Abstract: An equatorial anthropic radiation source and a method of making an equatorial anthropic radiation source are described. The radiation source is useful in diagnostic imaging applications in healthcare or other industries (e.g. computerized three-dimensional segmental imaging; Crompton scattering imaging techniques; radiation detector check and calibration, in particular CdZnTe detectors commonly used in medical imaging).

Abstract: A target irradiation system for irradiating a radioisotope target in a vessel penetration of a fission reactor, includes a target delivery assembly with a body defining a central bore, a basket that is slidably receivable within the central bore of the body, and a winch that is connected to the basket by a cable. The target delivery assembly is affixed to the vessel penetration of the reactor; and a target passage is in fluid communication with the target delivery assembly. The basket is configured to receive a radioisotope target therein via the target passage and be lowered into the vessel penetration of the reactor when irradiating the radioisotope target. The target delivery system forms a portion of the pressure boundary of the reactor when in fluid communication with the reactor.

Abstract: An apparatus for manufacturing a radioisotope comprises a container. The container comprising a portable neutron source and a solution that comprises a particular isotope. The portable neutron source is surrounded by the solution. The solution comprises at least one of copper phthalocyanine or copper salicylaldehyde o-phenylene diamine. The portable neutron source emits neutrons that react with the particular isotope resulting in the transformation of the particular isotope into the radioisotope.

Abstract: A method of producing radionuclides from irradiation targets in a nuclear reactor uses at least one instrumentation tube system of a commercial nuclear reactor. Irradiation targets and dummy targets are inserted into an instrumentation finger and the irradiation targets are activated by exposing them to neutron flux in the nuclear reactor core to form a radionuclide. The dummy targets hold the irradiation targets at a predetermined axial position in the reactor core corresponding to a pre-calculated neutron flux density sufficient for converting the irradiation targets to the radionuclide. Separating the dummy targets from the activated irradiation targets includes exposure to a magnetic field to retain either the dummy targets or the activated irradiation targets in the instrumentation tube system and release the other one of the activated irradiation target or the dummy target from the instrumentation tube system. An apparatus adapted to the above method is also provided.

Abstract: 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.

Abstract: Methods of producing and isolating 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc and solution targets for use in the methods are disclosed. The methods of producing 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc include irradiating a closed target system with a proton beam. The closed target system can include a solution target. The methods of producing isolated 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 52Mn, or 44Sc further include isolating 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 52Mn, or 44Sc 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.

Abstract: A method of collecting 3He from a nuclear reactor may include the steps of a) providing heavy water at least part of which is exposed to a neutron flux of the reactor, b) providing a cover gas in fluid communication with the heavy water, c) operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by ?? decay and at least a portion of the 3He gas escapes from the heavy water and mixes with the cover gas, d) extracting an outlet gas stream, the outlet gas stream comprising a mixture of the cover gas and the 3He gas and e) separating the 3He gas from the outlet gas stream.

Abstract: A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated target suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.

Abstract: A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants.

Abstract: In one embodiment, an irradiation target encapsulation assembly, includes a container, at least one first irradiation target disposed in the container, at least one second irradiation target disposed in the container, and a positioning structure configured to position the first irradiation target closer to an axial center of the container than the second irradiation target.

Abstract: A liquid lithium-cooled fission reactor optimized for producing radioactive materials. The reactor is designed to enhance the availability of rare radioactive materials that have significant value for national defense, industrial research, and medical care. This invention has tangible design attributes that can be tailored to create one or more scarce and valuable radioactive materials. In particular, the reactor design is optimized for low-cost production of large quantities of radioactive tritium needed in national-defense and fusion-breeder programs. There are four core designs applied to this invention, all of which produce tritium and surplus heat that can generate byproduct electricity, thereby reducing the cost of radioactive-material production.

Abstract: An apparatus for generating medical isotopes provides for the irradiation of dry-phase, granular uranium compounds which are then dissolved in a solvent for separation of the medical isotope from the irradiated compound. Once the medical isotope is removed, the dissolved compound may be reconstituted in dry granular form for repeated irradiation.

