The invention is related to nuclear technologies, in particular, to the technology of producing nuclear oxide fuel for fuel elements, this oxide fuel can be used for manufacturing palletized nuclear fuel from uranium dioxide to be consumed by NPPs. The essence of the invention: this method of producing palletized nuclear fuel from uranium dioxide involves preparation of uranium dioxide moulding powder with/without uranium oxide, at this point powdered uranium dioxide is used as a raw material for preparation of moulding powder. Powdered uranium dioxide should be in the following proportion: O/U=2.370.04, it is obtained using a renowned methodby air heating of powdered uranium dioxide (ceramic grade) with the following proportion O/U=2.012.15. The technical result of the invention is increased mechanical strength of sintered pellets and a larger grain size of sintered pellets. 3 appendices, 1 diagram, 2 figures

A process for recovering uranium from components contaminated with uranium oxide includes providing a cleaning apparatus with a cleaning solution for dissolving the uranium oxide of the components, carrying out a cleaning process by introducing a batch of components into the cleaning apparatus, and carrying out a measurement for determining the uranium content of the components. The cleaning and the measuring are repeated if a limit value for the uranium content is exceeded. The components are discharged from the process if the uranium content falls below a limit value. The cleaning is carried out on a plurality of successive batches of components until a control measurement indicates an unsatisfactory cleaning action of the cleaning solution. The uranium oxide dissolved in the cleaning solution is recovered after indication of the unsatisfactory cleaning action.

Electrical devices operating in a range of 273 C. to 100 C. are disclosed. The devices include an insulating substrate. A U0.sub.2+x crystal or oriented crystal U0.sub.2+x film is on a first portion of the substrate. The U0.sub.2+x crystal or film originates and hosts a non-equilibrium polaronic quantum phase-condensate. A first lead on a second portion of the substrate is in electrical contact with the U0.sub.2+x crystal or film. A second lead on a third portion of the surface is in electrical contact with the U0.sub.2+x crystal or film. The leads are isolated from each other. A U0.sub.2+x excitation source is in operable communication with the UO.sub.2+x crystal or film. The source is configured to polarize a region of the crystal or film thereby activating the non-equilibrium quantum phase-condensate. One source state causes the UO.sub.2+x crystal or film to be conducting. Another source state causes the U0.sub.2+x crystal or film to be non-conductive.

In alternative embodiments, the invention provides processes and methods for the recovery, removal or extracting of, and subsequent purification of uranium from a wet-process phosphoric acid using a continuous ion exchange processing approach, where the uranium is recovered from a phosphoric acid, or a phos-acid feedstock using either a dual or a single stage extraction methodology. In both cases an intermediate ammonium uranyl-tricarbonate solution is formed. In alternative embodiments, in the dual cycle approach, this solution is contacted in a second continuous ion exchange system with a strong anion exchange resin then subsequently recovered as an acidic uranyl solution that is further treated to produce an intermediate uranyl peroxide compound which is ultimately calcined to produce the final uranium oxide product. In alternative embodiments, in the single cycle case, the intermediate ammonium uranyl-tricarbonate solution is evaporated to decompose the ammonium carbonate and produce an intermediate uranium carbonate/oxide solid material. These solids are digested in an acid medium, and then processed in the same manner as the secondary regeneration solution from the dual cycle process to produce an intermediate uranyl peroxide that is calcined to produce a final uranium oxide product.

In alternative embodiments, the invention provides processes and methods for the recovery, removal or extracting of, and subsequent purification of uranium from a wet-process phosphoric acid using a continuous ion exchange processing approach, where the uranium is recovered from a phosphoric acid, or a phos-acid feedstock using either a dual or a single stage extraction methodology. In both cases an intermediate ammonium uranyl-tricarbonate solution is formed. In alternative embodiments, in the dual cycle approach, this solution is contacted in a second continuous ion exchange system with a strong anion exchange resin then subsequently recovered as an acidic uranyl solution that is further treated to produce an intermediate uranyl peroxide compound which is ultimately calcined to produce the final uranium oxide product. In alternative embodiments, in the single cycle case, the intermediate ammonium uranyl-tricarbonate solution is evaporated to decompose the ammonium carbonate and produce an intermediate uranium carbonate/oxide solid material. These solids are digested in an acid medium, and then processed in the same manner as the secondary regeneration solution from the dual cycle process to produce an intermediate uranyl peroxide that is calcined to produce a final uranium oxide product.

