35results about "Nuclear fuel reprocessing" patented technology

A metal particulate fuel system is described. The metal fuel system may include particulate metal fuel for use in nuclear reactors. The particulate metal fuel may include a plurality of particles of at least one enriched alloy where the particles are compacted into a fuel column. The metal particulate fuel system may also include a cladding and / or a gas-filled plenum.

Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium / transuranic product recovery and fission product immobilization in engineered waste forms.

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.

A method of fabricating metallic fuel from surplus plutonium may include combining plutonium oxide powder and uranium oxide powder to obtain a mixed powder with reduced proliferation potential. The mixed powder may be electroreduced in a bath of molten salt so as to convert the mixed powder to a first alloy. The first alloy may be pressed to remove a majority of the molten salt adhered to the first alloy to form a pressed alloy-salt mixture. The first alloy may be isolated from the salt by melting the pressed alloy-salt mixture. The first alloy may be further processed to fabricate a fuel rod. Accordingly, the metallic fuel produced may be used in a fast reactor system, such as a Power Reactor Innovative Small Module (PRISM).

The present invention provides a polymeric extractant with pendant chelator groups for selective sequestration of actinides and / or lanthanides comprising: a first block polymer comprising one or more CMPO monomers each having a first backbone bound to a pendant carbomylmethylphosphine oxide group; a second block polymer comprising one or more second monomers each having a second backbone bound to a second pendant group, wherein the first backbone is polymerized to the second backbone to form a polymer backbone with pendent blocks of the second pendant group and pendent blocks of pendent carbomylmethylphosphine oxide group that sequester the actinides and / or lanthanides; and optionally a third block polymer comprising one or more third monomers each having a third backbone bound to a third pendant group, wherein the one or more third monomers is polymerized to the first backbone and the second backbone.

The present invention relates to a method for processing a nuclear fuel comprising a fissile material, SiC and possibly carbon, said method comprising the contacting of said fuel with a chlorine / oxygen mixture at a temperature below 950° C., and more particularly at a temperature between 400 and 900° C., so as to remove the SiC, and the carbon if this is present, from said fuel. The method of the invention makes it possible for example to declad TRISO or BISO nuclear fuel particles, i.e. particles enabling the nuclear fuel to be confined in a sheath or cladding, or to remove an SiC matrix from a fuel having a heterogeneous SiC matrix. The present invention therefore has many applications, especially in the reprocessing of irradiated nuclear fuels.

The present invention provides a redox mediated polymer composition for the selective sequestration and separation of multivalent metal ions comprising: 2 or more monomers polymerized to form a polymeric thiophene backbone, wherein each of the 2 or more monomers comprise 2 thiophene portion groups for polymerization and 2 pendant carbomylmethyl-phosphine oxide or diglycolamide groups that sequester selectively the multrivalent metal ions.

A separation and recycling method for recycling uranium and fluoride from a waste liquid sequentially and separately is disclosed. The method comprises a uranium-recycling process and a fluoride-recycling process. In the uranium-recycling process, an alkali metal compound or monovalent cation and a coagulant aid are added into the waste liquid to promote the precipitation of uranium. In the fluoride-recycling process, an alkaline earth metal compound, a strong acid and a coagulant aid are added into the uranium-removed waste liquid to precipitate fluoride. By means of the method of the present invention, the uranium and fluoride contents of the uranium-removed and fluoride-removed waste liquid are compliant with the effluent standards of the environmental laws.

The method concerns processing irradiated (spent) nuclear fuel (SNF), it is primarily aimed at isolating and trapping tritium, and can be used in nuclear power industry for treating SNF. This method provides for a two-phase voloxidation of a reaction mass using gas-air mixture, the reaction mass including fragmented uranium dioxide SNF elements with containers. The first phase is carried out at 400-650° C. in the presence of air and additional carbon dioxide. The second phase is carried out at 350-450° C. using a stream of an air-vapor mixture that can be oxygen-enriched. Both phases are carried out with a repeated mechanical activation of the reaction mass. Provided in the course of the voloxidation is the gas replacement at the hour rate of about 10-50 fold the reaction chamber gas volume. Before being introduced into the reaction chamber, the gas is preheated up to the chamber internal temperature.

The invention pertains to a process for separating at least one first chemical element E1 from at least one second chemical element E2 coexisting in a mixture in the form of oxides, comprising the following steps:a) a step to solubilise a powder of one or more oxides of the said at least one first chemical element E1 and a powder of one or more oxides of the said at least one second chemical element E2 in a medium comprising at least one molten salt of formula MF—AlF3 wherein M is an alkaline element, after which there results after this step a mixture comprising the said molten salt, a fluoride of the said at least one first chemical elements E1 and a fluoride of the said at least one second chemical element E2;b) a step to contact the mixture resulting from step a) with a medium comprising a metal in the liquid state, the said metal being a reducing agent capable of predominantly reducing the said at least one first chemical element E1 relative to the said at least one second chemical element E2, after which there results after this step a two-phase medium comprising a first phase called metal phase comprising the said at least one first chemical element E1 in oxidation state 0, and a second phase called saline phase comprising the molten salt of above-mentioned formula MF—AlF3 and a fluoride of the said at least one second chemical element E2.

A method for the treatment of an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, allowing the uranium and plutonium contained in the solution to be extracted, separated and decontaminated in a single cycle, without requiring any operation involving a reductive stripping of plutonium. Applications for the method include the processing of uranium-based and / or plutonium-based spent nuclear fuels.

The novel lipophilic metal extractants of the class dialkyldiaza-tetraalkyloctanediamide (DADA) useful to selectively separate trivalent americium (sup. 241 Am) from trivalent lanthanides are represented by the formula 1:Wherein R is a C1 to C5 normal alkyl and R′ is a C4 to C8 normal and branched alkyl group. The compounds are synthesized at high yield and purity by the reaction of corresponding N,N′-dialkylethylenediamine and N,N-dialkyl-2-chloroacetamide. The separation is achieved by utilizing the soft-soft interaction between the trivalent actinides and ‘N’ atoms of the extractant. Both soft donor “n” and hard donor ‘O’ sites are incorporated in the molecule for better extraction of trivalent actinides over trivalent lanthanides. Thus, this molecule can be used as selective extractant to separate trivalent actinides from trivalent lanthanides.

A method using diglycolamide for increasing the separation factor between americium and curium and / or between lanthanides during an extraction operation. The operation comprising putting an acid aqueous phase, in which are found the americium, curium and / or lanthanides, in contact with an organic phase non-miscible with water, containing at least one extractant in an organic diluent. The aqueous and organic phases are then separated, and the diglycolamide is added to the aqueous phase. This method can be used for processing and recycling irradiated nuclear fuels, in particular for selectively recovering americium from high activity aqueous solutions such as raffinates stemming from the processing of irradiated nuclear fuels with a PUREX or COEX™ method; processing of rare earth ores of the monazite, xenotime or bastnaesite type, in order to facilitate separation of <<lightweight>> rare earths from <<heavy>> rare earths and of yttrium, or that of two rare earths with adjacent or close atomic numbers.

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.