Abstract: Methods for separating Ra from Pb, Bi, and Th are provided, the methods can include: providing a first mixture comprising Ra, Pb, Bi, and/or Th; providing a system that can include: a first vessel housing a first media; a second vessel in fluid communication with the first vessel, the second vessel housing a second media; and a third vessel in fluid communication with the second vessel, the third vessel housing a third media; and exposing the first mixture to the first media within the first vessel then, through the fluid communication, exposing the first remainder to the second media in the second vessel, then, through fluid communication, exposing the next remainder to the third media in the third vessel, the exposing separating the Th and Bi from the Ra and Pb, and the Ra from the Pb. Methods for separating Ra from being associated with a media are also provided. The methods can include: exposing the Ra and media to a chelating agent to form a mixture comprising the Ra complexed with the chelating agent.

Abstract: 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.

Abstract: The present invention relates to a radioactive chemical waste treatment apparatus including an adsorption unit including an radioactive chemical waste adsorption member for adsorbing and separating radioactive chemical wastes from radioactive chemical waste-containing fluid, and a regeneration unit which is in fluidic communication with the adsorption unit and is for regenerating the radioactive chemical waste adsorption member by desorbing the radioactive chemical wastes from the adsorption member with the radioactive chemical wastes adsorbed thereonto, and to a radioactive chemical waste treatment method including (A) adsorbing radioactive chemical wastes onto a radioactive chemical waste adsorption member and separating the radioactive chemical wastes from a radioactive chemical waste-containing fluid, and (B) desorbing the radioactive chemical wastes from the radioactive chemical waste adsorption member with the radioactive chemical wastes adsorbed thereonto, and regenerating the radioactive chemical wast

Abstract: The invention provides a system for containing irradiated particles, the system having a housing having a closed upstream end and an open downstream end; an axially disposed tunnel extending from the downstream end to the upstream end, wherein longitudinally extending regions of the tunnel define a plurality of threaded surfaces; a sample cup positioned within the tunnel and proximal to the closed upstream end; threaded plugs matingly received by the threaded surfaces so as to provide at least one seal between the sample cup and the atmosphere external of the housing; and a plurality of deformable substrates disposed between the plugs.

Abstract: A system for externally cooling a cask containing heat-emitting spent nuclear fuel includes the cask comprising a radiation shielding body defining an internal cavity configured to hold a canister containing the spent nuclear fuel. A continuously annular cooling jacket extends circumferentially around an external surface of the cask body. The cooling jacket may have a double shell construction including an internal cavity for a cooling medium which provides an external heat sink for absorbing heat radiated from the external wall surface of the cask generated by the spent nuclear fuel. The heat emitted by the spent nuclear fuel is absorbed by the cooling medium in the cooling jacket, thereby in turn cooling the cask. In one embodiment, the cooling medium may be dry ice which undergoes sublimation by absorbing the heat to change from solid to gaseous phase directly. The jacket may be formed of multiple segments.

Abstract: A system for analyzing radiochemical purity of an eluate produced from a generator is provided. The system has an elution detector made of a shielding metal. The elution detector houses a flow path template for flow of an eluate solution and an elutable isotope contained in the generator, and a monitoring device for simultaneously monitoring the generator and the eluate exiting the generator. The generator is a Technetium-99m (Tc-99m) generator, and the elutable isotope may be contained in a packed bed of a column in the Tc-99m generator. A method for analyzing radiochemical purity of eluate produced from a Technetium-99m (Tc-99m) generator or column is also provided.

Abstract: The present disclosure provides fluorinated crown ethers. The fluorinated crown ethers have one or more pendant fluorinated groups (e.g., thioether groups with a terminal fluorinated group). The fluorinated crown ethers have desirable solubility in supercritical carbon dioxide. Also provided are methods and systems for removal of lithium (e.g., lithium ions) from aqueous samples using the fluorinated crown ethers coupled with lipophilic cation exchangers.

