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 Ac-225 and/or Bi-213 may thus be obtained.

Abstract: The present invention provides a new chemical process, a new cassette configuration, and new software for the automated production of multiple batches of an [18F]labelled compound on a single cassette. The invention allows one synthesizer in one hot cell to produce sequentially a plurality of batches of [18F]-labelled PET tracer in the same day. In particular, the present invention provides a novel arrangement useful for the trapping of [18F]fluoride and recovery of [18O]water.

Abstract: The present invention provides a new chemical process, a new cassette configuration, and new software for the automated production of multiple batches of an [18F]-labelled compound on a single cassette. The invention allows one synthesizer in one hot cell to produce sequentially a plurality of batches of [18F]-labelled PET tracer in the same day. In particular, the present invention provides a novel arrangement useful for purifying two consecutive batches of a reaction mixture comprising an 18F-labelled compound.

Abstract: It is intended to provide an industrially advantageous method for obtaining zirconium tungsten phosphate that is useful as a negative thermal expansion material and exhibits a single phase in X-ray diffraction. The method for producing zirconium tungsten phosphate according to the present invention comprises using a mixture of a tungsten compound and an amorphous compound containing phosphorus and zirconium as a reaction precursor and calcining the reaction precursor. Preferably, the reaction precursor has an infrared absorption peak at least at 950 to 1150 cm?1, and the maximum value of the infrared absorption peak in this range appears at 1030 (±20) cm?1.

Abstract: A method of reconstructing time-varying position of individual radioactive point sources directly from Positron Emission Tomography (PET) measurements is provided that includes using a PET scanner to acquire list-mode coincidence events of a moving radioactive point source, using an appropriately programmed computer to model a trajectory of the moving radioactive point source as a 3D function of a temporal variable, then apply an optimization procedure to find the trajectory that minimizes a distance between the trajectory and the recorded list-mode coincidence events, and using the PET scanner to output a real time position of the radioactive point source.

Abstract: Systems and methods for separation or isolation of technetium radioisotopes from aqueous solutions of radioactive or non-radioactive molybdate salts using a polyalkyl glycol-based cross-linked polyether polymer. Some embodiments can be used for the effective purification of radioactive technetium-99m produced from low specific activity 99Mo.

Abstract: Disclosed is an exhaust gas purification catalyst carrier which includes a phosphate salt represented by formula: MPO4 (wherein M represents Y, La, or Al) or a zirconium phosphate represented by formula ZrP2O7; an exhaust gas purification catalyst containing a noble metal at least containing Rh and supported on the carrier; and an exhaust gas purification catalyst product having a catalyst support made of a ceramic or metallic material, and a layer of the exhaust gas purification catalyst, the layer being supported on the catalyst support.

Abstract: In various embodiments, a method for coating a workpiece is provided. The method may include drying a workpiece, the workpiece being coated with at least one oxide layer as an uppermost layer; depositing a dielectric layer over the uppermost layer of the dried workpiece; wherein the workpiece is continuously subject to a pressure which is lower than atmospheric pressure during the drying process and during the depositing process.

Abstract: A system for measuring asphaltene content of crude oil, includes a microfluidic chip, the microfluidic chip having a crude oil sample inlet port, a solvent port, a mixer and reactor section in fluid communication with the crude oil sample inlet port and the solvent port, and a filter in fluid communication with the mixer and reactor section, the filter having an inlet side and an outlet side, a waste port in fluid communication with the inlet side of the filter, and a product port in fluid communication with the outlet side of the filter. The system further includes an optical cell in fluid communication with the product port.

Abstract: The invention relates to a method of producing radionuclides. According to the method, a target medium comprising at least a target nuclide material is irradiated in an irradiation zone with neutron irradiation. Radionuclides form in the target nuclide material as a result of the irradiation, and at least some of the formed radionuclides are ejected from the target nuclide material. The ejected radionuclides are then captured and collected in a carbon-based recoil capture material which does not have an empty cage structure at crystallographic level.