Abstract: Recycling of isotopically enriched molybdenum metal targets that are suitable for the large scale cyclotron production of 99mTc or 94mTc includes the charged particle irradiation of an enriched molybdenum metal target to produce a technetium isotope, separation of the technetium isotope following irradiation of the molybdenum, re-claiming the molybdenum metal and reformation of the molybdenum target for a further irradiation step. This process may then be repeated. Separation of the technetium isotope preferably is achieved by oxidative dissolution of the molybdenum thereby removing it from a target support plate, and forming molybdate and pertechnetate. The technetium isotope is isolated by various means, such as the ABEC process. To reuse the molybdenum, additional steps of isolating the molybdate and reducing it back to molybdenum metal are required.

Abstract: 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.

Abstract: A method of collecting 3He from a nuclear reactor may include the steps of a) providing heavy water at least part of which is exposed to a neutron flux of the reactor, b) providing a cover gas in fluid communication with the heavy water, c) operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by ?? decay and at least a portion of the 3He gas escapes from the heavy water and mixes with the cover gas, d) extracting an outlet gas stream, the outlet gas stream comprising a mixture of the cover gas and the 3He gas and e) separating the 3He gas from the outlet gas stream.

Abstract: A target supply apparatus includes a tank for storing a liquid target material, a nozzle for outputting the liquid target material in the tank, and a gas supply source for supplying gas into the tank, and controls a gas pressure inside the tank with a pressure of the gas supplied from the gas supply source which is provided with a pressure regulator. The target supply apparatus also includes a pressure-decrease gas passage of which one end is connected to the tank and the other end forms an exhaust port, a pressure-decrease valve provided on the pressure-decrease gas passage, and a controller for controlling open/close of the pressure-decrease valve. The controller, when the target material is caused not to output from the nozzle, opens the pressure-decrease valve and decreases the pressure inside the tank.

Abstract: Example embodiments and methods are directed to irradiation target positioning devices and systems that are configurable to permit accurate irradiation of irradiation targets and accurate production of daughter products, including isotopes and radioisotopes, therefrom. These include irradiation target plates having precise loading positions for irradiation targets, where the targets may be maintained in a radiation field. These further include a target plate holder for retaining and positioning the target plates and irradiation targets therein in the radiation field. Example embodiments include materials with known absorption cross-sections for the radiation field to further permit precise, desired levels of exposure in the irradiation targets. Example methods configure irradiation target retention systems to provide for desired amounts of irradiation and daughter product production.

Abstract: Method of producing molybdenum-99, comprising accelerating ions by means of an accelerator; directing the ions onto a metal target so as to generate neutrons having an energy of greater than 10 MeV; directing the neutrons through a converter material comprising techentium-99 to produce a mixture comprising molybdenum-99; and, chemically extracting the molybdenum-99 from the mixture.

Abstract: Example embodiments are directed to apparatuses and methods for producing radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable from nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes.

Abstract: Provided are methods to separate an isotope from a first solution including uranium. The methods may include (a) cleaning the first solution to form a second solution including the uranium and a third solution including the isotope; (b) oxidizing the third solution to form an oxidized isotope; and (c) separating the oxidized isotope.

Abstract: Methods of producing cesium-131. The method comprises dissolving at least one non-irradiated barium source in water or a nitric acid solution to produce a barium target solution. The barium target solution is irradiated with neutron radiation to produce cesium-131, which is removed from the barium target solution. The cesium-131 is complexed with a calixarene compound to separate the cesium-131 from the barium target solution. A liquid:liquid extraction device or extraction column is used to separate the cesium-131 from the barium target solution.