Abstract: Methods for purifying 89Zr are provided, 89Zr compositions are provided, isotope compositions are provided that can include: a radio isotope and a nanoparticle, and methods for radio labeling monoclonal antibodies are provided.

Abstract: The use of titanium oxide as a stationary phase in a radio-isotope generator device, wherein said titanium oxide comprises particles having a d50 comprised between 10 and 350 ?m, and having a BET specific surface area comprised between 30 and 300 m2/g, preferably greater than 60 m2/g, and of a generator comprising the stationary phase used.

Abstract: Systems and methods for efficient, effective, and safe separation and isolation of multiple isotopes (e.g., Mo, Zr, Ba, Sr, Te, and lanthanide isotopes) from fission products includes use of a plurality of chromatography columns, each containing a chromatographic resin formulated to target one or more particular isotopes. The system is operable in a “series” configuration to load the multiple columns by a single pass of the sample. Then, the system may be transitioned (e.g., using valves) to a “parallel” configuration in which multiple columns of the system may be operated simultaneously to elute targeted isotopes. Additional parallel operations of the columns, using different eluent compositions, may be used to elute different targeted isotopes. The system may be reconditioned in preparation for a subsequent sample.

Abstract: 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.

Abstract: A method for producing a metal nanoparticle-loaded biopolymer microgel involving the reduction of metal ions in a metal ion-loaded biopolymer microgel. The method yields a metal nanoparticle-loaded biopolymer microgel. Also disclosed is a method of catalytically reducing an organic pollutant in water using the metal nanoparticle-loaded biopolymer microgel and a reducing agent.

Abstract: 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.

Abstract: 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.

Abstract: Microporous glutaraldehyde-crosslinked chitosan sorbents include a plurality of nanoparticles of a high Z element. The nanoparticles are disposed in the cross-linked chitosan-gluteraldehyde composite matrix and integrated with the cross-linked chitosan-gluteraldehyde composite matrix to reduce primary impact of high radiation flux and minimize radiolytic effect on said cross-linked chitosan-gluteraldehyde composite matrix. The plurality of nanoparticles is made from the high Z element such as hafnium (Hf). Methods of making and using the microporous glutaraldehyde-crosslinked chitosan sorbents, and a generator for the radioisotope 99Mo containing the sorbents.

Abstract: A system for selecting a number of transmitting ports in an antenna at an access node includes an access node configured to deploy a first radio air interface to provide wireless services to wireless devices. The access node includes an antenna having ports and configured to communicate data with wireless devices through the ports and the first radio air interface. The access node also includes a processor configured to determine a trigger for dynamically selecting a number of transmitting ports in the antenna based on data received from the wireless devices relating to changes in levels of signals received at the wireless devices. The processor is also configured to estimate an amount of change based on the data relating to the changes. The processor is further configured to select a number of transmitting ports in the antenna based on the estimated amount of change.

Abstract: The invention provides a method for isolating scandium, the method having the steps of dissolving titanium nuclear targets to create a solution; contacting the solution with a resin so as to retain scandium on the resin and generate an eluent containing titanium; contacting the scandium-containing resin with acid of a first concentration to remove impurities from the resin; and contacting the scandium-containing resin with an acid of a second concentration to remove scandium from the resin.

Abstract: The present invention discloses an extracting agent for separating lithium isotopes and an organic extraction phase containing the extracting agent; the organic extraction phase easily enriches 7Li and achieves the separation of lithium isotopes. The present invention also discloses a high-efficiency method for separating lithium isotopes in an aqueous solution, in which the organic extraction phase of the present invention is used, said organic extraction phase being suitable for single-stage and multi-stage extraction processes.

Abstract: 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.

