solution comprising uranium which has been used as reactor fuel in a homogeneous reactor , the acidic solution i) , ii) or iii) having been treated to harvest Mo-99 , the process comprising contacting the treated acidic solution with an adsorbent comprising ammonium molybdophosphate (AMP).2. A process according to claim 1 , wherein the acidic solution i) is obtained by dissolving an irradiated solid target comprising uranium in an acidic medium.3. A process according to claim 1 , wherein the AMP is combined with an organic or inorganic polymeric support.4. A process according to claim 3 , wherein the AMP combined with an inorganic polymeric support is AMP synthesised within hollow aluminosilicate microspheres (AMP-C).5. A process according to claim 1 , wherein the process further comprises a step of harvesting Mo-99 from the solution i) claim 1 , ii) or iii) prior to contacting said solution with the adsorbent.6. A process according to wherein Cs-137 is extracted from an acidic solution comprising uranium which has ...

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Номер записи: 66
07-05-2020 дата публикации

SYSTEMS AND METHODS FOR PREPARING TAILORED RADIOACTIVE ISOTOPE SOLUTIONS

Номер: US20200143953A1
Автор: Drera Saleem S.
Принадлежит:

The present disclosure relates to systems and methods for producing tailored solutions or medicaments containing radioactive isotopes (e.g., alpha particle emitting radioactive isotopes). The solutions may be produced by appropriate aging and separation steps. Therapeutically effective amounts of Pb-212 and/or Bi-213 may thus be obtained. 1. A method comprising: '(i) radioactively decaying at least some of the Th-232 cations, thereby producing an aged starting actinide element solution having at least some progeny divalent cations, wherein the progeny divalent include one or more of Ra-228, Ra-224, Pb-212, and Pb-208; and', '(a) aging a starting actinide element solution comprising Th-232 cations, wherein the aging comprises(b) flowing the aged starting actinide element solution through a column having an adsorbent, thereby adsorbing, by the adsorbent, at least some of the progeny divalent cations of the aged actinide element solution.2. The method of claim 1 , comprising:after the flowing, contacting the adsorbent with an extraction solution, thereby desorbing at least some of the progeny divalent cations from the adsorbent; andrecovering an extraction effluent solution, wherein the extraction effluent solution comprises at least some of the extraction solution and at least some progeny divalent cations, wherein at least 75% of the adsorbed progeny divalent cations are recovered in the extraction effluent solution.3. The method of claim 2 , comprising: radioactively decaying at least some of the progeny divalent cations, thereby producing an aged extraction effluent solution comprising at least some Ac-228 cations and at least some Th-228 cations;', 'wherein the aged extraction effluent solution comprises at least some of the progeny divalent cations., 'aging the extraction effluent solution, wherein the aging the extraction effluent solution comprises4. The method of claim 3 , wherein the adsorbent is a first adsorbent claim 3 , the method comprising: 'wherein the ...

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Номер записи: 67
24-06-2021 дата публикации

SYSTEMS, METHODS AND DEVICES FOR PRODUCING, MANUFACTURING AND CONTROL OF RADIOPHARMACEUTICALS

Номер: US20210187467A1
Автор: Alcaide Derrick, Benecke Herman, Bouton Chad E., Danhof Scott N., ESHIMA Dennis, Garbark Daniel B., Gleeson James B., Hassenpflug Eric, HOLLEY David A., HUSNU Mehmet, KELLEY Brian C., KLAUSING Thomas A., Padgett Henry, Prescott James A., Stone James, STROUP Jeffrey T., ZAMBANINI Joseph E.
Принадлежит:

Systems, methods, and devices for generating radionuclides for use in production of radiopharmaceuticals; synthesizing the radionuclides generated and removing any unwanted products; measuring the quantity and activity level of the synthesized radionuclides; distributively delivering the radionuclides in appropriate quantities to modular cassette synthesis units in a modular cassette subsystem for contemporaneous/parallel production of radiopharmaceutical output and that allow reuse and/or quick, safe, and disposable replacement of portions of the subsystem; delivering non-radionuclide components to the modular cassette synthesis units as part of production of radiopharmaceutical output; measuring the quantity and activity level of each stream of radiopharmaceutical output; purifying the radiopharmaceutical output; dispensing individual doses in sterile vial(s); automatically producing labeling and dose related information; performing automated quality control on extracted samples of produced radiopharmaceutical output; and providing software and hardware controls for overall and sub-portion operation for optional remote data collection, communication, and/or control. 173-. (canceled)74. A radiopharmaceutical production system comprising:a cyclotron configured to produce a radionuclide from a precursor;a synthesis unit configured to produce at least one dose of a radiopharmaceutical comprising the radionuclide and one or more pharmaceutical reagents, wherein the synthesis unit comprises at least one modular, disposable cassette unit configured to receive the radionuclide from the cyclotron and the one or more pharmaceutical reagents from a reagent pack;a vial filler configured to dispense the at least one dose of the radiopharmaceutical produced by the synthesis unit into a vial;a quality control unit configured to perform automated quality control on a sample of the at least one dose of the radiopharmaceutical; anda control platform configured to receive input from ...

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Номер записи: 68
07-06-2018 дата публикации

METHOD AND SYSTEM FOR PRODUCING GALLIUM-68 RADIOISOTOPE BY SOLID TARGETING IN A CYCLOTRON

Номер: US20180158559A1
Автор: Abbasi Ali A., Easwaramoorthy Balu
Принадлежит:

In a system and a method for making carrier-free radioactive isotopic Gallium-68, stable enriched Zinc-68 is formed into a solid target of very high purity. The solid target of enriched Zinc-68 is exposed to a proton beam provided by irradiation in a cyclotron to change the enriched Zinc-68 into Gallium-68. After irradiation, the solid target contains high concentrations of Gallium-68 with only trace amounts of enriched Zinc-68 and isotopic Gallium-67. Gallium-68 is then further purified to remove the impurities resulting in a Gallium-68 composition with high purity and specific activity and without Germanium-68, Also provided are radiopharmaceutical agents that are labeled with the Gallium-68 compositions made by solid targeting in a cyclotron. 1. A method of making carrier free radioactive isotope Gallium-68 , the method comprising:irradiating a solid target of substantially pure enriched Zinc-68 with a proton beam provided by a cyclotron to produce Gallium-68.2. The method according to claim 1 , wherein the solid target is 99% enriched Zinc-68.3. The method according to claim 1 , wherein the solid target is a foil.4. The method according to claim 1 , wherein the solid target is about 0.05 to about 1.0 mm thick.5. The method according to claim 1 , wherein the solid target has a molar content of enriched Zinc-68 that is about 0.01 to about 1.0 mmol.6. The method according to claim 1 , wherein the proton beam has an intensity of about 10 to about 16 MeV.7. The method according to claim 1 , wherein the solid target is irradiated for about 1 to about 2 hours.8. The method according to claim 1 , wherein the proton beam is directed at the solid target with an angle of incidence of about 10 to about 90 degrees.9. The method of claim 1 , further comprising:dissolving the irradiated solid target in a dissolving acid;isolating Gallium-68 from the dissolved solid target;washing with at least one washing solution; andrecovering purified Gallium-68.10. The method of claim 9 , ...