Abstract: A method is described for obtaining purified 68Ga. The method comprises eluting 68Ga from a sorbent in and/or on which are sorbed 68Ga and 68Ge, so as to generate a crude 68Ga solution. The sorbent should have a higher affinity for 68Ge than for 68Ga. The crude 68Ga solution is then applied to a medium cation exchange resin, and the resin is eluted with an aqueous alcohol so as to retain the 68Ga on the resin and remove unwanted species from the resin. The resin is then eluted with an eluent, which may be an acidic solution, an alkaline solution or a solution of a species capable of complexing 68Ga ions, so as to obtain an eluate comprising purified 68Ga ions.

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: This invention relates to a method for making strontium-phosphate microparticles that incorporate radioisotopes for radiation therapy and imaging.

Abstract: This invention relates to strontium-phosphate microparticles that incorporate radioisotopes for radiation therapy and imaging.

Abstract: The invention provides the Magnetoelectric Effect Material consisted of a single isotope, the alloy of isotopes, or the compound of isotopes. The invention applies enrichment and purification to increase the isotope abundance, to create the density of nuclear exciton by irradiation, and therefore increase the magnetoelectric effect of the crystal of single isotope, the alloy crystal of isotopes and the compound crystal of isotopes. The invention provides the manufacturing method including the selection rules of isotopes, the fabrication processes and the structure of composite materials. The invention belongs to the area of the nuclear science and the improvement of material character. The invention using the transition of entangled multiple photons to achieve the delocalized nuclear exciton. The mix of selected isotopes adjusts the decay lifetime of nuclear exciton and the irradiation efficiency to generate the nuclear exciton.

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 losing 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: Cavitand compositions that comprise void spaces are disclosed. The void spaces may be empty, which means that voids are free of guest molecules or atoms, or the void spaces may comprise guest molecules or atoms that are normally in their gas phase at standard temperature and pressure. These cavitands may be useful for industrial applications, such as the separation or storage of gasses. Novel cavitand compounds are also disclosed.

Abstract: Methods of synthesizing 100Mo2C and 99mTcO4? are disclosed. Methods of 100Mo2C generation involve thermally carburizing 100MoO3. Methods of 99mTcO4 generation involve proton bombardment of 100Mo2C in a cyclotron. Yields of 99mTcO4 can be increased by sintering 100Mo2C prior to bombardment. The methods also include recycling of 100Mo2C to form 100MoO3. SPECT images obtained using 99mTcO4 generated by the disclosed methods are also presented.

Abstract: A system and method for tritium separation systems using a mixed bed catalytic exchange process in a Liquid Phase Catalytic Exchange/Closed Loop Continuous Process (LPCE/CLCP) system, that operates as a low temperature and low pressure continuous balanced process, designed to rapidly, economically and safely extract and isolate isotope specific products without generating unwanted products in the form of new waste streams.

Abstract: A process for preparing a mesoporous metal oxide, i.e., transition metal oxide, Lanthanide metal oxide, a post-transition metal oxide and metalloid oxide. The process comprises providing a micellar solution comprising a metal precursor, an interface modifier, a hydrotropic ion precursor, and a surfactant; and heating the micellar solution at a temperature and for a period of time sufficient to form the mesoporous metal oxide. A mesoporous metal oxide prepared by the above process. A method of controlling nano-sized wall crystallinity and mesoporosity in mesoporous metal oxides. The method comprises providing a micellar solution comprising a metal precursor, an interface modifier, a hydrotropic ion precursor, and a surfactant; and heating the micellar solution at a temperature and for a period of time sufficient to control nano-sized wall crystallinity and mesoporosity in the mesoporous metal oxides. Mesoporous metal oxides and a method of tuning structural properties of mesoporous metal oxides.

Abstract: A process for removal of tritium from materials that are contaminated thereby envisages the use of a detritiation reactor RT, in which the reaction for the removal of tritium from the waste takes place, the waste being recovered by a flow of moist inert gas in which an extremely low percentage of humidity is used. The heated waste releases a current of tritiated gases, the current of gases being removed from the reactor via the moist inert gas, which conveys it into a membrane reactor RM for decontamination. The membrane reactor, in fact, is able to remove selectively the tritium present in the mixture of gases: there is thus the dual advantage of purifying the mixture of gases and of recovering the tritium contained therein.