Abstract: Systems and methods using a double-walled portable container with pressurized gaseous Xe-124 are used as a target for thermal neutron irradiation that generates Xe-125. The portable container is transferred, while submerged in the reactor pool, to a mobile radiation shield container, which are then removed from the reactor pool and connected to the production apparatus that provides handling and recovery functions while properly shielded to minimize radiation exposure. A rapid and efficient transfer of induced Xe-125 and remaining Xe-124 is then accomplished into a clean spiral trap container in which the Xe-125 radioactivity is converted to Iodine-125. After the decay period is completed, Xe-124 and remaining Xe-125 are recovered leaving I-125 deposited on the internal surface of the spiral trap. I-125 is then removed with appropriate solvents.

Abstract: A method of isolating a radioisotope according to example embodiments may include vaporizing a source compound containing a first isotope and a second isotope of an element, wherein the second isotope may have at least one of therapeutic and diagnostic properties when used as a radiopharmaceutical. The vaporized source compound may be ionized to form charged particles of the first and second isotopes. The charged particles may be separated to isolate the particles of the second isotope. The isolated charged particles of the second isotope may be collected with an oppositely-charged collector. Accordingly, the isolated second isotope may be used to produce therapeutic and/or diagnostic radiopharmaceuticals having higher specific activity.

Abstract: The present invention provides a method for large scale production of cesium-131 (Cs-131) with low cesium-132 (Cs-132) content, where the Cs-131 is produced via barium-131 (Ba-131) decay. Uses of the Cs-131 produced by the method include cancer research and treatment, such as for use in brachytherapy. Cesium-131 is particularly useful in the treatment of faster growing tumors.

Abstract: 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.

Abstract: Methods and apparatus are provided for producing and extracting Mo-99 and other radioisotopes from fission products that overcome the drawbacks of previously-known systems, especially the excessive generation of radioactive wastes, by providing gas-phase extraction of fission product radioisotopes from a nuclear fuel target using a mixture including halide and an oxygen-containing species with heat to convert the fission product radioisotopes to gas (e.g., Mo-99 to MoO2Cl2 gas). The gaseous species are evacuated to a recovery chamber where the radioisotopes solidify for subsequent processing, while the substantially intact uranium target made available for further irradiation and extraction cycles.

Abstract: 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.

Abstract: 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 (m, f) reactions in U-238. The U-238 can be deployed as layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material.

Abstract: Example embodiments are directed to methods and apparatuses for generating desired isotopes within water rods of nuclear fuel assemblies. Example methods may include selecting a desired irradiation target based on the target's properties, loading the target into a target rod based on irradiation target and fuel assembly properties, exposing the target rod to neutron flux, and/or harvesting isotopes produced from the irradiation target from the target rod. Example embodiment target rods may house one or more irradiation targets of varying types and phases. Example embodiment securing devices include a ledge collar and/or bushing that support target rods within a water rod and permit moderator/coolant flow through the water rod. Other example embodiment securing devices include one or more washers with one or more apertures drilled therein to hold one or more example embodiment target rods in a water rod while permitting coolant/moderator to flow through the water rod.

Abstract: Methods of producing cesium-131. The method comprises dissolving at least one non-irradiated barium source in water or a nitric acid solution to produce a barium target solution. The barium target solution is irradiated with neutron radiation to produce cesium-131, which is removed from the barium target solution. The cesium-131 is complexed with a calixarene compound to separate the cesium-131 from the barium target solution. A liquid:liquid extraction device or extraction column is used to separate the cesium-131 from the barium target solution.

Abstract: An apparatus for containing and cooling enriched water for the production of activated fluorine (18F). A target assembly includes internal cooling channels in which developed flow of a coolant removes the heat from the target liquid in the target chamber. In one embodiment, the target assembly is fabricated of tantalum.