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Номер записи: 69
22-09-2022 дата публикации

METHOD FOR PRODUCING 225ACTINIUM FROM 226RADIUM

Номер: US20220301735A1
Автор: BANIK Nidhu Lal, CACIUFFO Roberto, MALMBECK Rikard
Принадлежит: THE EUROPEAN UNION, REPRESENTED BY THE EUROPEAN COMMISSION

actinium is produced from radium by irradiating a liquid radium target by means of protons, deuterons or gamma irradiation in an irradiation device () and by extracting the produced actinium out of the irradiated liquid target solution in a first extraction device (). The liquid target solution from which the actinium has been removed is then irradiated again to produce further actinium therein. The liquid target solution is preferably circulated, in a closed loop (), over the irradiation device and in a further closed loop () over the first extraction device (). An advantage of such a method is that the irradiated target solution does not need to be dried and re-dissolved to be able to separate the produced actinium from the radium and no further drying and re-dissolving step is needed for producing the liquid target again starting from the separated radium. The radium target can thus be recycled in a more efficient and safer way, especially in view of the radon gas which is continuously produced by the decay of radium. 2425. The method according to claim 1 , wherein said liquid target solution is circulated during said irradiation step in a first closed loop () over said irradiation device () and over a heat exchanger ().376. The method according to claim 1 , wherein said liquid target solution is circulated during said first extraction step in a second closed loop () over said first extraction device ().4412716. The method according to claim 2 , wherein said liquid target solution is circulated during said irradiation step claim 2 , in said first closed loop () claim 2 , over a container () and said irradiation device () and during said first extraction step claim 2 , in said second closed loop () claim 2 , over said container () and said first extraction device ().5. The method according to claim 1 , wherein said liquid target solution is irradiated for less than 16 days claim 1 , before at least part of said actinium is extracted from the liquid target solution ...

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Номер записи: 70
22-09-2022 дата публикации

System, Emanation Generator, and Process for Production of High-Purity Therapeutic Radioisotopes

Номер: US20220301737A1
Автор: OHara Matthew J.
Принадлежит: BATTELLE MEMORIAL INSTITUTE

An isotope production system, emanation generator, and process are disclosed for production of high-purity radioisotopes. In one implementation example, high-purity Pb-212 and/or Bi-212 isotopes are produced suitable for therapeutic applications. In one embodiment the process includes transporting gaseous radon-220 from a radium-224 bearing generator which provides gas-phase separation of the Rn-220 from the Ra-224 in the generator. Subsequent decay of the captured Rn-220 accumulates high-purity Pb-212 and/or Bi-212 isotopes suitable for direct therapeutic applications. Other high-purity product isotopes may also be prepared. 1. An emanation system for production of ultrapure radioisotopes , comprising:an emanation device having an emanation source comprising a source isotope therein that emanates a radioactive gas therefrom; anda collection device configured to collect the radioactive gas retaining same for a time sufficient to yield one or more high purity radioactive daughter isotopes therein.2. The system of wherein the source isotope is selected from Thorium-228 and/or Radium 224; Thorium-227 and/or Radium-223; or Thorium-230 and/or Radium-226.3. The system of wherein the radioactive gas is selected from Radon-220; Radon-219; and Radon-222.4. The system of wherein the radioactive gas is a radioactive noble gas.5. The system of wherein the source isotope is disposed on a particle surface or a permeable support.6. The system of wherein the source isotope is disposed on magnetic or paramagnetic metal oxide particles.7. The system of wherein the source isotope is disposed on a gas-permeable support.8. The system of wherein the collection device includes a cooling device configured to cool the radioactive gas emanated from the emanation device.9. The system of wherein the collection device includes a soluble salt configured as a thin film or a packed salt claim 1 , or a lipophilic liquid configured as a thin film or a thin film coating on a solid support to extract ...

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Номер записи: 71
14-05-2020 дата публикации

Nuclear Microbattery

Номер: US20200152344A1
Автор: Barnett Anna Megan, BUTERA Silvia, LIOLIOU Grammatiki
Принадлежит:

A nuclear microbattery is disclosed comprising: a radioactive material that emits photons or particles; and at least one diode comprising a semiconductor material arranged to receive and absorb photons or particles and generate electrical charge-carriers in response thereto, wherein said semiconductor material is a crystalline lattice structure comprising Aluminium, Indium and Phosphorus. 1. A nuclear microbattery comprising:a radioactive material that emits photons or particles; andat least one diode comprising a semiconductor material arranged to receive and absorb photons or particles and generate electrical charge-carriers in response thereto, wherein said semiconductor material is a crystalline lattice structure comprising Aluminium, Indium and Phosphorus.2. The microbattery of claim 1 , wherein the crystalline lattice structure is AlInP.3. The microbattery of claim 2 , wherein the lattice structure has a lattice composition AlInP or AlInP.4. The microbattery of claim 1 , comprising electrodes for collecting an electrical current generated by said at least one diode due to the generation of said electrical charge-carriers.5. The microbattery of claim 1 , wherein said at least one diode comprises a plurality of the diodes electrically connected in parallel or series.6. The microbattery of claim 1 , wherein said at least one diode comprises a plurality of the diodes claim 1 , wherein one or more of the diodes is arranged on a first side of the radioactive material claim 1 , and wherein one or more of the diodes is arranged on a second claim 1 , opposite side of the radioactive material.7. The microbattery of claim 1 , wherein a gap is arranged between said at least one diode and the radioactive material claim 1 , and wherein the gap is filled with a gas maintained at sub-atmospheric pressure.8. The microbattery of claim 1 , wherein a gap is arranged between said at least one diode and the radioactive material claim 1 , and wherein the gap is filled with a noble ...

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Номер записи: 72
15-06-2017 дата публикации

METHOD FOR PRODUCING BETA EMITTING RADIOPHARMACEUTICALS, AND BETA EMITTING RADIOPHARMACEUTICALS THUS OBTAINED

Номер: US20170169908A1
Автор: ANDRIGHETTO Alberto
Принадлежит: ISTITUTO NAZIONALE DI FISICA NUCLEARE

The present invention relates to a method for producing beta emitting radiopharmaceuticals. The method provides to produce, through a primary accelerator, a low energy proton beam, namely with an energy lower than 70 MeV, preferably with an energy ranging from 32 to 45 MeV, more preferably with energy ranging from 38 to 42 MeV; the low energy proton beam is irradiated on a source target so as to generate a neutral atom beam; the neutral atoms are ionized, extracted by acceleration and preferably subjected to a first focusing; the first focused beam is subjected to a mass separation such to generate a isobaric beam of radioisotopes. The isobaric beam therefore is preferably subjected to a second focusing and it is sent for a predetermined time on a deposition target. Then the irradiated deposition target is subjected to chemical treatment so as to obtain pure beta emitting radiopharmaceuticals. 113-. (canceled)14. A method for producing pure beta emitting radiopharmaceuticals by pure nuclear fission processes comprising the steps of:i. producing a low energy proton beam through a primary accelerator;ii. irradiating said low energy proton beam on a source target so as to generate, thanks to the reaction of pure nuclear fission, a neutral atom beam;iii. subjecting said neutral atom beam to a positive ionization in an ionizer so as to produce a ionized radioisotope beam;iv. accelerating said ionized radioisotope beam in an accelerator extractor so as to produce an accelerated ionized isotope beam;v. separating said accelerated ionized isotope beam in a mass separator so as to generate an isobaric beam of radioisotopes;vi. irradiating said isobaric beam of radioisotopes on a deposition target for a predetermined time so as to produce an irradiated deposition target; andvii. after said predetermined time, sending said irradiated deposition target to an extraction and purification chemical treatment in a chemical unit so as to obtain pure beta emitting radiopharmaceutical ...

