Abstract: An anti-proliferation technique is disclosed to reduce the likelihood of nuclear proliferation due to the use fissionable fuel salts. The technique includes doping the fuel salt with one or more elements (referred to herein as activation dopants) that, upon exposure to neutrons such as would occur in the fuel salt when a reactor is in operation, undergo a nuclear reaction to, directly or indirectly, form highly active “protecting isotopes” (of the same element as the activation dopant or a different element). A sufficient mass of activation dopants is used so that the Figure of Merit (FOM) of the fuel salt is decreased to below 1.0 within some target number of days of fission. This allows the FOM of the fuel salt to be controlled so that the fuel becomes too dangerous to handle before to the creation of a significant amount of weaponizable isotopes.

Abstract: Sodium-tin and sodium-tin-lead compositions have been identified and created that exhibit better reactivity characteristics (i.e., are less reactive) than sodium metal under the same conditions, making these compositions safer alternatives to sodium metal for use as a coolant. These compositions include compositions having at least 90% sodium (Na), from 0-10% lead (Pb) and the balance being tin (Sn).

Abstract: Embodiments of the invention are directed to a method for production of a nuclear fuel pellet by spark plasma sintering (SPS), wherein a fuel pellet with more than 80% TD or more than 90% TD is formed. The SPS can be performed with the imposition of a controlled uniaxial pressure applied at the maximum temperature of the processing to achieve a very high density, in excess of 95% TD, at temperatures of 850 to 1600° C. The formation of a fuel pellet can be carried out in one hour or less. In an embodiment of the invention, a nuclear fuel pellet comprises UO2 and a highly thermally conductive material, such as SiC or diamond.

Abstract: The invention relates to the ROANEX method, which extracts actinides from used nuclear fuel in a single purification cycle. The used nuclear fuel contains actinides, U, Am, Pu, Np. and Cm, and fission products, Cs, Sr and Tc. The fission products are separated first from the used nuclear fuel. The actinides are reduced to their lowest oxidation states and then oxidized to their highest oxidations states. Uranium, Pu and Np move to an organic phase solution and Am and Cm move to a nitrate solution. Uranium, Pu, and Np are stripped from the organic phase solution, and then treated with an oxalic acid to form a precipitate. Americium and Cm are treated with a potassium carbonate solution and Am precipitates. Actinides Am, U, Pu, and Np precipitates are heated in an oven and then blended together to form a mixed oxide fuel of UO2, PuO2, NpO2 and AmO2.

Abstract: A method for preparing a powder of an alloy based on uranium and molybdenum in a metastable ? phase is provided, which comprises: a) putting at least one first reagent selected from uranium oxides and mixtures thereof, uranium fluorides and mixtures thereof, into contact with a second reagent consisting in molybdenum and a third reagent consisting in a reducing metal, the first, second and third reagents being in a divided form; b) reacting the reagents at a temperature?the melting temperature of the third reagent and under an inert atmosphere, whereby this reaction leads to the formation of the alloy comprising uranium and molybdenum in the form of a powder, for which the particles are covered with a reducing metal oxide or fluoride layer; c) cooling the so formed powder at a rate at least equal to 450° C./hour; and d) removing the reducing metal oxide or fluoride layer which covers the particles of the powder of the alloy comprising uranium and molybdenum.

Abstract: This invention relates to a method of preparing nuclear fuel including the step of depositing at least two adjacent series of layers (16, 18) around a kernel (12) of fissile material, each series comprising a layer of pyrolytic carbon (16) contiguous with a layer of silicon carbide (18) and each layer (16, 18) having a thickness of at most (10) micrometers, such that alternate layers of (16, 18) of pyrolytic carbon and silicon carbide are deposited around the kernel (12). The invention extends to a nuclear fuel element (10).

Abstract: The present invention includes a composition of LiF—ThF4—UF4—PuF3 for use as a fuel in a nuclear engine.