A method to activate U.sub.3O.sub.8 for conversion of this uranium oxide to hydrated UO.sub.4 via reaction with hydrogen peroxide H.sub.2O.sub.2, wherein the following successive steps are performed: a) an aqueous suspension is prepared containing a powder of U.sub.3O.sub.8 and hydrogen peroxide; b) the aqueous suspension containing a powder of U.sub.3O.sub.8 and hydrogen peroxide is contacted with ozone, whereby an aqueous suspension is obtained of a powder of activated U.sub.3O.sub.8; c) optionally the powder of activated U.sub.3O.sub.8 is separated from the aqueous suspension. A method to convert U.sub.3O.sub.8 to hydrated UO.sub.4 of formula UO.sub.4, nH.sub.2O where n is 2 or 4, comprising at least one step at which hydrogen peroxide H.sub.2O.sub.2 is added to the aqueous suspension of a powder of activated U.sub.3O.sub.8 obtained at the end of step b) of the activation method or to an aqueous suspension prepared by placing in suspension in water the powder of activated U.sub.3O.sub.8 obtained at the end of step c) of the activation method.

A method to activate U.sub.3O.sub.8 for conversion of this uranium oxide to hydrated UO.sub.4 via reaction with hydrogen peroxide H.sub.2O.sub.2, wherein the following successive steps are performed: a) an aqueous suspension is prepared containing a powder of U.sub.3O.sub.8 and hydrogen peroxide; b) the aqueous suspension containing a powder of U.sub.3O.sub.8 and hydrogen peroxide is contacted with ozone, whereby an aqueous suspension is obtained of a powder of activated U.sub.3O.sub.8; c) optionally the powder of activated U.sub.3O.sub.8 is separated from the aqueous suspension. A method to convert U.sub.3O.sub.8 to hydrated UO.sub.4 of formula UO.sub.4, nH.sub.2O where n is 2 or 4, comprising at least one step at which hydrogen peroxide H.sub.2O.sub.2 is added to the aqueous suspension of a powder of activated U.sub.3O.sub.8 obtained at the end of step b) of the activation method or to an aqueous suspension prepared by placing in suspension in water the powder of activated U.sub.3O.sub.8 obtained at the end of step c) of the activation method.

In alternative embodiments, the invention provides processes and methods for the recovery, removal or extracting of, and subsequent purification of uranium from a wet-process phosphoric acid using a continuous ion exchange processing approach, where the uranium is recovered from a phosphoric acid, or a phos-acid feedstock using either a dual or a single stage extraction methodology. In both cases an intermediate ammonium uranyl-tricarbonate solution is formed. In alternative embodiments, in the dual cycle approach, this solution is contacted in a second continuous ion exchange system with a strong anion exchange resin then subsequently recovered as an acidic uranyl solution that is further treated to produce an intermediate uranyl peroxide compound which is ultimately calcined to produce the final uranium oxide product. In alternative embodiments, in the single cycle case, the intermediate ammonium uranyl-tricarbonate solution is evaporated to decompose the ammonium carbonate and produce an intermediate uranium carbonate/oxide solid material. These solids are digested in an acid medium, and then processed in the same manner as the secondary regeneration solution from the dual cycle process to produce an intermediate uranyl peroxide that is calcined to produce a final uranium oxide product.

In alternative embodiments, the invention provides processes and methods for the recovery, removal or extracting of, and subsequent purification of uranium from a wet-process phosphoric acid using a continuous ion exchange processing approach, where the uranium is recovered from a phosphoric acid, or a phos-acid feedstock using either a dual or a single stage extraction methodology. In both cases an intermediate ammonium uranyl-tricarbonate solution is formed. In alternative embodiments, in the dual cycle approach, this solution is contacted in a second continuous ion exchange system with a strong anion exchange resin then subsequently recovered as an acidic uranyl solution that is further treated to produce an intermediate uranyl peroxide compound which is ultimately calcined to produce the final uranium oxide product. In alternative embodiments, in the single cycle case, the intermediate ammonium uranyl-tricarbonate solution is evaporated to decompose the ammonium carbonate and produce an intermediate uranium carbonate/oxide solid material. These solids are digested in an acid medium, and then processed in the same manner as the secondary regeneration solution from the dual cycle process to produce an intermediate uranyl peroxide that is calcined to produce a final uranium oxide product.

An apparatus (1) for thermal denitration of a uranyl nitrate hydrate to uranium trioxide UO3. The apparatus (1) comprises a burner (114) and a reaction chamber (110) configured to carry out thermal denitration of uranyl nitrate hydrate and to form uranium trioxide UO3 in the form of particles. The apparatus also comprises a separating chamber (120) suitable for separating UO3 particles from the gases resulting from the thermal denitration carried out in the reaction chamber (110), and at least one filter (130) configured for purifying the gases. The separating chamber (120) is a decanting chamber into which the reaction chamber (110) directly opens out. The filter (130) is capable of performing the separation at a temperature greater than or equal to 350 C. The invention also relates to use of such an apparatus, to a thermal denitration process and to UO3 particles obtained by such a process.