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Номер записи: 73
29-09-2022 дата публикации

TARGETRY COUPLED SEPARATIONS

Номер: US20220310281A1
Автор: Czerwinski Ken, Dunckley Christopher P., Walter Joshua C.
Принадлежит: TerraPower, LLC

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art. 120-. (canceled)21. A method of manufacturing a radionuclide metal that is a fission product of the fissioning of uranium , the method comprising:fissioning a molten fuel salt containing uranium, thereby generating an irradiated fuel salt mixture containing molten uranium salt and fission products, wherein the fission products include the radionuclide metal;contacting at least some of the irradiated fuel salt mixture with supercritical carbon dioxide containing a ligand that forms a metal complex with the radionuclide metal but does not form a metal complex with the uranium in the fuel salt, thereby forming a combined fuel salt and supercritical carbon dioxide mixture containing an amount of radionuclide metal complexes;separating at least some of the radionuclide metal complexes from the combined fuel salt and supercritical carbon dioxide mixture; anddecomposing the radionuclide metal complexes to obtain the radionuclide metal.22. The method of claim 21 , wherein the contacting further comprises:removing the irradiated fuel salt mixture from a molten ...

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Номер записи: 74
01-07-2021 дата публикации

Method for the manufacture of highly purified 68Ge material for radiopharmaceutical purposes

Номер: US20210198116A1
Автор: Harfensteller Mark, Jernström Jussi, Kelmendi Nevzat, Zhernosekov Konstantin
Принадлежит:

A method for the manufacture of highly purified Ge material for radiopharmaceutical purposes. The invention particularly concerns the production of Ge-API (API=Active Pharmaceutical Ingredient) solution complying with the Guidelines for good manufacturing practices (GMP). Starting material for the method of the present invention can be a Ge stock solution of commercial or other origin as raw material. Such Ge containing raw solutions are purified from potential metal and organic impurities originating from production processes. The radiochemical method disclosed is based on a twofold separation of Ge from organic and metallic impurities with two different adsorbent materials. During the first separation phase Ge is purified from both organic and metallic impurities by adsorption in germanium tetrachloride form, after which hydrolyzed Ge is purified from remaining metallic impurities by cation exchange. The final Ge-API-product e.g. fulfills the regulatory requirements for specifications of the GMP production of Ge/Ga generators. 110-. (canceled)11. A method for the manufacture of highly purified Ge material for radiopharmaceutical purposes ,characterized by{'sup': 68', '68', '68, 'sub': '4', '(a) adjusting a Ge-containing solution containing organic and metallic impurities to a HCl concentration of 6.5 to 12 M in order to convert the Ge contained in the solution, to a GeCl-containing material;'}{'sup': '68', 'sub': '4', '(b) loading the solution comprising the GeCl-containing material obtained in step (a) to a resin matrix, wherein said resin matrix is a hydrophilic, macroporous, acrylic ester polymeric resin;'}{'sup': 68', '68', '68', '68, 'sub': 4', '4', '4, '(c) eluting said resin matrix with water in order to hydrolyze the GeCl-containing material and to release Ge essentially in germanic acid form [Ge(OH)] from the GeCl-containing material which was adsorbed to the resin matrix in step (b);'}{'sup': '68', '(d) adjusting an eluate solution containing Ge obtained ...

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Номер записи: 75
25-06-2015 дата публикации

System and method for generating molybdenum-99 and metastable technetium-99, and other isotopes

Номер: US20150179290A1
Автор: Clayton James E.
Принадлежит:

An accelerator based systems are disclosed for the generation of isotopes, such as molybdenum-98 (“99Mo”) and metastable technetium-99 (“99mTc”) from molybdenum-98 (“98Mo”). Multilayer targets are disclosed for use in the system and other systems to generate 99mTc and 98Mo, and other isotopes. In one example a multilayer target comprises a first, inner target of 98Mo surrounded, at least in part, by a separate, second outer layer of 98Mo. In another example, a first target layer of molybdenum-100 is surrounded, at least in part, by a second target layer of 98Mo. In another example, a first inner target comprises a Bremsstrahlung target material surrounded, at least in part, by a second target layer of molybdenum-100, surrounded, at least in part, by a third target layer of 98Mo. 1. A system for generating isotopes , comprising:an accelerator;a source of charged particles coupled to the accelerator to inject charged particles into the accelerator;a target comprising;a first, inner target material, comprising a first isotope of a first material; and a second, outer target material comprising a second isotope of a second material, the second outer target material at least partially surrounding the first, inner target material, the second, outer target material defining a passage for accelerated charged particles to the first, inner target material.2. The system of claim 1 , wherein the first material and the second material are the same and the first isotope and the second isotope are different isotopes of the first material.3. The system of claim 2 , wherein the first claim 2 , inner target material and the second claim 2 , outer target material are separated by a gap.4. The system of claim 3 , wherein the first isotope and the second isotope each comprise molybdenum-98.5. The system of claim 2 , wherein the first isotope comprises molybdenum-100 and the second isotope comprises molybdenum-98.6. The system of claim 1 , wherein the target further comprises a layer of ...

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Номер записи: 76
21-05-2020 дата публикации

Targetry coupled separations

Номер: US20200161015A1
Автор: Joshua C. Walter, Ken Czerwinski
Принадлежит: TerraPower LLC

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

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Номер записи: 77
18-09-2014 дата публикации

PROCESS AND APPARATUS FOR CONDENSATION REPRESSING ISOTOPE SEPARATION BY LASER ACTIVATION

Номер: US20140270035A1
Автор: Eerkens Jozef W.
Принадлежит:

Isotope enrichment by laser activation wherein a multi-isotopic element Q, like Uranium, Silicon, Carbon is incorporated into gaseous QF, QF, QF, QOF, etc and diluted in gas G like He, N, Ar, Xe, SFor other inert gas; and wherein that mixture is cooled by adiabatic expansion or other means encouraging formation of dimers QF:G in a supersonic super-cooled free jet; and wherein that jet is exposed to laser photons at wavelengths that selectively excite predetermined molecules QFto QF*, thereby inducing rapid VT conversions and dissociations of QF*:G→QF+G+kT, while leaving non-excited dimers QF:G intact; and wherein a skimmer separates the supersonic free-jet core stream containing heavier QF:G dimers from lighter core-escaped QF-enriched rim gases. Particularly an advanced technique is disclosed to enrich UFby free jet expansion and isotope-selective dimerization suppression, utilizing a molecular CO laser and intra-cavity UFirradiation with laser lines overlapping predetermined UFabsorptions; and providing multiple free jet separator units irradiated by one laser beam, thereby enhancing process economics. 1. A process for enriching a selected isotope of uranium from a mixture of gaseous Uisotopomers in a supersonic low-pressure flow chamber , comprising the steps of:{'sub': '6', 'a. super-cooling a supersonic free jet comprising UFand carrier gases by adiabatically expansion of said supersonic free jet in a flow chamber, said supersonic free jet comprising a supersonic core region surrounded radially by a barrel shock and rim gases;'}{'sub': 6', '6, 'b. selectively exciting UFisotopomers containing a selected isotope of UFusing photons;'}c. separating the supersonic core region of the supersonic free jet from said rim gases with a skimmer.2. The process of claim 1 , wherein the supersonic free jet comprises a mixture of carrier gases G and UFin a predetermined UF/G molecular ratio.3. The process of claim 1 , wherein the selected uranium isotope is one of U-232 claim ...

