Abstract: The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

Abstract: A method for producing 99mTc may include: providing a solution comprising 100Mo-molybdate-ions; providing a proton beam having an energy suitable for inducing a 100Mo(p,2n)99mTc-nuclear reaction when exposing 100Mo-molybdate-ions; exposing the solution to the proton beams and inducing a 100Mo(p,2n)99mTc-nuclear reaction; and applying an extraction method for extracting the 99mTc from the solution. Further, a device for producing 99mTc may include: a solution with 100Mo-molybdate-ions; an accelerator for providing a proton beam with energy which is suitable for inducing a 100Mo(p,2n)99mTc-nuclear reaction when exposing 100Mo-molybdate-ions, for exposing the solution and for inducing a 100Mo(p,2n)99mTc-nuclear reaction; and an extraction step for extracting 99mTc from the solution.

Abstract: Methods for producing Tc-99m radioisotope by proton irradiation of a fluid target matrix. A method of producing Tc-99m includes irradiating a fluid target matrix comprising Mo-100 with a proton beam to transform at least a portion of Mo-100 to Tc-99m. Optionally, the fluid target matrix further includes at least one of O-18, O-16, or N-14, which upon exposure to the proton beam concurrently transform at least a portion of O-18 to F-18, at least a portion of O-16 to N-13, at least a portion of the O-16 to O-15, or at least a portion of N-14 to C-11. The method further includes isolating Tc-99m and optionally at least one of F-18, N-13, O-15, or C-11 from the irradiated fluid target matrix. An additional source of Tc-99m is available from the decay of Mo-99 that is co-produced from the Mo-100 during irradiation with the proton beam.

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: An apparatus for producing 99Mo from a plurality of 100Mo targets through a photo nuclear reaction on the 100Mo 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 100Mo 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.

Abstract: A system and a process for producing selected isotopic daughter products from parent materials characterized by the steps of loading the parent material upon a sorbent having a functional group configured to selectively bind the parent material under designated conditions, generating the selected isotopic daughter products, and eluting said selected isotopic daughter products from the sorbent. In one embodiment, the process also includes the step of passing an eluent formed by the elution step through a second sorbent material that is configured to remove a preselected material from said eluent. In some applications a passage of the material through a third sorbent material after passage through the second sorbent material is also performed.

Abstract: The present invention provides a method for producing molybdenum-99 comprising: i) providing an electron accelerator; ii) providing a molybdenum converter/target unit (Mo-CTU) comprising one or more metallic components, wherein each one of said metallic components is made of a material selected from the group consisting of natural molybdenum, molybdenum-100, molybdenum-98, and mixtures thereof; iii) directing an electron beam generated via said electron accelerator onto said Mo-CTU to produce a braking radiation (bremsstrahlung); iv) employing said bremsstrahlung onto said Mo-CTU to produce molybdenum-99 and neutrons via a photo-neutron reaction; v) slowing down the neutrons produced in step iv) with a low atomic liquid, e.g. distilled water; and optionally vi) employing the neutrons produced in step iv) to produce a complementary amount of molybdenum-99 via a neutron capture reaction on said Mo-CTU. The invention further provides an apparatus for producing molybdenum-99.

Abstract: A capsule for holding, irradiating, and eluting a material is provided. Methods of fabricating and using the capsule are also provided. The capsule may include a multidiameter tube with a first end region, a second end region, and a middle region. Washers and filters are provided in the end regions and the end regions may be sealed using various methods and materials with the end caps press fit into the end regions. The middle region is designed to store a material to be irradiated by a neutron flux source. The capsule components may be made from materials having a low nuclear cross section so that the capsule may be handled safely after an irradiation step is performed. The capsule is also designed to have a symmetric configuration as an elution and irradiation column so that the same capsule may be used to elute the material within the middle region of the capsule after an irradiation step is performed.

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: To provide a method of producing radioactive molybdenum solution suitable for extracting 99mTc to be used as radioactive diagnostic drug by way of establishing a production process for high-density MoO3 pellets with a lower amount of insoluble content when dissolving the pellets. The method has the steps of: preparing MoO3 powder, fabricating a MoO3 pellet by filling said MoO3 powder in a heated die and sintering in an air, oxidizing said MoO3 pellet, producing neutron-irradiated MoO3 pellets by irradiating a neutron on said oxidized MoO3 pellet, and obtaining radioactive molybdenum solution by dissolving said neutron-irradiated MoO3 pellet.