Abstract: A material is exposed to a neutron flux by distributing it in a neutron-diffusing medium surrounding a neutron source. The diffusing medium is transparent to neutrons and so arranged that neutron scattering substantially enhances the neutron flux to which the material is exposed. Such enhanced neutron exposure may be used to produce useful radio-isotopes, in particular for medical applications, from the transmutation of readily-available isotopes included in the exposed material. It may also be used to efficiently transmute long-lived radioactive wastes, such as those recovered from spent nuclear fuel. The use of heavy elements, such as lead and/or bismuth, as the diffusing medium is particularly of interest, since it results in a slowly decreasing scan through the neutron energy spectrum, thereby permitting very efficient resonant neutron captures in the exposed material.

Abstract: A method of isolating 186Re according to example embodiments may include vaporizing a source compound containing 185Re and 186Re. The vaporized source compound may be ionized to form negatively-charged molecules containing 185Re and 186Re. The negatively-charged molecules may be separated to isolate the negatively-charged molecules containing 186Re. The isolated negatively-charged molecules containing 186Re may be collected with a positively-charged collector. Accordingly, the isolated 186Re may be used to produce therapeutic and/or diagnostic radiopharmaceuticals having higher specific activity.

Abstract: In a method for producing a radionuclide, a target chamber is filled with target fluid and pressurized. A particle beam is applied to the target chamber to irradiate target material of the target fluid, and the target fluid becomes heated. The heated target liquid may expand out from the target chamber through a lower opening. A space including target fluid vapor may be created in an upper region of the target chamber. The upper region is sealed to maintain the vapor space.

Abstract: An apparatus for producing a radionuclide includes a target chamber including a beam strike region for containing a liquid and a condenser region for containing a vapor. A particle beam source is operatively aligned with the beam strike region, and a lower liquid conduit communicates with the beam strike region. The condenser region is disposed above the beam strike region in fluid communication therewith for receiving heat energy from the beam strike region and transferring condensate to the beam strike region. The lower liquid conduit transfers liquid to and from the beam strike region. In operation, the target chamber acts as a thermosyphon that is self-regulating in response to heat energy deposited by the particle beam source. A portion of the liquid expands into the lower liquid conduit prior to boiling. After boiling begins, a vapor void is created above the liquid and an evaporation/condensation cycle is established, with additional liquid being displaced into the lower liquid conduit.

Abstract: A new composition of matter includes 195mPt characterized by a specific activity of at least 30 mCi/mg Pt, generally made by method that includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A new composition of matter includes 195mPt characterized by a specific activity of at least 30 mCi/mg Pt, generally made by method that includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A method of preparing high-specific-activity 195mPt includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A method for transmuting spent fuel from a nuclear reactor includes the step of separating the waste into components including a driver fuel component and a transmutation fuel component. The driver fuel, which includes fissile materials such as Plutonium239, is used to initiate a critical, fission reaction in a reactor. The transmutation fuel, which includes non-fissile transuranic isotopes, is transmuted by thermal neutrons generated during fission of the driver fuel. The system is designed to promote fission of the driver fuel and reduce neutron capture by the driver fuel. Reacted driver fuel is separated into transuranics and fission products using a dry cleanup process and the resulting transuranics are mixed with transmutation fuel and re-introduced into the reactor. Transmutation fuel from the reactor is introduced into a second reactor for further transmutation by neutrons generated using a proton beam and spallation target.

Abstract: A method of preparing high-specific-activity 195mPt includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A method for transmuting spent fuel from a nuclear reactor includes the step of separating the waste into components including a driver fuel component and a transmutation fuel component. The driver fuel, which includes fissile materials such as Plutonium239, is used to initiate a critical, fission reaction in a reactor. The transmutation fuel, which includes non-fissile transuranic isotopes, is transmuted by thermal neutrons generated during fission of the driver fuel. The system is designed to promote fission of the driver fuel and reduce neutron capture by the driver fuel. Reacted driver fuel is separated into transuranics and fission products using a dry cleanup process and the resulting transuranics are mixed with transmutation fuel and re-introduced into the reactor. Transmutation fuel from the reactor is introduced into a second reactor for further transmutation by neutrons generated using a proton beam and spallation target.