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Номер записи: 78
28-06-2018 дата публикации

METHOD FOR THE QUANTIFICATION OF 227AC IN 223RA COMPOSITIONS

Номер: US20180180585A1
Автор: HJELLUM Gro Elisabeth
Принадлежит:

A method for the quantification of Ac in a Ra composition comprising passing the composition through a first solid phase extraction column A, wherein said column comprises a thorium specific resin, passing the eluate of column A through a second solid phase extraction Column B, wherein said column comprises an actinium specific resin and recovering the Ac absorbed onto the resin in column B and determining the amount thereof. 121-. (canceled)22. Apparatus for the quantification of Ac in a Ra composition comprisinga first solid phase extraction column A, wherein said column A comprises a thorium specific resin, anda second solid phase extraction column B, wherein said column B comprises an actinium specific resin.23. The apparatus of claim 22 , wherein column A and column B are arranged in series.24. The apparatus of claim 22 , wherein the thorium specific resin comprises a phosphonate extractant.25. The apparatus of claim 24 , wherein the phosphonate extractant is an alkyl phosphonate extractant.27. The apparatus of claim 26 , wherein the dialkyl alkyl phosphonate extractant is a dipentyl pentylphosphonate extractant29. The apparatus of claim 28 , wherein the tetra-alkyl diglycolamide extractant is a N claim 28 ,N claim 28 ,N′ claim 28 ,N′-tetra-n-octyldiglycolamide (DGA) extractant. The present invention relates to a novel method for quantifying levels of Ac in Ra compositions, in particular a method which involves solid phase extraction followed by quantification via the in-growth of the Th daughter via γ-spectrometry. The invention further relates to the use of the method of the invention in determining the level of Ac in a Ra composition and to an apparatus for use in the method of the invention.A substantial percentage of cancer patients is effected by skeletal metastases. As many as 85% of patients with advanced lung, prostate and breast carcinoma develop bony metastates (Garret 1993, Nielsen et al, 1991). They are associated with a decline in health and ...

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Номер записи: 79
08-07-2021 дата публикации

RADIONUCLIDE PREPARATION SYSTEM, STORAGE MEDIUM READABLE BY COMPUTER STORING RADIONUCLIDE PREPARATION PROGRAM, RADIONUCLIDE PREPARATION METHOD, AND TERMINAL DEVICE

Номер: US20210210244A1
Автор: ICHIMURA Soichiro, KANDA Akimitsu, SHINOHARA Atsushi, TOYOSHIMA Atsushi, ZHANG Zijian
Принадлежит: OSAKA UNIVERSITY

There is provided a radionuclide manufacturing system, a computer-readable storage medium storing a radionuclide manufacturing program, a radionuclide manufacturing method, and a terminal device for more stably manufacturing a radionuclide. 1. A radionuclide manufacturing system comprising:a heating unit including a first end into which carrier gas is introduced and a second end from which the carrier gas is discharged, the heating unit being configured to internally house a target holding a radionuclide;a gas supply unit including a first end connected to a gas retention unit that retains the carrier gas, and a second end connected to the first end of the heating unit;an adsorption unit including a first end connected to the second end of the heating unit and introducing the carrier gas, and a second end from which the carrier gas is discharged, the adsorption unit being configured to adsorb the radionuclide;a solvent supply unit including an end connected to the second end of the adsorption unit;a storage unit configured to store a predetermined instruction; anda control unit configured to control the heating unit to heat the target at a temperature at which the radionuclide held in the target is allowed to volatilize, to control the gas supply unit to supply the carrier gas to the heating unit in order to transport the radionuclide volatilized in the heating unit to the adsorption unit, and to control the solvent supply unit in order to supply a solvent for eluting the radionuclide adsorbed to the adsorption unit to the adsorption unit based on the instruction.2. The radionuclide manufacturing system according to claim 1 , further comprising a warming unit disposed to cover a part of the adsorption unit and configured to warm the radionuclide transported by the carrier gas.3. The radionuclide manufacturing system according to claim 2 , wherein the control unit is configured to control the warming unit to warm the part covered with the warming unit to a ...

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Номер записи: 80
08-07-2021 дата публикации

METHOD FOR PRODUCING AC-225 FROM RA-226

Номер: US20210210245A1
Автор: BEYER Gerd-Jürgen, COMOR Jozef, GEETS Jean-Michel
Принадлежит: Ion Beam Applications

The embodiments of the present disclosure provide a method for producing Ac-225 from Ra-226, comprising submitting Ra-226 to a photo-nuclear process, collecting an electrochemical precipitation of an Ac-225 on a cathode in a recipient, removing the cathode from the recipient after the electrochemical precipitation of the Ac-225, transferring the cathode to a hot cell environment, and extracting the Ac-225 from the cathode in the hot cell environment. The Ra-226 may comprise a liquid solution in the recipient, and submitting Ra-226 to the photo-nuclear process may comprise irradiating the Ra-226 to produce Ra-225. The Ra-225 may decay into Ac-225 upon irradiation of the Ra-226. 1. A method for producing Ac-225 from Ra-226 , the method comprising:submitting Ra-226 to a photo-nuclear process, wherein the Ra-226 comprises a liquid solution in a recipient;irradiating the Ra-226 to produce Ra-225, wherein the Ra-225 decays into Ac-225 upon irradiation of the Ra-226;collecting an electrochemical precipitation of the Ac-225 on a cathode in the recipient;removing the cathode from the recipient after the electrochemical precipitation of the Ac-225;transferring the cathode to a hot cell environment; andextracting the Ac-225 from the cathode in the hot cell environment.2. The method as claimed in claim 1 , wherein the Ra-226 is irradiated with photons having an energy of at least 6.4 MeV claim 1 , and wherein the energy is generated as Bremsstrahlung from an e-beam having an energy higher than 6.4 MeV.3. The method as claimed in claim 1 , wherein the cathode is made of a chemically inert metal.4. The method as claimed in claim 1 , wherein the Ac-225 extracted from the cathode is chemically purified to remove traces of coprecipitated Ra-226 and Ra-225.5. The method as claimed in claim 1 , wherein the liquid solution comprises a nitric acid (HNO) solution with a pH below 6.6. The method as claimed in claim 1 , wherein the recipient comprises a first recipient and a third ...

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Номер записи: 81
30-06-2016 дата публикации

Targetry coupled separations

Номер: US20160189816A1
Автор: Czerwinski Ken, Walter Joshua C.
Принадлежит: TerraPower, LLC

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art. 1. A method for manufacturing Mo radioisotope , the method comprising:providing a source containing a first mass of uranium, the source being in a form in which a majority of uranium atoms are within a selected distance from an available surface of the source;{'sup': 99', '99, 'exposing the source to neutrons, thereby reducing the first mass of uranium in the source to a second mass of uranium less than the first mass and creating at least some atoms of the Mo radioisotope and thereby also causing at least some of the newly created atoms of the Mo radioisotope to move toward an available surface of the source; and'}{'sup': '99', 'after exposing the source to neutrons, removing at least some of the atoms of the Mo radioisotope from the source without substantially removing uranium from the second mass of uranium in the source.'}2. The method of wherein the removing operation removes less than 0.01% of the uranium from the second mass of uranium in the source.3. The method of wherein the removing operation removes less than 0.1% of the uranium from the ...

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Номер записи: 82
22-07-2021 дата публикации

SYSTEM AND METHOD FOR GERMANIUM-68 ISOTOPE PRODUCTION

Номер: US20210225546A1
Автор: WHYTE Raymond Jeffery, Wong Chung Kay Michael
Принадлежит:

A system and method for producing Germanium-68 (Ge-68) isotopes is provided. The method includes irradiating a plated or encapsulated target containing Gallium, dissolving the irradiated target with an acid, and purifying the dissolved target by distillation to produce purified Ge-68. A dissolution cell assembly is provided for use in the dissolution step of the method. 1. A method for producing Germanium-68 isotopes , the method comprising:irradiating a solid target plated with a Ga—Ni alloy to form an irradiated Ga—Ni alloy plated target,dissolving the irradiated Ga—Ni alloy plated solid target with an acid, andpurifying the dissolved irradiated plated target by distillation to produce purified Germanium-68.2. The method according to claim 1 , wherein the solid target comprises silver.3. The method according to claim 1 , wherein the solid target is plated with an insulation layer.4. The method according to claim 3 , wherein the insulation layer is selected from the group consisting of copper claim 3 , aluminum claim 3 , nickel claim 3 , tungsten claim 3 , silver-copper alloy claim 3 , tungsten-silver alloy claim 3 , rhodium claim 3 , rhodium-gallium alloy claim 3 , niobium claim 3 , and a combination thereof.5. The method according to claim 1 , wherein the acid is Hydrochloric acid (HCl) claim 1 , sulfuric acid claim 1 , nitric acid claim 1 , or a combination of other strong acids.6. The method according to claim 1 , wherein dissolving occurs at a temperature in a range of 70 degrees C. to 80 degrees C.7. The method according to claim 1 , further comprising condensing vapor produced during dissolution.8. The method according to claim 1 , further comprising increasing the temperature after dissolution.9. The method according to claim 8 , wherein the temperature is increased to 90 degrees C. to 100 degrees C.10. The method according to claim 7 , wherein the vapor is condensed at a distillation condenser.11. The method according to claim 1 , further comprising ...