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: A method and apparatus for producing Mo-99 from Mo-100 for the use of the produced Mo-99 in a Tc-99m generator without the use of uranium is presented. Both the method and apparatus employ high energy gamma rays for the transformation of Mo-100 to Mo-99. The high energy gamma rays are produced by exposing a metal target to a moderated neutron output of between 6 MeV and 14 MeV. The resulting Mo-99 spontaneously decays into Tc-99m and can therefore be used in a Tc-99m generator.

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: A large enhancement of neutron flux is realized when a primary target of D2O and H2O is contained in a vessel, is irradiated by an electron beam incident on a gamma converter and where the vessel is enclosed within a neutron reflector material including Nickel and Polyethylene. A very large enhancement of neutron flux is realized when a secondary target of LEU is mixed with the primary target resulting in a very large enhanced production of Molybdenum-99. The primary target and the secondary target is contained in cylindrical or spherical vessels. A process for extracting produced Mo-99 from the reaction vessel solution.

Abstract: A combinatorial heterogeneous-homogeneous reactor configuration in which an array or groups of homogeneous fuel assemblies are interlinked together in a heterogeneous lattice. The present invention removes the limitation of a homogeneous reactor by providing a reactor concept that utilizes the inherent advantages of homogeneous fuel elements but in a heterogeneous fuel lattice arrangement that limits the power density of any one homogeneous fuel element and yet forms a reactor arrangement that is capable of producing any product demand of interest. The present invention provides a method for producing medical isotopes by the use of a modular reactor core comprised of homogeneous fuel assemblies arranged in a regular rectangular or triangular pitch lattice.

Abstract: A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith.

Abstract: A system and method for eluting one or more radioisotope processing elements, such as generators and/or concentration and purification components, with a gas over eluent delivery mechanism is disclosed. The system is adapted to an elution/concentration system using two such mechanisms, one for an initial elution of one or more generators, and a second for "re-elution" of concentrated and purified radioisotope contained in a concentration subsystem. The operation of such a dual mechanism system may also be completely or partially automated. The invention also discloses a concentration subsystem containing all single-use components needed for concentration and/or purification, as a single use, self sealed, cartridge that can safely store waste eluate for safe and simple disposal.

Abstract: The current invention involves a means for the production and extraction of the isotope molybdenum-99 for medical purposes in a waste free, simple, and economical process. Mo-99 is generated in the uranyl sulphate nuclear fuel of a homogeneous solution nuclear reactor and extracted from the fuel by a solid polymer sorbent with a greater than 90% purity. The sorbent is composed of a composite ether of a maleic anhydride copolymer and .alpha.-benzoin-oxime.

Abstract: Insuring a constant supply of radioisotopes is of great importance to medicine and industry. This invention addresses this problem, and helps to solve it by introducing a new apparatus for transmutation of isotopes which enables swift and flexible production on demand.

Abstract: An improved method for producing .sup.99m Tc compositions. .sup.100 Mo metal is irradiated with photons in a particle (electron) accelerator to produce .sup.99 Mo metal which is dissolved in a solvent. A solvated .sup.99 Mo product is then dried to generate a supply of .sup.99 MoO.sub.3 crystals. The crystals are thereafter heated at a temperature which will sublimate the crystals and form a gaseous mixture containing vaporized .sup.99m TcO.sub.3 and vaporized .sup.99m TcO.sub.2 but will not cause the production of vaporized .sup.99 MoO.sub.3. The mixture is then combined with an oxidizing gas to generate a gaseous stream containing vaporized .sup.99m Tc.sub.2 O.sub.7. Next, the gaseous stream is cooled to a temperature sufficient to convert the vaporized .sup.99m Tc.sub.2 O.sub.7 into a condensed .sup.99m Tc-containing product. The product has high purity levels resulting from the use of reduced temperature conditions and ultrafine crystalline .sup.99 MoO.sub.3 starting materials with segregated .sup.

