344 results about "Spent nuclear fuel" patented technology

A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

A canister apparatus, basket apparatus and combinations thereof for transporting and/or storing high level radioactive waste, such as spent nuclear fuel. The canister apparatus comprises a cavity for receiving the spent nuclear fuel that is surrounded by two independent gas-tight containment boundaries. The structures that form the two independent gas-tight containment boundaries are in substantially continuous surface contact with one another, thereby facilitating sufficient heat removal from the cavity. In another aspect, the invention is a basket apparatus having a plurality of disk-like grates arranged in a stacked and spaced arrangement so that the cells of the disk-like grates are aligned. In still another aspect, the invention can be a basket apparatus having a disk-like grate having a ring-like structure encompassing a gridwork of beams specially arranged to achieve a unique cell configuration.

The invention relates to a nuclear fuel cladding totally or partially made of a composite material with a ceramic matrix containing silicon carbide (SiC) fibers as a matrix reinforcement and an interphase layer provided between said matrix and said fibers, the matrix including at least one carbide selected from titanium carbide (TiC), zirconium carbide (ZrC), or ternary titanium silicon carbide (Ti₃SiC₂).

When irradiated and at temperatures of between 800° C. and 1200° C., said cladding can mechanically maintain the nuclear fuel within the cladding while enabling optimal thermal-energy transfer towards the coolant.

The invention also relates to a method for making the nuclear fuel cladding.

The invention relates to the separation of lanthanides and actinides by nanofiltration complexation. The object of the invention is to satisfy the existing need for a simple, efficient and economical technique for separating lanthanides and actinides. This object is achieved by a process consisting of using ligands of the polyamino acid type, such as EDTA or DTPA, for complexing lanthanides and/or actinides before separating them by nanofiltration. The invention further relates to novel polyamino acid ligands incorporating ligand structures additional to EDTA and DTPA. Application to the production of rare earths or nuclear waste processing, especially to recycling operations carried out on spent nuclear fuels is also discussed.

An apparatus and method for inter-unit transfer of spent nuclear fuel. In one aspect, the invention is a method of transferring high level radioactive waste comprising: a) loading high level radioactive waste into a water-filled cavity of a canister body having an open top end at a first location; b) coupling a lid to the canister body to enclose the open top end; c) removing a volume of water from the cavity so that a water level of the water within the cavity is above a top end of the high level radioactive waste and a space exists between the water level and a bottom surface of the lid; d) hermetically sealing the cavity; and e) transferring the canister to a second location, the water level remaining above the top end of the high level radioactive waste during the transfer.

Methods and compositions to extract radionuclides such as various actinides and lanthanides from organic and/or aqueous solutions by utilizing extractant functionalized carbon nanotubes are disclosed. More particularly, phosphorous-containing (such as phosphine oxides, phosphoric acids or phosphates) organic extractants and other predesigned extractants (such as crown ethers, calncrown derivatives, malonamide and diglycolamide derivatives, polyethylene glycol derivatives, cobalt dicarbollite derivatives, and N-donating heterocyclic ligands) can be covalently and/or non-covalently employed on the surfaces and/or ends (tips) of carbon nanotubes for the purpose of removal radionuclides such as various actinides and lanthanides from organic and/or aqueous solutions. Extractant functionalized carbon nanotubes can be used for extracting radioactive nuclides from nuclear waste or spent nuclear fuel, which are produced and/or reprocessed from the nuclear power generation or other nuclear application. The invention also relates to the solid-liquid separation process based on the contact of liquid radioactive waste by using the invention extracting agents as the solid phase.

Disclosed is a method for manufacturing a tube and a sheet of niobium-containing zirconium alloys for the high burn-up nuclear fuel. The method comprises melting Nb-added zirconium alloy to ingot; forging the ingot at beta phase range; beta-quenching the forged ingot after solution heat-treatment at 1015-1075° C.; hot-working the quenched ingot at 600-650° C.; cold-working the hot-worked ingot in three to five passes, with intermediate vacuum annealing; and final vacuum annealing the cold-worked ingot at 440-600° C., wherein temperatures of intermediate vacuum annealing and final vacuum annealing after beta-quenching are changed so as to attain the condition under which precipitates in the alloy matrix are limited to an average diameter of 80 nm or smaller and the accumulated annealing parameter (SIGMA A) is limited to 1.0x10-18 hr or lower.

The invention relates to a FeCrAl base alloy material for a nuclear fuel cladding material. With the total weight of the FeCrAl base alloy material as a standard, the alloy material comprises 6%-16% of Cr, 3%-8% of Al, 0.001%-1% of Y, 0.1%-4% of Mo, 0.01%-0.5% of Si, 0.001%-0.5% of C, less than or equal to 500 ppm of N, less than or equal to 1000 ppm of O, less than or equal to 500 ppm of P, less than or equal to 500 ppm of S and the balance Fe. A FeCrAl base alloy shows the excellent antioxygenic property in the 1200 DEG C vapor environment, the excellent accident fault-tolerant ability is achieved and is obviously superior to that of a Zr base alloy, the machining performance is good, and the FeCrAl base alloy material can be used as core structural materials such as fuel element cladding, fuel element composite cladding, fuel cladding coatings and positioning grillage stripes in nuclear power station pressurized water reactors.

Disclosed is a process for recovery of uranium from a spent nuclear fuel using a carbonate solution, characterized by excellent proliferation resistance of preventing leaching of transuranium element (TRU) nuclides such as Pu, Np, Am, Cm, etc. from the spent nuclear fuel as well as environmental friendliness of minimizing waste generation, wherein a highly alkaline carbonate solution is used to separate uranium alone from the spent nuclear fuel.