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Номер записи: 83
13-07-2017 дата публикации

DEVICE AND METHOD FOR THE PRODUCTION OF RADIOISOTOPES

Номер: US20170200520A1
Автор: Wilson Taylor Ramon
Принадлежит:

A dense plasma focus (DPF) to produce positron emitters is provided, where a pulsed device has an anode and a cathode arranged in a vacuum chamber, the anode and cathode being subjected to a high voltage. When the vacuum chamber is filled with a reaction gas and a high voltage generated is applied, a plasma sheath is created and a reaction between the electrodes take place to produce plasmoids resulting in an ion beam that interacts with a reactive gas to produce radio-isotopes. 1. A device for producing isotopes , the device comprising:a first chamber including an anode and at least one accelerating gas;a second chamber including at least one target gas or target liquid; anda voltage source configured to apply a voltage between the anode and the first chamber; whereina reaction of the accelerated gas is produced in the first chamber as a result of the applied voltage, the reaction resulting in a plasma; anda nuclear reaction between the plasma and the target gas is produced in the second chamber.2. The device of claim 1 , wherein the nuclear reaction results in a production of one or more isotopes.3. The device of claim 1 , whereina beam window separates the first chamber and the second chamber; andthe plasma travels from the first chamber to the second chamber through the beam window.4. The device of claim 3 , wherein the beam window comprises Beryllium.5. The device of claim 1 , wherein the second chamber includes conduits to insert or remove components of the nuclear reaction without disturbing the first chamber.6. The device of claim 1 , wherein the anode is an elongated hollow cylinder.7. The device of claim 6 , wherein the anode is covered with a thermal and electrical insulator.8. The device of claim 7 , wherein the thermal and electrical insulator comprises a glass layer.9. The device of claim 6 , whereinthe elongated anode includes a recess at a first end opposite to a second end that is coupled to the voltage source; andthe plasma is created at the first ...

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Номер записи: 84
13-07-2017 дата публикации

APPARATUS AND METHODS FOR TRANSMUTATION OF ELEMENTS

Номер: US20170200521A1
Автор: Dent, JR. William Vaden
Принадлежит:

Examples of apparatus and methods for transmutation of an element are disclosed. An apparatus can include a neutron emitter configured to emit neutrons with a neutron output, a neutron moderator configured to reduce the average energy of the neutron output to produce a moderated neutron output, a target configured to absorb neutrons when exposed to the moderated neutron output, the absorption of the neutrons by the target producing a transmuted element, and an extractor configured to extract the desired element. A method can include producing a neutron output, reducing the average energy of the neutron output with a neutron moderator to produce a moderated neutron output, absorbing neutrons from the moderated neutron output with the target to generate a transmuted element, and eluting a solution through the target to extract a desired element. In some examples, the target includes molybdenum-98, and the desired element includes technetium-99m. 1. (canceled)2. A method of transmutating a target comprising powdered molybdenum dioxide , the method comprising:producing a neutron output;reducing an average energy of the neutron output with a neutron moderator to produce a moderated neutron output;absorbing neutrons from the moderated neutron output with the powdered molybdenum dioxide including to generate technetium-99m;flowing an eluting solution comprising saline through the powdered molybdenum dioxide; andextracting technetium-99m from the eluting solution after it has flowed through the powdered molybdenum dioxide.3. The method of claim 2 , further comprising multiplying neutrons in the moderated neutron output to produce a moderated and multiplied neutron output claim 2 , wherein absorbing neutrons from the moderated neutron output comprises absorbing neutrons from the moderated and multiplied neutron output.4. The method of claim 2 , wherein absorbing neutrons from the moderated neutron output with the powdered molybdenum dioxide comprises forming molybdenum-99.5. ...

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Номер записи: 85
30-07-2015 дата публикации

Medical unit for injecting a patient with rubidium 82

Номер: US20150209504A1
Автор: Pierre-Marie Lemer
Принадлежит: Lemer Protection Anti X SAS

A medical unit for injecting a patient with an elution solution containing rubidium 82, includes, elements for receiving a strontium/rubidium generator capable of producing an elution solution that contains the rubidium 82 and that is capable of being contaminated by strontium 82 and/or strontium 85. This medical unit includes: —own elements thereof for acquiring a value related to at least one safety parameter that is associated with a maximum threshold value corresponding to a potentially excessive contamination of the elution solution with strontium 82 and/or strontium 85, and —control elements including safety elements that are controlled in an active configuration when the acquired value reaches the maximum threshold value of the safety parameter, the safety elements being capable, in the active configuration, of operating the infusion elements to the stop position in order to prevent an injection of the elution solution into the patient.

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Номер записи: 86
20-07-2017 дата публикации

HIGH PURITY SN-117M COMPOSITIONS AND METHODS OF PREPARING SAME

Номер: US20170204497A1
Автор: Crump Druce, George St. George, Simon Jaime, Stevenson Nigel R.
Принадлежит:

A method of purifying a high specific activity Sn-117m composition is provided that includes extracting an iodide complex of Sn-117m with an organic solvent from an acidic aqueous cadmium solution comprising a dissolved irradiated cadmium target, an acid, and a source of iodide. The organic solvent layer comprising the iodide complex of Sn-117m is substantially reduced in cadmium content. The Sn-117m may be back extracted into an aqueous solution. 1. A method of purifying a high specific activity Sn-117m composition , comprising:extracting an iodide complex of Sn-117m with an organic solvent from an acidic aqueous cadmium solution, said cadmium solution comprising a dissolved irradiated cadmium target, an acid, and a source of iodide, to provide an organic solvent layer comprising the iodide complex of Sn-117m that is substantially reduced in cadmium content and an acidic cadmium aqueous layer.2. The method of claim 1 , wherein the organic solvent is selected from the group consisting of aromatic ring solvents and halogenated alkane solvents and mixtures thereof.3. The method of claim 1 , wherein the organic solvent comprises toluene.4. The method of claim 1 , wherein source of iodide is selected from hydrogen iodide claim 1 , an iodide salt and mixtures thereof.5. The method of claim 4 , wherein the iodide salt is selected from the group consisting of alkali metal salts and alkaline earth metal salts and mixtures thereof.6. The method of claim 1 , further comprising:dissolving an irradiated cadmium target comprising a quantity of high specific activity Sn-117m to form said cadmium solution.7. The method of claim 1 , further comprising:separating the organic solvent layer comprising the iodide complex of Sn-117m from the acidic cadmium aqueous layer to provide an organic solution enriched in Sn-117m.8. The method of claim 7 , further comprising:back-extracting the Sn-117m into an acidic solution by washing the organic solvent layer with a hydrochloric acid solution ...