Abstract: An improved method for producing .sup.99m Tc compositions from .sup.99 Mo compounds. .sup.100 Mo metal or .sup.100 MoO.sub.3 is irradiated with photons in a particle (electron) accelerator to ultimately produce .sup.99 MoO.sub.3. This composition is then heated in a reaction chamber to form a pool of molten .sup.99 MoO.sub.3 with an optimum depth of 0.5-5 mm. A gaseous mixture thereafter evolves from the molten .sup.99 MoO.sub.3 which contains vaporized .sup.99 MoO.sub.3, vaporized .sup.99m TcO.sub.3, and vaporized .sup.99m TcO.sub.2. This mixture is then combined with an oxidizing gas (O.sub.2(g)) to generate a gaseous stream containing vaporized .sup.99m Tc.sub.2 O.sub.7 and vaporized .sup.99 MoO.sub.3. Next, the gaseous stream is cooled in a primary condensation stage in the reaction chamber to remove vaporized .sup.99 MoO.sub.3. Cooling is undertaken at a specially-controlled rate to achieve maximum separation efficiency.

Abstract: Insuring a constant supply of radioisotopes is of great importance to medicine and industry. This invention addresses this problem, and helps to solve it by introducing a new apparatus for transmutation of isotopes which enables swift and flexible production on demand.

Abstract: A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith.

Abstract: Medical isotopes are produced using a lower power, low cost nuclear reactor which permits the use of all the fission products produced in the reactor. Medical isotopes such as Molybdenum-99 are produced in a reactor operating at a power of 100 to 500 kilowatts.

Abstract: A process for selective separation of strontium-82 and strontium-85 from proton irradiated molybdenum targets is provided and includes dissolving the molybdenum target in a hydrogen peroxide solution to form a first ion-containing solution, passing the first ion-containing solution through a first cationic resin whereby ions selected from the group consisting of molybdenum, niobium, technetium, selenium, vanadium, arsenic, germanium, zirconium and rubidium remain in the first ion-containing solution while ions selected from the group consisting of rubidium, zinc, beryllium, cobalt, iron, manganese, chromium, strontium, yttrium and zirconium are selectively adsorbed by the first resin, contacting the first resin with an acid solution capable of stripping adsorbed ions from the first cationic exchange resin whereby the adsorbed ions are removed from the first resin to form a second ion-containing solution, evaporating the second ion-containing solution for time sufficient to remove substantially all of the acid

Abstract: A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

Abstract: A generator of radionuclides comprises a generator column (10) containing the radionuclide and provided with an inlet (12) and an outlet (14) for eluent, first and second reservoirs (16, 18) for eluent, and connecting means for passing a pre-determined column of eluent from the second reservoir through the generator column. A part defining the second reservoir, which part may be the second reservoir itself, is rotatable to determine the volume of eluent passed through the column. The second reservoir is preferably shaped as a sector of a cylinder rotatable around a horizontal axis.The generator is well suited for vacuum elution of a variable pre-determined volume of eluent into a single size collection vial with subsequent drying of the generator column.

Abstract: This invention relates to a standard component .sup.99m TC elution generator useful for medical purposes and consisting of prefabricated component parts. The main generator column of the device may be used both as an irradiation container and an elution container, enabling the user to supply activated or nonactivated parts. The main generator column made from by neutrns little activable materials serves first as reactor irradiation ampoule and after having been activated in the reactor by neutrons and after a simple adjustment serves directly as the generator column. It is filled with water insoluable molybdates or polymolybdates (with the molybdenum content in the range 10-40%), easily releasing .sup.99m Tc generated by radioactive decay of the mother .sup.99 Mo formed in it by neutron activation. This column filling serves originally as target material for reactor irradiations and afterwards it is directly used as the generator elution matrix.

Abstract: A method for decreasing the amount of hazardous radioactive reactor waste materials by separation from the waste of materials having long-term risk potential and exposing these materials to a thermal neutron flux. The utilization of thermal neutrons enhances the natural decay rates of the hazardous materials while the separation for recycling of the hazardous materials prevents further transmutation of stable and short-lived nuclides.