Abstract: The present invention relates to a column chromatographic method of manufacturing non-carrier-added high-purity 177Lu compounds for medicinal purposes. In the method in accordance with the invention a cation exchanger and a suitable chelating agent are used. With the method in accordance with the invention it is possible for the first time to provide non-carrier-added high-purity 177Lu compounds in milligram amounts for pharmaceutical-medicinal purposes from 176Yb compounds irradiated with thermal neutrons, the radionuclides 177Lu and 176Yb being present in an approximate mass ratio of 1:102 to 1:1010 for purification.

Abstract: A method for producing a reaction product containing 99mTC may include providing 100Mo-metal targets to be irradiated, irradiating the 100Mo-metal target with a proton stream having an energy for the induction of a 100Mo(p, 2n)99mTC core reaction, heating the 100Mo-metal target to over 300° C., recovering incurred 99mTc in a sublimation-extraction process with the aid of oxygen gas which is conducted over the 100 Mo-metal target forming 99mTc-Technetium oxide. Further, a device for producing the reaction product containing 99mTc may include a 100Mo metal target, an acceleration unit for providing a proton stream, which can be directed to the 100Mo-Metal target, such that a 100Mo(p, 2n)99mTC core reaction is induced upon irradiation of the 100Mo-metal target by the proton stream, a gas supply line for conducting oxygen gas onto the irradiated 100Mo-metal target to form 99mTC-Technetium oxide, and a gas discharge line to discharge the sublimated 99mTC-Technetium oxide.

Abstract: A hydro-metallurgical method 80 for the removal of uranium, thorium, radium, lead, bismuth and polonium and/or other radionuclides from a radioactive copper concentrate to produce an upgraded copper concentrate having lowered emission levels. The method comprises the step of: subjecting the copper concentrate to an acidic leaching process (NONOX leach) 120 using a sulfate and chloride containing lixiviant under electrochemically controlled conditions, to allow at least partial removal of one or more of the radionuclides to produce the lowered emission upgraded copper concentrate, wherein the leaching process is conducted at elevated temperature and under pressure to suppress boiling in the leaching process.

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

Abstract: A method for sequestering technetium is disclosed. The method includes contacting technetium with apatite under reducing conditions whereby the technetium is absorbed by the apatite.

Abstract: A method for sequestering selenium is disclosed. The method includes contacting selenium with a stannous modified apatite under conditions whereby the selenium is absorbed by the apatite.

Abstract: To use 99mTc as a raw material for a radioactive medicine, a very small amount of 99mTc in the high concentration Mo(99Mo) solution is purified and recovered with high yield without contamination of 99Mo. 99mTc with high purity is recovered by forming a high concentration Mo(99Mo) solution which contains radionuclides 99Mo which is the parent nuclide of 99mTc used for the radioactive medicine and the raw material for its labeled compound, forming a high concentration Mo(99Mo) solution which contains radionuclides 99Mo and 99mTc by generating 99mTc to a radioactive-equilibrium state, getting 99mTc in the high concentration Mo(99Mo) solution adsorbed to activated carbon selectively by feeding the solution to an adsorption column which has activated carbon, and undergoing desorption and purification treatment of 99mTc with a desorbent from the activated carbon to which 99mTc is adsorbed.

Abstract: Microporous glutaraldehyde-crosslinked chitosan sorbents, methods of making and using them, and a generator for the radioisotope 99Mo containing the sorbents.

Abstract: The invention relates to a strontium-82/rubidium-82 generator, comprising a column filled with a cationic exchanger loaded with strontium-82, and having an inlet and an outlet, and a liquid medium, wherein parts of the column, inlet and outlet coming into contact with the liquid medium are iron-free, preferably metal-free, to a method for producing rubidium-82, and to the obtained diagnostic agent.

Abstract: The present invention provides a simple and inexpensive method for removing metal and other impurities in a radioisotope solution. The invention further includes the development of a new parent/daughter generator system for collecting the daughter isotope in a concentrated solution.