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Номер записи: 87
27-06-2019 дата публикации

METHODS FOR PURIFYING MOLYBDENUM-99

Номер: US20190198186A1
Автор: Barbosa Luis Antonio M.M., Dekker Frits, Gerritsen Marjolijn, Hermeler Mark David
Принадлежит:

Methods for purifying the molybdenum-99 isotope are disclosed. Molybdenum-99 is loaded onto an anion exchange column and extracted. In some embodiments, the extraction solution may include nitric acid and nitrate salts. In other embodiments, a two stage elution is performed in which a nitric acid containing eluent and a hydroxide containing eluent are used in succession to extract molybdenum-99. 1. A method for purifying a molybdenum-99 feed stream comprising molybdenum-99 , the method comprising:contacting the molybdenum-99 feed stream with anion exchange media to adsorb molybdenum-99 onto the media; andcontacting the molybdenum-99 adsorbed media with an eluent solution comprising nitric acid and a nitrate salt.2. The method as set forth in wherein the molybdenum-99 feed stream comprises at least about 90 wt % molybdenum-99 on a dry basis.3. The method as set forth in wherein the nitrate salt is selected from the group consisting of alkali or alkaline earth-metal nitrate salts.4. The method as set forth in wherein the nitrate salt is selected from the group consisting of sodium nitrate and potassium nitrate.5. The method as set forth in wherein the molybdenum-99 feed stream comprises molybdenyl-99 ions.6. The method as set forth in wherein the molybdenum-99 feed stream is an aqueous hydroxide solution.7. The method as set forth in wherein the ratio of the mass of the eluent solution to the mass of molybdenum-99 in the molybdenum-99 feed stream is less than about 4:1.8. A method for purifying a molybdenum-99 feed stream comprising molybdenum-99 claim 1 , the method comprising:contacting the molybdenum-99 feed stream with anion exchange media to adsorb molybdenum-99 onto the media;contacting the molybdenum-99 adsorbed media with a nitric acid-containing eluent solution; andcontacting the molybdenum-99 adsorbed media with a hydroxide-containing eluent solution.9. The method as set forth in wherein the molybdenum-99 feed stream comprises at least about 90 wt % ...

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Номер записи: 88
27-06-2019 дата публикации

Rapid Isolation of Cyclotron-Produced Gallium-68

Номер: US20190198187A1
Автор: DeGrado Timothy R., Pandey Mukesh K.
Принадлежит:

Methods for rapid isolation of radionuclides (e.g., Ga) produced using a cyclotron and methods for recycling of the parent isotope (e.g., Zn) are disclosed. In one version of the method, a solution including a radionuclide (e.g., Ga) is created from a target including cations (e.g., Zn). The solution including the radionuclide is passed through a first column including a sorbent comprising a hydroxamate resin to adsorb the radionuclide on the sorbent, and the radionuclide is eluted off the sorbent. The cations (e.g., Zn) are recovered from a recovery solution that has passed through the first column by passing the recovery solution through a second column including a second sorbent comprising a cation exchange resin. 1. A method for producing a solution including a radionuclide , the method comprising:{'sup': '68', '(a) bombarding a target solution with protons to produce a solution including a radionuclide, wherein the radionuclide is Ga;'}(b) passing the solution including the radionuclide through a column including a sorbent to adsorb the radionuclide on the sorbent; and(c) eluting the radionuclide off the sorbent,wherein the sorbent comprises a hydroxamate resin.2. The method of claim 1 , wherein the target solution comprises Zn-enriched zinc nitrate.3. The method of claim 1 , wherein step (c) comprises eluting the radionuclide off the sorbent using hydrochloric acid.4. The method of claim 1 , wherein step (c) comprises eluting the radionuclide off the sorbent with an amount of eluent of 5 milliliters or less.5. The method of claim 1 , wherein the method takes 30 minutes or less.6. The method of claim 1 , wherein step (b) comprises adjusting pH of the solution including the radionuclide before passing the solution including the radionuclide through the column.7. The method of claim 6 , wherein the adjusting of the pH comprises a dilution with water.8. The method of claim 6 , wherein the adjusting of the pH comprises an addition of a base.9. The method of claim 8 ...

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Номер записи: 89
26-07-2018 дата публикации

PRODUCTION OF COPPER-67 FROM AN ENRICHED ZINC-68 TARGET

Номер: US20180209013A1
Автор: Gardner Tim, Stoner Jon
Принадлежит:

An apparatus including a heating element and a sublimation vessel disposed adjacent the heating element such that the heating element heats a portion thereof. A collection vessel is removably disposed within the sublimation vessel and is open on an end thereof. A crucible is configured to sealingly position a solid mixture against the collection vessel. 1. A method , comprising the steps of:positioning a metal target within a target assembly including one or more converter plates positioned adjacent the metal target;directing an electron beam at the one or more converter plates;irradiating the metal target with the electron beam via the one or more converter plates.2. The method of claim 1 , further comprising preparing the metal target prior to the irradiating by melting a metal and pouring the metal into a crucible via a funnel.3. The method of claim 1 , wherein the electron beam has an energy of at least 20 MeV and a power of at least 1 kW.4. The method of claim 1 , further comprising cooling the one or more converter plates and the metal target with a coolant fluid during the step of irradiating.5. The method of claim 1 , wherein the metal target includes zinc-68.6. A method claim 1 , comprising the steps of: a heating element,', 'a sublimation vessel disposed adjacent the heating element such that the heating element heats a portion thereof,', 'a collection vessel removably disposed within the sublimation vessel and being open on an end thereof, and', 'a crucible containing the solid mixture therein and being configured to sealingly position the solid mixture against the collection vessel; and, 'positioning a solid mixture including copper-67 and zinc-68 in a sublimation apparatus, the sublimation apparatus includingheating the solid mixture so as to form a metal vapor having greater than 90% by weight zinc-68,wherein the metal vapor condenses within the collection vessel.7. The method of claim 6 , wherein the zinc-68 has an appreciable vapor pressure at a ...

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Номер записи: 90
27-07-2017 дата публикации

Apparatus and Method for Stripping Tritium from Molten Salt

Номер: US20170213609A1
Автор: Dane F. Wilson, David E. Holcomb
Принадлежит: UT Battelle LLC

A method of stripping tritium from flowing stream of molten salt includes providing a tritium-separating membrane structure having a porous support, a nanoporous structural metal-ion diffusion barrier layer, and a gas-tight, nonporous palladium-bearing separative layer, directing the flowing stream of molten salt into contact with the palladium-bearing layer so that tritium contained within the molten salt is transported through the tritium-separating membrane structure, and contacting a sweep gas with the porous support for collecting the tritium.

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Номер записи: 91
27-07-2017 дата публикации

Container, method for obtaining same and target assembly for the production of radioisotopes using such a container

Номер: US20170213614A1
Автор: Conard Milo
Принадлежит:

The invention relates to a container () for the production of radioisotopes by irradation of a precursor material formed by a one-piece metal casing, the wall of said casing including one thin portion () having a thickness of between 5 and 100 μm, the remainder having a thickness greater than 100 μm. The invention also relates to a method for obtaining the container and to a target assembly using same. 1. A container for producing radioisotopes by irradiation of a precursor material , the container comprising: a metal jacket of integral construction , the metal jacking including a first wall portion having a first thickness between 5 μm and 100 μm , and a second wall portion having a second thickness larger than 100 μm.2. The container as claimed in claim 1 , wherein the jacket has a symmetry of revolution claim 1 , and the first wall portion extends over a fraction of a height of the jacket.3. The container as claimed in further including at least one end having a conical shape claim 1 , a base of the cone being oriented toward an exterior of the container.4. The container as claimed in claim 1 , wherein one end of the jacket is closed.5. The container as claimed in claim 1 , wherein the first wall portion has an outside diameter between 4 mm and 100 mm.6. The container as claimed in claim 1 , wherein the container is at least partially made from at least one of nickel claim 1 , titanium claim 1 , niobium claim 1 , tantalum claim 1 , iron claim 1 , chromium claim 1 , cobalt or a stainless steel.7. A method for obtaining a container as claimed in claim 1 , the method comprising:providing a matrix;electrodepositing on the matrix a thickness of a metallic material, until a first thickness between 5 μm and 100 μm is obtained;masking a fraction of a surface of the matrix;electrodepositing on an unmasked section until a thickness larger than 100 μm is obtained;removing the matrix.8. The method as claimed in claim 7 , wherein the matrix is removed by dissolution.9. A ...