Abstract: Neutron source comprising a composite, said composite comprising crystals comprising BeO and AmBe13, and an excess of beryllium, wherein the crystals have an average size of less than 2 microns; the size distribution of the crystals is less than 2 microns; and the beryllium is present in a 7-fold to a 75-fold excess by weight of the amount of AmBe13; and methods of making thereof.

Abstract: The invention refers to a nuclear fuel, a fuel element, a fuel assembly and a method of manufacturing a nuclear fuel. The nuclear fuel is adapted for use in a water cooled nuclear reactor, including light water reactors LWR, such as Boiling Water Reactors BWR and Pressure Water Reactors PWR. The nuclear fuel comprises an uranium-containing compound consisting of UN. The uranium content of the uranium-containing compound comprises less than 10% by weight of the isotope 235U. The nuclear fuel comprises an additive substantially consisting of at least one element, in elementary form or as a compound, selected from the group consisting of Zr, Mo, Si, Al, Nb and U.

Abstract: The present invention includes a composition of LiF—ThF4—UF4—PuF3 for use as a fuel in a nuclear engine.

Abstract: A ceramic waste immobilizing material for the encapsulation of high level radioactive waste (HLW), e.g. resulting from the reprocessing of irradiated nuclear fuel. The ceramic waste immobilising material enables waste ions from at least fission products in irradiated nuclear fuel to be dissolved in substantially solid solution form. The ceramic waste immobilising medium has a matrix comprising phases of hollandite, perovskite and zirconolite in which the waste ions are dissolved. The invention also includes a method of immobilizing HLW from reprocessed nuclear fuel assemblies comprising the steps of mixing a liquor containing the HLW with a precursor material comprising oxides or oxide precursors of at least titanium, calcium and barium to form a slurry, drying the slurry, and calcining the dried slurry under a reducing atmosphere to form a powder comprising 30–65 weight % waste.

Abstract: A non-corrosive cleaning composition for removing residues from a substrate. The composition comprises: (a) water; (b) at least one hydroxylammonium compound; (c) at least one basic compound, preferably selected from the group consisting of amines and quaternary ammonium hydroxides; (d) at least one organic carboxylic acid; and (e) optionally, a polyhydric compound. The pH of the composition is preferably between about 2 to about 6.

Abstract: A nuclear fuel based on UO.sub.2, ThO.sub.2 and/or PuO.sub.2 having improved retention properties for fission products. The fuel comprises a metal such as Cr or Mo able to trap oxygen in order to form an oxide having a free formation enthalpy equal to or below that of the superstoichiometric oxide or oxides (U, Th)O.sub.2+x and/or (U, Pu)O.sub.2+x (O<x.ltoreq.0.01). Thus, it is possible to trap oxygen atoms released during the fission of U, Th and/or Pu. This leads to an increase in the retention level of the fission products and a possibility of obtaining a high burn-up of nuclear fuel elements.

Abstract: Process for the production of nuclear fuel pellets based on mixed uranium and plutonium oxide having a specific plutonium content from a charge of UO.sub.2 and PuO.sub.2 powders by lubrification, pelletizing and sintering, in which a solid, sulphur, organic additive of the zwitterion type is incorporated into the mixture during the co-milling stage for the powders.

Abstract: A uranium-containing nuclear-fuel sintered pellet containing UO.sub.2, (U, Pu)O.sub.2, (U, Th)O.sub.2, (U, RE)O.sub.2, (U, Pu, Th)O.sub.2, (U, Pu, RE)O.sub.2, (U, Th, RE)O.sub.2 or (U, Pu, Th, RE)O.sub.2, wherein RE=rare earth, has a sintered-pellet surface layer being formed of at least 80% by volume of a chemical boron compound UB.sub.x or (U, . . . )B.sub.x, wherein x=2;4;6 or 12, and a remainder of the sintered pellet containing at most 5% by volume of the chemical boron compound. A nuclear-reactor fuel assembly has a fuel rod containing such a uranium-containing nuclear-fuel sintered pellet in a cladding tube with the boron as a burnable absorber for thermal neutrons. The surface layer having the chemical boron compound is obtained by treating the nuclear-fuel sintered pellet with boron or a boron-containing chemical compound at an appropriately high treatment temperature.