Abstract: A process for manufacturing magnetic and/or radioactive metal nanoparticles, the process comprising: preparing an electrolyte solution including metal ions and a stabilizer; generating a plasma at an interface of the electrolyte solution at atmospheric pressure; and recovering magnetic and/or radioactive metal nanoparticles. The magnetic metal nanoparticles can comprise magnetoradioactive nanoparticles. The magnetic metal nanoparticles can be used as MRI contrast agents and the magnetoradioactive nanoparticles can also be used as contrast agents and for dual PET/MRI applications. It also relates to a multi-plasma apparatus for synthesizing nanoparticles.

Abstract: Disclosed is a novel adsorbent for use in a 99Mo/99mTc generator, which is a medical diagnostic radioisotope generator, and in a 188W/188Re 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 99Mo or 188W 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 99mTc or 188Re.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: A gamma radiation source comprises 75Selenium wherein the 75Selenium is provided in the form of compounds, alloys or mixtures with one or more nonmetals which upon irradiation do not produce products capable of sustained emission of radiation which would unacceptably interfere with the gamma radiation of 75Selenium. A further gamma radiation source comprises 75Selenium wherein the 75Selenium is provided in the form of compounds, alloys or mixtures with one or more metals or nonmetals, the neutron irradiation of which does produce products capable of sustained emission of radiation which would acceptably complement the gamma radiation of 75Selenium. Further, the gamma radiation source may have components that are separately irradiated before being combined and the components may be of natural isotopic composition or of isotopically modified composition so that the subsequent radiation peaks may also be adjusted in relative frequency.

Abstract: A carbon nanotube array suitable for use in labeling is provided. The carbon nanotube array includes at least two different isotope-doped carbon nanotube sub-arrays. Each isotope-doped carbon nanotube sub-array includes a plurality of carbon nanotubes. The carbon nanotubes in different isotope-doped carbon nanotube sub-arrays are composed of different kinds of carbon isotopes.

Abstract: The invention relates to a detritiation device comprising i) a furnace (1) for melting tritiated waste, said furnace comprising a hearth for receiving tritiated waste and a bubbling device for introducing a hydrogenated bubbling gas into the hearth during the melting and treatment of the tritiated waste in the furnace (1), and ii) a catalytic reactor with a quadrupole membrane (2) for treating the gas resulting from the melting and treatment of the tritiated waste in the furnace (1), said reactor comprising a membrane (20) for separating two flows of gas, the membrane (20) being permeable to the hydrogen isotopes. The invention also relates to an associated detritiation method.

Abstract: The invention generally relates to the preparation of 18F-labeled radiopharmaceuticals. In particular, this invention relates to the advanced processes for an efficient eiution of [18F]fluoride trapped in a cartridge filled with quaternary ammonium polymer which comprises inert non-basic and non-nucleophilic counter anions. The said methods and polymer cartridges allow the rapid preparation of suitable [18F]fluoride solution, which is also less basic to reduce the formation of byproducts, finally to increase radiochemical yield and purity of 18F-radiopharmaceuticals.

Abstract: The present invention is related to a method to obtain reactive [18F] fluorides in an organic medium suitable for radiolabelling without any azeotropic evaporation step, by the use of a solid phase extraction column containing a modified non-ionic solid support.

Abstract: Systems, methods, and processes for a high throughput, low concentration processing of low activity tritiated light water include the electrolysis of at least some of the tritiated water to produce hydrogen and tritium gas. The hydrogen and tritium gas produced by electrolysis in some cases are combined with heated water vapor to increase throughput and passed through a liquid phase catalytic exchange column, which generally includes a catalyst that includes palladium coated with a hydrophobic polymer. As the hydrogen and tritium gas, along with heated water vapor, rise through the LPCE column, the tritium is retained on the catalyst. Deionized wash water passes down the column (i.e., in the opposite direction of the flow of the hydrogen gas and heated water vapor) and carries the retained tritium out of the LPCE column. Useful in separating tritium from radioactive waste materials and from the water from nuclear reactors.