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Номер записи: 92
26-07-2018 дата публикации

GALLIUM-69 ENRICHED TARGET BODIES

Номер: US20180211737A1
Автор: Pipes David, Uhland William
Принадлежит:

Gallium target bodies for producing germanium-68 are disclosed. The targets include an alloy of a base metal and gallium. The alloy is enriched in gallium-69 to increase germanium-68 production. Methods for producing such alloys by electroplating are also disclosed. 1. A target body for producing geramium-68 , the target body comprising:a target substrate plate; gallium, with greater than 60% of the gallium being gallium-69; and', 'a base metal selected from the group consisting of nickel, iron, cobalt, copper and tungsten., 'an alloy that forms an interface with the substrate plate, the alloy comprising2. The target body as set forth in wherein the base metal is nickel.3. The target body as set forth in wherein the target substrate plate comprises copper.4. The target body as set forth in wherein the alloy comprises at least about 10 wt % gallium.5. The target body as set forth in wherein the alloy comprises at least about 50 wt % gallium.6. The target body as set forth in wherein the alloy comprises at least about 75 wt % gallium.7. The target body as set forth in wherein the alloy comprises an amount of gallium-71 claim 1 , the molar ratio of gallium-69 to gallium-71 being at least about 1:1.8. The target body as set forth in wherein the alloy comprises an amount of gallium-71 claim 1 , the molar ratio of gallium-69 to gallium-71 being at least about 5:1.9. The target body as set forth in wherein at least about 65% of the gallium in the alloy is gallium-69.10. The target body as set forth in wherein at least about 95% of the gallium in the alloy is gallium-69.11. A method for forming germanium-68 claim 1 , the method comprising operating a cyclotron or linear accelerator to bombard the target body as set forth in claim 1 , the bombarded target decaying to produce germanium-68.12. A method for forming a target body claim 1 , the method comprising:contacting a target substrate plate with a plating bath comprising a base metal selected from the group consisting of ...

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Номер записи: 93
04-07-2019 дата публикации

CHROMATOGRAPHIC SEPARATION OF MO-99 FROM W-187

Номер: US20190206584A1
Автор: Booij Arend
Принадлежит:

The present disclosure provides a method of separating Mo-99 from W-187 from a solution comprising Mo-99 and W-187. The method comprises contacting a tridentate diglycolanude ligand with a solution comprising Mo-99 and W-187 and eluting W-187 from the tridentate diglycolanude ligand to thereby an eluate comprising W-187. 1. A method of separating Mo-99 from W-187 from a solution comprising Mo-99 and W-187 , the method comprising:contacting a tridentate diglycolamide ligand with a solution comprising Mo-99 and W-187; andeluting the W-187 from the tridentate diglycolamide ligand to thereby yield an eluate comprising W-187.2. The method of wherein the tridentate diglycolamide ligand comprises N claim 1 ,N claim 1 ,N′ claim 1 ,N′-tetraoctyldiglycolamide.3. The method of wherein the Mo-99 is in a molybdate salt.4. The method of wherein the W-187 is in a tungstate salt.5. The method of wherein the tridentate diglycolamide ligand is contacted with the solution comprising Mo-99 and W-187 in a chromatography column.6. The method of wherein the tridentate diglycolamide ligand is contacted with the solution comprising Mo-99 and W-187 in a chromatography cartridge.7. The method of wherein the solution comprising Mo-99 and W-187 further comprises an acid.8. The method of wherein the solution comprising Mo-99 and W-187 further comprises an acid in a concentration of at least about 2.0 M.9. The method of wherein the solution comprising Mo-99 and W-187 further comprises an acid selected from the group consisting of hydrochloric acid claim 1 , nitric acid claim 1 , sulfuric acid claim 1 , and any combination thereof.10. The method of wherein the solution comprises W-187/Mo-99 in an activity ratio of between about 500 kBq and about 18 KBq W-187 activity to about 370 MBq of Mo-99 activity.11. The method of wherein at least about 99% of Mo-99 is removed from the solution.12. The method of wherein the yield of W-187 is at least about 80% claim 1 , or at least about 85% claim 1 , or at ...

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Номер записи: 94
02-08-2018 дата публикации

AUTOMATIC PROCESS PLATFORM FOR THE PRODUCTION OF ASTATINE-211 [At-211]-RADIOPHARMACEUTICALS

Номер: US20180215680A1
Автор: Aneheim Emma, Lindegren Sture
Принадлежит:

A system and method for automatic production of astatine-211 labeled molecules is described. The invention represents a significant advantage in the preparation of At-211 radiopharmaceuticals including better reproducibility, reduced production time and increased radiation safety. The invention also enables routine automatic synthesis of radiopharmaceuticals in a clinical setting, in conjunction or at short distance from a cyclotron unit capable of producing the radionuclide. 2125. The process according to claim 1 , wherein the At-211 is obtained by scraping an irradiated bismuth target to At-211 powder () target material.3120. The process of claim 2 , wherein in the scraping of the irradiated bismuth target is performed using a scraping unit ()4. The process according to claim 1 , wherein in step B) the transfer liquid is an organic solvent5. The process according to claim 1 , wherein in step C) the organic solvent is evaporated leaving a dry residue of At-2116. The process of claim 1 , wherein in step B) the transfer liquid is an adaptive solvent oxidizing At-211.7. The process of claim 1 , wherein in step B) the transfer liquid is an adaptive solvent reducing At-211.8. The process of claim 1 , wherein the process comprises the further step of F) purifying the reaction product from the reaction mixture.9. (canceled)10100106. The process according to claim 1 , wherein an inert gas is used to transport dry-distilled At-211 from the quartz glass receptacle () to the cooling unit () and transfer liquids and solvents within the system.11. (canceled)12. The process according to claim 1 , wherein a prompt reduced pressure is applied to confine At-211 in the system and to speed up rate of distillation.13106. The process according to claim 1 , wherein the cooling unit () is a cryotrap.14. (canceled)15. (canceled)16. (canceled)17. The process according to claim 1 , wherein the precursor molecule is selected from the group comprising inorganic molecules claim 1 , organic ...

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Номер записи: 95
03-08-2017 дата публикации

SOLUTION TARGET FOR CYCLOTRON PRODUCTION OF RADIOMETALS

Номер: US20170221594A1
Автор: Byrne John, DeGrado Timothy R., Pandey Mukesh K.
Принадлежит:

Methods of producing and isolating Ga, Zr, Cu, Zn, Y, Cu, Tc, Ti, N, Mn, or Sc and solution targets for use in the methods are disclosed. The methods of producing Ga, Zr, Cu, Zn, Y, Cu, mTc, Ti, N, Mn, or Sc include irradiating a closed target system with a proton beam. The closed target system can include a solution target. The methods of producing isolated Ga, Zr, Cu, Zn, Y, CU, TC, -Ti, Mn, or Sc by ion exchange chromatography. An example solution target includes a target body including a target cavity for receiving the target material; a housing defining a passageway for directing a particle beam at the target cavity; a target window for covering an opening of the target cavity; and a coolant gas flow path disposed in the passageway upstream of the target window. 1. A solution target for production of a radionuclide from a target material , the solution target comprising:a target body including a target cavity for receiving the target material;a housing defining a passageway for directing a particle beam at the target cavity;a target window foil for covering an opening of the target cavity;a coolant gas flow path disposed in the passageway upstream of the target window; anda coolant fluid flow path disposed in a coolant housing encompassing the target body.2. The solution target of claim 1 , further comprising an energy degrading foil disposed within the passageway and upstream of the coolant flow path.3. The solution target of claim 2 , wherein the energy degrading foil decreases an energy value of the particle beam to 15 MeV or less.4. The solution target of claim 3 , wherein the energy degrading foil comprises aluminum or an aluminum alloy.5. The solution target of claim 1 , wherein the target window foil comprises a cobalt alloy.6. (canceled)7. (canceled)8. (canceled)9. (canceled)10. (canceled)11. (canceled)12. (canceled)13. The solution target of claim 1 , wherein the target cavity is defined by a conical wall.14. The solution target of claim 13 , wherein ...