Abstract: Microporous glutaraldehyde-crosslinked chitosan sorbents, methods of making and using them, and a generator for the radioisotope 99Mo containing the sorbents.

Abstract: A method for separating an amount of osmium from a mixture containing the osmium and at least one other additional metal is provided. In particular, method for forming and trapping OsO4 to separate the osmium from a mixture containing the osmium and at least one other additional metal is provided.

Abstract: A method for extracting a radioisotope from an aqueous solution, the method comprising: a) intimately mixing a non-chelating ionic liquid with the aqueous solution to transfer at least a portion of said radioisotope to said non-chelating ionic liquid; and b) separating the non-chelating ionic liquid from the aqueous solution. In preferred embodiments, the method achieves an extraction efficiency of at least 80%, or a separation factor of at least 1×104 when more than one radioisotope is included in the aqueous solution. In particular embodiments, the method is applied to the separation of medical isotopes pairs, such as Th from Ac (Th-229/Ac-225, Ac-227/Th-227), or Ra from Ac (Ac-225 and Ra-225, Ac-227 and Ra-223), or Ra from Th (Th-227 and Ra-223, Th-229 and Ra-225).

Abstract: The present invention relates to a column chromatographic method of manufacturing non-carrier-added high-purity 177Lu compounds for medicinal purposes. In the method in accordance with the invention a cation exchanger and a suitable chelating agent are used. With the method in accordance with the invention it is possible for the first time to provide non-carrier-added high-purity 177Lu compounds in milligram amounts for pharmaceutical-medicinal purposes from 176Yb compounds irradiated with thermal neutrons, the radionuclides 177Lu and 176Yb being present in an approximate mass ratio of 1:102 to 1:1010 for purification.

Abstract: 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).

Abstract: There is provided a process for decontaminating and converting volumetrically contaminated radioactive metals, especially nickel, and recovering a decontaminated metal hydroxide or metal carbonate. The process includes the use of hydrogen peroxide to oxidize and remove nucleotides.

Abstract: A solid composition comprises: MnO2; 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 MnO2 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.

Abstract: The invention is directed to a method for the purification of Radium, in particular 226Ra, for target preparation for an essentially pure 225Ac production from available radioactive sources, using an extraction chromatography in order to separate chemically similar elements such as Ba, Sr, and Pb from the desired Ra; wherein said extraction chromatography has an extractant system on the basis of a crown ether. The invention is related to a method for recycling of 226Ra, for target preparation for 225Ac production from radium sources irradiated with accelerated protons (p,2n), after separation of the produced 225Ac. In this method a combination of the above extraction chromatography and a cation exchange chromatography is used. The obtained 226Ra is essentially free of the following chemical contaminants consisting of Ag, Al, As, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, Ga, K, Li, Mg, Mn, Na, Ni, Pb, Sr, V, Zn, and Ba.

Abstract: An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

Abstract: A method for extracting a radioisotope from an aqueous solution, the method comprising: a) intimately mixing a non-chelating ionic liquid with the aqueous solution to transfer at least a portion of said radioisotope to said non-chelating ionic liquid; and b) separating the non-chelating ionic liquid from the aqueous solution. In preferred embodiments, the method achieves an extraction efficiency of at least 80%, or a separation factor of at least 1×104 when more than one radioisotope is included in the aqueous solution. In particular embodiments, the method is applied to the separation of medical isotopes pairs, such as Th from Ac (Th-229/Ac-225, Ac-227/Th-227), or Ra from Ac (Ac-225 and Ra-225, Ac-227 and Ra-223), or Ra from Th (Th-227 and Ra-223, Th-229 and Ra-225).