Abstract: A gamma radiation source comprises 75Selenium wherein the 75Selenium is provided in the form of compounds, alloys or mixtures with one or more nonmetals which upon irradiation do not produce products capable of sustained emission of radiation which would unacceptably interfere with the gamma radiation of 75Selenium. A further gamma radiation source comprises 75Selenium wherein the 75Selenium is provided in the form of compounds, alloys or mixtures with one or more metals or nonmetals, the neutron irradiation of which does produce products capable of sustained emission of radiation which would acceptably complement the gamma radiation of 75Selenium. Further, the gamma radiation source may have components that are separately irradiated before being combined and the components may be of natural isotopic composition or of isotopically modified composition so that the subsequent radiation peaks may also be adjusted in relative frequency.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

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 for concentrating nitrogen isotope to obtain a final product nitrogen, in which the stable nitrogen isotope, 15N, is concentrated, by low-temperature distillation of raw material nitrogen containing trace amounts of oxygen and argon using a plurality of distillation columns in a cascade arrangement. An argon-oxygen mixture is discharged from the bottom of the final column and 15N-concentrated nitrogen is extracted from a lower intermediate point of the final column. The flow rate of the argon-oxygen mixture discharged from the bottom of the final column is controlled based on the reading of a thermometer installed below the point where the final product 15N-concentrated nitrogen is extracted, thereby providing a stable nitrogen product.

Abstract: A process for removal of tritium from materials that are contaminated thereby envisages the use of a detritiation reactor RT, in which the reaction for the removal of tritium from the waste takes place, the waste being recovered by a flow of moist inert gas in which an extremely low percentage of humidity is used. The heated waste releases a current of tritiated gases, the current of gases being removed from the reactor via the moist inert gas, which conveys it into a membrane reactor RM for decontamination. The membrane reactor, in fact, is able to remove selectively the tritium present in the mixture of gases: there is thus the dual advantage of purifying the mixture of gases and of recovering the tritium contained therein.

Abstract: Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

Abstract: A method for providing 11C-labeled cyanides from 11C labeled oxides in a target gas stream retrieved from an irradiated high pressure gaseous target containing O2, wherein 11C labeled oxides are reduced with H2 in the presence of a nickel catalyst under a pressure and a temperature sufficient to form a product stream comprising at least about 95% 11CH4, the 11CH4 is then combined with an excess of NH3 in a carrier/reaction stream flowing at an accelerated velocity and the combined 11CH4 carrier/reaction stream is then contacted with a platinum (Pt) catalyst particulate supported on a substantially-chemically-nonreactive heat-stable support at a temperature of at least about 900° C., whereby a product stream comprising at least about 60% H11CN is provided in less than 10 minutes from retrieval of the 11C labeled oxide.

Abstract: A method of drying a radioisotope solution having radioisotopes includes passing the radioisotope solution through a solid phase extraction column containing an anion exchange group, thereby trapping the radioisotopes in the column. The method also includes passing an eluent through the column, thereby removing the radioisotopes from the column. The eluent includes a cation trapping agent/salt complex, less than 4% water, and the remainder is a solvent. A method of producing the eluent includes reacting a cation trapping agent with a salt in the presence of less than 4% water and a solvent to form solubilized cation trapping agent/salt complex, wherein one of the cation trapping agent and the salt is present in en excess of a stoichiometric amount and ending the reaction when a predetermined amount of solubilized cation trapping agent/salt complex has been formed.

Abstract: One gamma radiation source comprises 75Selenium or a precursor thereof, wherein the 75Selenium is provided in the form of one or more thermally stable compounds, alloys or mixtures with one or more nonmetals which upon irradiation do not produce products capable of sustained emission of radiation which would unacceptably interfere with the gamma radiation of 75Selenium. A further gamma radiation source comprises 75Selenium or a precursor thereof, wherein the 75Selenium is provided in the form of one or more thermally stable compounds, alloys or mixtures with one or more metals or nonmetals, the neutron irradiation of which does produce products capable of sustained emission of radiation which would acceptably complement the gamma radiation of 75Selenium.

Abstract: A method is presented for producing and isolating an isotopically enriched compound of a desired isotope from a parent radionuclide. The method includes forming, or placing, a precipitate containing a parent radionuclide of the desired daughter isotope in a first reaction zone and allowing sufficient time for the parent to decay into the desired gaseous daughter radioisotope. The method further contemplates collecting the desired daughter isotope as a solid in a second reaction zone through the application of temperatures below the freezing point of the desired isotope to a second reaction zone that is connected to the first reaction zone. Specifically, a method is presented for producing isotopically enriched compounds of xenon, including the radioactive isotope Xe-131m and the stable isotope Xe-131.