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Номер записи: 96
10-08-2017 дата публикации

SYSTEM AND METHOD FOR METALLIC ISOTOPE SEPARATION BY A COMBINED THERMAL-VACUUM DISTILLATION PROCESS

Номер: US20170229202A1
Автор: ABEYSEKERA Brian, GELBART William Z., JOHNSON Richard Ray, MATEI Lidia, McRae Glenn
Принадлежит:

A process for the separation of Tc from molybdenum targets is described. The method for separation of Tc isotope from molybdenum targets includes: i) providing an initial multicomponent mixture of elements, the mixture containing Tc; ii) dissolving the multicomponent mixture of elements with an oxidizing agent to oxidize the mixture of elements; iii) heating the mixture of elements at a temperature sufficiently high enough to sublimate a vaporized compound containing Tc; iv) condensing the vaporized compound containing Tc to form a reaction product; v) adding a base to the condensed reaction product to dissolve the Tc containing reaction product to form sodium pertechnetate (NaTcO); and vii) purifying the crude solution of sodium pertechnetate NaTc04 using column chromatography to provide the Tc isotope as a radiochemical compound. 1. A method for separating a Tc isotope from a molybdenum target , the method comprising the steps of:{'sup': '99m', 'providing an initial multicomponent mixture of elements, said mixture containing Tc;'}dissolving said multicomponent mixture of elements with an oxidizing agent to oxidize the mixture of elements;{'sup': '99m', 'heating said mixture of elements at a temperature sufficient to sublimate the mixture and generate a vaporized compound containing Tc;'}{'sup': '99m', 'condensing the vaporized compound containing Tc to form a reaction product;'}{'sup': 99m', '99m, 'adding a base to the condensed reaction product to dissolve the Tc containing reaction product to form a salt of an acid containing Tc;'}{'sup': '99m', 'collecting said dissolved Tc reaction product as a crude solution; and'}{'sup': 99m', '99m, 'purifying the crude solution containing Tc using column chromatography to provide the Tc isotope as a radiochemical compound.'}2. The method of claim 1 , wherein the oxidizing agent is selected from the group consisting of HO claim 1 , HNOand HSO.3. The method of claim 1 , wherein the oxidation occurs at a temperature range of ...

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Номер записи: 97
10-08-2017 дата публикации

TARGET, APPARATUS AND PROCESS FOR THE MANUFACTURE OF MOLYBDENUM-100 TARGETS

Номер: US20170231080A1
Автор: ABEYSEKERA Brian, GELBART William Z., JOHNSON Richard Ray, MATEI Lidia, McRae Glenn
Принадлежит: Best Theratronics Ltd

Apparatuses and methods for production of molybdenum targets, and the formed molybdenum targets, used to produce Tc-99m are described. The target includes a copper support plate having a front face and a back face. The copper support plate desirably has dimensions of thickness of about 2.8 mm, a length of about 65 mm and a width of about 30 mm; and the copper support plate desirably has either a circular or an elliptical cavity centrally formed therein by pressing molybdenum powder into the front face with a depth of about 200-400 microns. Also, the copper support plate includes cooling channels dispensed at the back face; wherein the copper support plate is water cooled by a flow of water during irradiation by a proton beam. Molybdenum powder is embedded and compressed onto the cavity of the copper support plate thereby creating a thin layer of molybdenum onto the copper support plate. 1. A method for manufacturing a target for the production of Tc-99m , comprising the steps of:providing a target support plate including copper, the target support plate having a front face and a back face;placing a molybdenum material in association with the front face of the target support plate; andapplying a pressing force to the molybdenum material to embed the molybdenum material in a cavity formed in the front face by the pressing force, the cavity forming a target material receptacle for the molybdenum material, the pressing force creating a thin substantially uniform layer of the molybdenum material in the formed cavity of the target support plate to form a molybdenum target for forming Tc-99m.2. The method for manufacturing a target for the production of Tc-99m of claim 1 , wherein the molybdenum material comprises a molybdenum powder or a molybdenum disk or pellet.3. The method for manufacturing a target for the production of Tc-99m of claim 2 , wherein the thin substantially uniform layer of the molybdenum material has an elliptical or circular shape corresponding to a ...

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Номер записи: 98
27-08-2015 дата публикации

Techniques for On-Demand Production of Medical Isotopes Such as Mo-99/Tc-99m and Radioactive Iodine Isotopes Including I-131

Номер: US20150243396A1
Автор: Francis Y. Tsang
Принадлежит: GLOBAL MEDICAL ISOTOPE SYSTEMS LLC

A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron's probability of causing fission (n, f) reactions in U-238. The U-238 can be deployed as one or more layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material. The gaseous fission products can be withdrawn from the irradiation chamber on a continuous basis, and the radioactive iodine isotopes (including I-131) extracted.

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Номер записи: 99
06-11-2014 дата публикации

QUASI-NEUTRAL PLASMA GENERATION OF RADIOISOTOPES

Номер: US20140326900A1
Автор: Haaland Peter, Papadopoulos Konstantinos (Dennis), Zigler Arie
Принадлежит:

Methods and apparatus for synthesizing radiochemical compounds are provided. The methods include generating a quasi-neutral plasma jet, and directing the plasma jet onto a radionuclide precursor to provide one or more radionuclides. The radionuclides can be used to prepare radiolabeled compounds, such as radiolabeled biomarkers. 1. A method for production of radioisotopes , the method comprising:generating a quasi-neutral plasma jet; anddirecting the plasma jet onto a radionuclide precursor.2. The method of claim 1 , where the quasi-neutral plasma jet is produced by impinging a light pulse less than about 10seconds in duration onto a target material;{'sub': 'o,', 'sup': 18', '2, 'wherein the dimensionless vector potential, α=0.6λ √I, is greater than about one, where λ is the wavelength in μm and I is the intensity in units of 10W/cm.'}3. The method of claim 2 , where the target material is a solid film or particle; or the target material is a liquid film claim 2 , jet claim 2 , or droplet.4. (canceled)5. The method of claim 2 , where the target material is a gas jet whose number density in the focal region of the light pulse is greater than about 10nuclei per cubic centimeter.6. The method of claim 2 , where light energy that precedes the light pulse and whose vector potential α<10is intercepted by one or more plasma mirrors or thin claim 2 , sacrificial foils.7. The method of claim 2 , where the light pulse is focused on the target material by one or more plasma microlenses.8. The method of claim 2 , where the light pulse is produced by a laser having a wavelength of about 0.4 μm to about 20 μm.9. The method of claim 2 , where the intensity of light impinging on the target prior to the light pulse has α<10.10. The method of claim 1 , where the quasi neutral plasma jet passes from an evacuated region through a window to interact with the radionuclide precursor at a region of higher pressure.11. The method of claim 10 , whereinthe evacuated region is at a pressure of ...

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Номер записи: 100