Abstract: Disclosed is a method for rapidly separating radioactive copper from nickel that contains radioactive copper and radioactive cobalt, which comprises dissolving nickel that contains radioactive copper and radioactive cobalt in an acid solution and leading it to pass through a chelating-ion exchange resin-filled column to thereby make nickel, radioactive copper and radioactive cobalt held by the chelating-ion exchange resin, and then leading an acid solution to pass through the chelating-ion exchange resin-filled column to elute nickel and radioactive cobalt, and thereafter leading an acid solution having a higher concentration than that of the previous acid solution to pass through the chelating-ion exchange resin-filled column after nickel and radioactive elution therefrom to thereby elute radioactive copper.

Abstract: Processes, systems, and methods for selectively regenerating an ion exchange resin generally comprises washing the ion exchange resin with an elution agent that encourages only selected contaminants, and especially selected radioactive isotopes, to disengage or decouple from the resin and enter solution in the elution agent, which thereafter is identified as the elution agent solution. The elution agent solution is then passed through a column of isotope-specific media (ISM). When the selected radioactive isotopes within the elution agent solution come into contact with the constituent media isotopes of the ISM, the selected radioactive isotopes are retained on the reactive surface areas of the ISM or within the interstitial spaces of the porous structures of the constituent media isotopes of the ISM. In some embodiments, the constituent media isotopes of the ISM are embedded, impregnated, or coated with the specific radioactive isotope that the particular ISM are adapted to separate.

Abstract: The present invention provides an apparatus and rapid methods for extracting strontium ions from urine to provide a concentrated and purified strontium-90 extract suitable for scintillation measurements. The methods remove organic compounds, pigments, and alkali metal ions that can interfere with quantitative determination of strontium-90 in urine.

Abstract: A solid composition comprises: MnO2; 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 MnO2 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.

Abstract: Disclosed herein is an radioisotope adsorbent including a bifunctional organosilane compound, including an organosiloxane functional group and a phosphate group, bonded on the surface of silica which is a bone structure of the adsorbent, and a method of preparing the radioisotope adsorbent, and a strontium/yttrium generator using the radioisotope adsorbent. Since the radioisotope adsorbent has a high adsorption capacity for 90Y such as 95% extraction efficiency or more from ICi 90Sr/90Y by using a column packed with 0.4 g of the adsorbent with a high-purity of 90Y, it can be usefully to be employed in the fields requiring 90Y.

Abstract: The present invention provides a method for the generation of 223Ra of pharmaceutically tolerable purity comprising i) preparing a generator mixture comprising 227Ac, 227Th and 223Ra; ii) loading said generator mixture onto a strong base anion exchange resin; iii) eluting said 223Ra from said strong base anion exchange resin using a first mineral acid in an alcoholic aqueous solution to give a first eluted 223Ra solution; iv) loading the 223Ra of the first eluted 223Ra solution onto a strong acid cation exchange resin; and v) eluting the 223Ra from said strong acid cation exchange resin using a second mineral acid in aqueous solution to provide a second eluted solution. The invention additionally provides products of corresponding purity and/or products obtained or obtainable by such a method.

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 for separating an amount of osmium from a mixture containing the osmium and at least one other additional metal is provided. In particular, method for forming and trapping OsO4 to separate the osmium from a mixture containing the osmium and at least one other additional metal is provided.

Abstract: A precipitator comprises a counter-current circulation between the reacting substances (8, 6, 7 ?12) and a non-miscible and chemically inert organic confinement phase (10?9) to maintain the phase containing the reagent in an emulsion. The walls of the precipitator are hydrophobic to prevent the adhesion of the precipitate. The emulsion is maintained by a mobile stirrer body (2). The precipitate is removed continually by a scavenging flow rate device (16).

Abstract: An alteration of the traditional zinc/zinc-amalgam reduction procedure which eliminates both the hazardous mercury and dangerous hydrogen gas generation. In order to avoid the presence of water and hydrated protons in the working solution, which can oxidize Eu2+ and cause hydrogen gas production, a process utilizing methanol as the process solvent is described. While methanol presents some flammability hazard in a radiological hot cell, it can be better managed and is less of a flammability hazard than hydrogen gas generation.