Abstract: The present disclosure relates to a compound comprising an oxygen-15 and a process for preparation thereof. It also relates to the use of the compound in positron and/or other nuclide imaging and use of the compound in obtaining a perfusion and/or metabolic image in an animal and/or human body. A process for preparing the compound comprises irradiating a compound comprising oxygen via irradiation energy in the range of 10 MeV to 430 MeV. The oxygen atom in the compound may then be allowed to be converted to an oxygen-15 positron nuclide through a photonuclear reaction. Provided that the molecular structure of the irradiated compound is not disrupted, a compound comprising an oxygen-15 is prepared.

Abstract: An isotope-doped nano-structure is provided. The isotope-doped nano-structure includes at least one isotope-doped nano-structure segment having at least two isotopes of the element. The at least two isotopes of the element are mixed uniformly in a certain proportion. The isotope-doped nano-structure can be used for isotope labeling one type of the unlabeled structures such as DNAs, proteins, glucoses, gluconic acids, starches, biotin enzymes, sorbitols, or organic amines. An isotope labeled structure is also provided.

Abstract: Disclosed is a novel adsorbent for use in a 99Mo/99mTc generator, which is a medical diagnostic radioisotope generator, and in a 188W/188Re 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 99Mo or 188W 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 99mTc or 188Re.

Abstract: Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI2, CuI, or Bi5O7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425° C. to 550° C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500° C. (below the silver iodide sublimation temperature of 500° C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

Abstract: A process comprises (a) combining (1) at least one base and (2) at least one metal carboxylate salt comprising (i) a metal cation selected from metal cations that form amphoteric metal oxides or oxyhydroxides and (ii) a carboxylate anion comprising from one to four alkyleneoxy moieties, or metal carboxylate salt precursors comprising (i) at least one metal salt comprising the metal cation and a non-interfering anion and (ii) at least one carboxylic acid comprising from one to four alkyleneoxy moieties, at least one salt of the carboxylic acid and a non-interfering, non-metal cation, or a mixture thereof; and (b) allowing the base and the metal carboxylate salt or metal carboxylate salt precursors to react.

Abstract: This disclosure concerns a new method for preparing radioisotopes, such as molybdenum-99, by alpha particle irradiation, such as by alpha particle irradiation of zirconium-96. Molybdenum-99 is a precursor to the medically-significant radioisotope technetium-99m. Also disclosed are novel compositions containing one or more of technetium-99m, molybdenum-99 and zirconium species. Systems for producing molybdenum-99 and technetium-99m, including alpha particle generators and irradiation targets, also are described.

Abstract: Disclosed in a catalyst which enables to reduce the carbon monoxide concentration in a product gas to 5 ppm by volume or less when carbon monoxide in a raw material gas containing hydrogen and carbon monoxide is selectively oxidized. The catalyst comprises a support of an inorganic oxide and ruthenium loaded thereon, and the relative loading depth X(Ru) of ruthenium in the radial direction in a redial cross-section of the catalyst satisfies the requirement defined by the following formula (1) X(Ru)?15??(1).

Abstract: The present invention relates to SiC nanostructures, including SiC nanopowder, SiC nanowires, and composites of SiC nanopowder and nanowires, which can be used as catalyst supports in membrane electrode assemblies and in fuel cells. The present invention also relates to composite catalyst supports comprising nanopowder and one or more inorganic nanowires for a membrane electrode assembly.

Abstract: The present invention relates to SiC nanostructures, including SiC nanopowder, SiC nanowires, and composites of SiC nanopowder and nanowires, which can be used as catalyst supports in membrane electrode assemblies and in fuel cells. The present invention also relates to composite catalyst supports comprising nanopowder and one or more inorganic nanowires for a membrane electrode assembly.

Abstract: The present invention relates to SiC nanostructures, including SiC nanopowder, SiC nanowires, and composites of SiC nanopowder and nanowires, which can be used as catalyst supports in membrane electrode assemblies and in fuel cells. The present invention also relates to composite catalyst supports comprising nanopowder and one or more inorganic nanowires for a membrane electrode assembly.