Abstract: The present invention relates to a process of decontaminating zirconium-based alloy claddings tubes used in nuclear reactor fuel rods. The process involved the use of a permanganate in a dilute acid solution. The process renders the alloy suitable for uncontrolled release into a non-radioactive environment.

Abstract: A method of preparing active, sinterable, finely-divided plutonium oxide (PuO.sub.2) powder from plutonium metal is disclosed. The process yields plutonium fissile material which can be easily blended to form a uniformly homogeneous powder for the fabrication of high-quality light water reactor ceramic fuel pellets. Such homogeneous fuels are required to prevent hot spots from developing in a reactor using the fuel.

Abstract: A chemical heat source comprising metal nitride, metal oxide and carbon, particularly useful in smoking articles, and methods of making the heat source are provided. The metal nitride of the heat source has an ignition temperature substantially lower than conventional carbonaceous heat sources, while at the same time provides sufficient heat to release a flavored aerosol from a flavor bed for inhalation by the smoker. Upon combustion the heat source produces virtually no carbon monoxide. The metal nitride is prepared by pre-forming the starting materials into a desired shape, and converting them to metal nitride in situ, without substantially altering the shape of the starting materials.

Abstract: To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Abstract: The disclosed invention consists of a ceramic form of fuel for a nuclear reactor comprising an oxide of a fissionable element and containing therein a consumable neutron absorbing agent. The ceramic fissionable fuel product provides longer fuel performance in nuclear reactor service.

Abstract: Process for the purification of reprocessing uranium from which have previously been separated the fission products generated in a nuclear reactor and consisting of eliminating the U232 daughter products appearing during storage, by passing said uranium in its hexafluoride form through a chemically inert porous material.

Abstract: For the production of plate-shaped fuel elements for research reactors having charges of more than 26 volume % of uranium compound in the aluminum matrix according to the known picture frame technique, 0.01 to 0.3 mm thick aluminum layers are applied by rolling to the picture before inserting it into the frame in order to avoid dragging the uranium into the fuel-free zones.

Abstract: The invention discloses a metal-actinide mononitride composition with dimensional stability in extended nuclear reactor operations, with a method of operation at surface temperatures in excess of 1700.degree. C. The preferred embodiment and operating method uses a mononitride of uranium and a metal selected from the group consisting of titanium or yttrium. Parameters for determination of the metal element to stabilize the fuel are disclosed.

Abstract: A method is disclosed for producing a dimensionally stable UO.sub.2 fuel pellet of large grain mole %, and relatively large pore size. A dopant containing an element selected from the group consisting of aluminum, calcium, magnesium, titanium, zirconium, vanadium, niobium, and mixtures thereof is added to a highly sinterable UO.sub.2 powder, which is a UO.sub.2 powder that is sinterable to at least 97% theoretical density at 1600.degree. C. in one hour, and the UO.sub.2 powder can then be formed into fuel pellets. Alternatively, the dopant can be added at a step in the process of producing ammonium diuranate. The ammonium diuranate is collected with about 0.05 to about 1.7 mole%, based on UO.sub.2, of a compound containing said element. That mixture is then calcined to produce UO.sub.2 and the UO.sub.2 is formed into a fuel pellet. The addition of the dopant can also be made at the hydrolysis stage in the manufacture of UO.sub.2 by a dry conversion process.

Abstract: The present invention is directed to a method for making a wide variety of general-purpose cermets and for radioactive waste disposal from ceramic powders prepared from urea-dispersed solutions containing various metal values. The powders are formed into a compact and subjected to a rapid temperature increase in a reducing atmosphere. During this reduction, one or more of the more readily reducible oxides in the compact is reduced to a selected substoichiometric state at a temperature below the eutectic phase for that particular oxide or oxides and then raised to a temperature greater than the eutectic temperature to provide a liquid phase in the compact prior to the reduction of the liquid phase forming oxide to solid metal. This liquid phase forms at a temperature below the melting temperature of the metal and bonds together the remaining particulates in the cermet to form a solid polycrystalline cermet.