Abstract: A system and method for rapidly analyzing elemental abundances in rock or soil samples (14) under field conditions. The system uses a portable neutron source (12) to allow neutron activation analysis of elements having identifiable radioactive decay characteristics. A radiation detector (18) detects radiation released by the sample (14) and provides radiation testing results to an amplifier (26) for computing the concentration of trace elements in the sample with a high degree of accuracy.

Abstract: A neutron detection system for detection of contaminants contained within a bulk material during recycling includes at least one neutron generator for neutron bombardment of the bulk material, and at least one gamma ray detector for detection of gamma rays emitted by contaminants within the bulk material. A structure for analyzing gamma ray data is communicably connected to the gamma ray detector, the structure for analyzing gamma ray data adapted. The identity and concentration of contaminants in a bulk material can also be determined. By scanning the neutron beam, discrete locations within the bulk material having contaminants can be identified.

Abstract: Inert gaseous fission products, including beryllium, rubidium, and krypton isotopes, resulting from the operation of a uranyl sulfate water solution nuclear reactor are passed through a delaying device to precipitate out strontium-90, then passed to a second delaying device to precipitate out the desired strontium-89.

Abstract: An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention.

Abstract: Inert gaseous fission products, including beryllium, rubidium, and krypton isotopes, resulting from the operation of a uranyl sulfate water solution nuclear reactor are passed through a delaying device to precipitate out strontium-90, then passed to a second delaying device to precipitate out the desired strontium-89.

Abstract: The present invention includes the use of N-isomers as a source of energy and of neutrons, and the use of K-isomers as a source of energy when associated with a source of neutrons. Although there is strong indirect evidence for the existence of shape isomers in nuclei lighter than actinides, super-deformed (SD) isomeric states have not yet been directly observed. However, rotational bands from such SD states have been observed through &ggr;-ray transitions within high-energy rotational states of this band, as populated by HI reactions. The lifetimes for the shape isomers are likely to be small, but may be increased by effects like the odd-even effects already observed for fission isomers. By contrast, K-isomers have been observed and investigated. If N-isomers are found with the required properties (especially with sufficiently long lifetimes) and produced in sufficient quantities, portable neutron sources more intense than existing neutron sources could be obtained.

Abstract: Methods for the production of radionuclides suitable for use in radiopharmaceuticals for diagnostic and therapeutic applications, and specifically, to the production of 186Re, 188Re and other radionuclides such as 195mPt and 198Au using an inorganic Szilard-Chalmers reaction. Thin-film and powdered 185 or 187Reo metal targets, and 185 or 187Re oxide/metal oxide target compositions with rhenium in a lower, relatively reduced oxidation state are prepared. The thin-film rhenium targets are aged for at least about 24 hours and then irradiated with neutrons in the present of an oxidizing medium sufficient to form a product nuclide in the higher oxidized state of perrhenate, ReO4−. Significantly, the rate and/or extent of oxidation of target nuclides which do not react with a neutron is controlled. For example, oxidation of such non-bombarded target nuclides is minimized by irradiating under vacuum, controlling the amount of oxidizing agent present, cooling during irradiation, etc.

Abstract: Radioactive stents used in angioplasty on sclerotic coronary arteries without the risk of restenosis can be produced by ion injecting 133Xe into the surfaces of stents as a nuclide that has a shorter half-life and emits a smaller maximum energy of &bgr;-rays than 32p Uniform ion injection is accomplished using an apparatus capable of uniform irradiation of the stents with 133Xe ion beams. The source of 133Xe is a nuclear fission product generated from 235U in the fuel rods in nuclear reactor.

Abstract: A method of producing and purifying Cs-131 comprising the steps of dissolving irradiated Ba comprised of natural or enriched Ba including Ba-130, Ba-131, and Cs-131 from the decay of Ba-131, in an acid, precipitating the Ba, separating the Cs-131 using an ion exchange media, and eluting the Cs-131 from the exchanger to recover the purified Cs-131.