The invention relates to a process for reprocessing a spent nuclear fuel and for preparing a mixed uranium-plutonium oxide, which process comprises: a) the separation of the uranium and plutonium from the fission products, the americium and the curium that are present in an aqueous nitric solution resulting from the dissolution of the fuel in nitric acid, this step including at least one operation of coextracting the uranium and plutonium from said solution by a solvent phase; b) the partition of the coextracted uranium and plutonium to a first aqueous phase containing plutonium and uranium, and a second aqueous phase containing uranium but no plutonium; c) the purification of the plutonium and uranium that are present in the first aqueous phase; and d) a step of coconverting the plutonium and uranium to a mixed uranium/plutonium oxide. Applications: reprocessing of nuclear fuels based on uranium oxide or on mixed uranium-plutonium oxide.

Various embodiments of a nuclear fuel assembly and related methods for processing and managing spent nuclear fuel are disclosed. According to one exemplary embodiment, a nuclear fuel may include a plurality of first fuel rods having a plurality of first fuel elements and a plurality of second fuel rods having a plurality of second fuel elements. Each of the first fuel elements may include uranium dioxide fuel, and each of the second fuel elements may include a plurality of tristructural isotropic fuel particles embedded in a silicon carbide matrix. The plurality of first fuel rods and the plurality of second fuel rods are arranged in a fuel assembly.

The invention relates to a structure and a preparation method of a nuclear fuel cladding tube with a crack expansion resisting capability. The structure comprises a two-layered or three-layered SiC-based nuclear fuel cladding tube, and is characterized in that an inner layer of the SiC-based nuclear fuel cladding tube is used as a secondary inner layer structure; one layer of a crack expansion resisting layer is introduced into the interior of a secondary inner layer to be used as an inner layer structure, so that a three-layered or four-layered SiC-based nuclear fuel cladding tube is formed; the modulus of the material of the introduced inner layer structure is lower than that of an anti-irradiation material with low modulus of a SiC material of the secondary inner layer. By virtue of the three-layered or four-layered SiC-based nuclear fuel cladding tube, the effect of preventing cracks from being formed on a SiC ceramic layer of the inner wall of the nuclear fuel cladding tube.

A transfer cask for maximizing the radiation shielding for spent nuclear fuel during cask transfer procedures has a cylindrical inner shell forming a cavity within which a spent nuclear fuel canister can be placed; a cylindrical outer shell concentric with and surrounding the inner shell to form an annulus with the inner shell, the annulus adapted for receiving gamma absorbing material; a jacket shell concentric with and surrounding the second shell to form a jacket for holding a neutron absorbing liquid; the jacket shell having filling and drainage systems; and a removable bottom lid so that a canister can be lowered from the cavity into a transport cask or permanent storage cask.

A process of producing ceramic-ceramic composites, including but not limited to nuclear fuels, and composites capable of exhibiting increased thermal conductivities. The process includes milling a first ceramic material to produce a powder of spheroidized particles of the first ceramic material, and then co-milling particles of a second ceramic material with the spheroidized particles of the first ceramic material to cause the particles of the second ceramic material to form a coating on the spheroidized particles of the first material. The spheroidized particles coated with the particles of the second ceramic material are then compacted and sintered to form the ceramic-ceramic composite, in which the second ceramic material forms a continuous phase completely surrounding the spheroidized particles of the first ceramic material.

This disclosure describes various configurations and components of a molten fuel fast or thermal nuclear reactor for managing the operating temperature in the reactor core. The disclosure includes various configurations of direct reactor auxiliary cooling system (DRACS) heat exchangers and primary heat exchangers as well as descriptions of improved flow paths for nuclear fuel, primary coolant and DRACS coolant through the reactor components.

A improved device and process for recycling spent nuclear fuels, in particular uranium metal, that facilitates the refinement and recovery of uranium metal from spent metallic nuclear fuels. The electrorefiner device comprises two anodes in predetermined spatial relation to a cathode. The anodese have separate current and voltage controls. A much higher voltage than normal for the electrorefining process is applied to the second anode, thereby facilitating oxidization of uranium (III), U⁺, to uranium (IV), U⁺⁴. The current path from the second anode to the cathode is physically shorter than the similar current path from the second anode to the spent nuclear fuel contained in a first anode shaped as a basket. The resulting U⁺⁴ oxidizes and solubilizes rough uranium deposited on the surface of the cathode. A softer uranium metal surface is left on the cathode and is more readily removed by a scraper.

Disclosed are methods for transporting and canistering nuclear spent fuel. In one embodiment, the method for transporting and canistering nuclear spent fuel comprises providing a canister that includes a mechanical closure for a lid on a container. The canister containing the nuclear spent fuel is loaded into a transport cask, and the transport cask is sealed at the power plant. It is preferred that the lid is not installed on the canister at the power plant. The transport cask containing the canister and nuclear spent fuel is transported to a remote facility. At the remote facility, the transport cask is opened and the lid of the canister is installed using the mechanical closure to seal the canister after the canister is transported to the remote facility. The canister containing the spent nuclear fuel is placed into an overpack, which is stored at the remote facility.

An apparatus for transferring spent nuclear fuel in the form of a cask having a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel, an intermediate shell disposed concentrically around and spaced apart from the inner shell and an outer shell disposed concentrically around and spaced apart from the intermediate shell. A bottom flange is affixed to bottoms of each of the shells, and a bottom lid is removably affixed to the bottom flange. A top flange is affixed to tops of each of the shells, and a top lid is seated on the top flange. An annulus for air flow may be formed between the inner shell and the canister; the bottom lid may include an impact zone including impact absorbing structure; and the top flange may have integrally formed trunnions.