175results about "Nuclear elements" patented technology

A present invention provides an internal circulating irradiation capsule available for the production of iodine-125 and a related production method. The irradiation capsule filled with xenon gas has a lower irradiation part, an upper irradiation part, and a neutron control member. The lower irradiation part is inserted into an irradiation hole of a reactor core and irradiated with a large quantity of neutron directly. When neutron is radiated to the xenon gas, iodine-125 is produced from xenon gas. The upper irradiation part protrudes from the irradiation hole, and iodine-125 is transferred to the upper irradiation part by convection and solidified in the upper part. The neutron control member reduces neutron in the upper part to produce iodine-125 of high purity and radioactivity in a large quantity.

In one aspect the invention provides a system for treating a wastestream, particularly a radwaste, for safe disposal and, in final processing converting it into one or both forms including an aqueous form for safe discharge to the environment and a solidified form for safe disposal. In another aspect the invention provides the capacity to employ a step where a specific target element strategy can be set up synchronizing sorbent substance choices and multiple recycle options to remove target substances from wastestream as a part of its Sorption or Powder Sorbent Isotopic Reduction step (II). Other steps cooperate with Sorption step (II) including Oxidation (I), Solid-Liquid separation (III), and Selective Ion Exchange (IV) to deliver the wastestream to final processing.

An externally powered, self cleaning strainer incorporating a projectile shield, which is capable of operating for an extended period of time. A suitably shaped, motor driven, impeller creates a localized, radially outward flow of fluid in the vicinity of the strainer inlet. The projectile shield has a lower surface shaped to deflect fluid to the strainer at a constant velocity, enabling the impeller to eject debris more efficiently. Maintaining a constant flow through the strainer also avoids additional head loss associated with accelerating flow. The self cleaning strainer may also include a brush attached diametrically opposite to the impeller to aid in removing debris from the inlet side of the strainer. The impeller may also be shaped so that when it is swept past the inlet side of the strainer, it causes a localized, reverse flow through the strainer, thereby removing debris particles from within the strainer.

Herein is a method of segregating chemical species contained in spent nuclear reactor fuel without employing conventional acid dissolution. Particularly, pellets of spent fuel are ground to talc sized particles. Heat is added. The preferred heating is by flow through a plasma arc producing micron sized liquid drops suspended in helium flow. The vapor pressure of the chemical species is significantly greater than uranium dioxide. the ultra volatile chemical species evolve from the drops into the helium flow. The gas phase is separated from the mist by a gas/liquid separator (demister). Heavy mist drops of UO₂ impact the walls, coalesce and flow down to the separator drain, becoming legally transportable. Helium flow exhausts from the separator vertically. The gaseous chemical species will condense in sequentially cooler stages and separate from the helium down to the cryogenic temperatures of liquid radioactive xenon and krypton. Non-condensed helium is recycled.

The process of the invention is the separation of minor actinides from lanthanides in a fluid mixture comprising, fission products, lanthanides, minor actinides, rare earth elements, nitric acid and water by addition of an organic chelating aid to the fluid; extracting the fluid with a solvent comprising a first extractant, a second extractant and an organic diluent to form an organic extractant stream and an aqueous raffinate. Scrubbing the organic stream with a dicarboxylic acid and a chelating agent to form a scrubber discharge. The scrubber discharge is stripped with a simple buffering agent and a second chelating agent in the pH range of 2.5 to 6.1 to produce actinide and lanthanide streams and spent organic diluents. The first extractant is selected from bis(2-ethylhexyl)hydrogen phosphate (HDEHP) and mono(2-ethylhexyl)2-ethylhexyl phosphonate (HEH(EHP)) and the second extractant is selected from N,N,N,N-tetra-2-ethylhexyl diglycol amide (TEHDGA) and N,N,N′,N′-tetraoctyl-3-oxapentanediamide (TODGA).

Apparatuses, processes and methods for the separation, isolation, or removal of specific radioactive isotopes from liquid radioactive waste, these processes and methods employing isotope-specific media (ISM). In some embodiments, the processes and methods further include the vitrification of the separated isotopes, generally with the ISM; this isotope-specific vitrification (ISV) is often a step in a larger scheme of preparing the radioactive isotopes for long-term storage or other disposition. A variety of ISM are disclosed.

In accordance with the present invention, there is provided an increased efficiency strainer system which is particularly suited for use in the emergency core cooling system of a nuclear power plant. In certain embodiments of the present invention, the strainer system includes one or more strainer cassettes or cartridges, with each such cassette or cartridge including a plurality of strainer pockets disposed in side-by-side relation to each other. In these embodiments, multiple cassettes or cartridges may be assembled together to form a strainer module of the strainer system. The strainer pockets of the cartridge each define an inflow end. Within the cartridge, or the module including multiple cartridges, the inflow ends of one or more of the strainer pockets may be enclosed by an elastic metal membrane. When in a closed position, the membrane prevents liquid flow into the corresponding strainer pocket via the inflow end thereof. The membrane remains closed when only a low pressure load is exerted thereon, but is deflected or deformed into an open position when a high pressure load is exerted thereon. The movement of the membrane to its open position effectively opens the corresponding strainer pocket, thus allowing for the flow of liquid into the interior of the strainer pocket via the inflow end thereof.

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.

A method for the extraction and purification of molybdenum, the method comprising the steps of: transferring an irradiated fuel solution to an extraction system, the irradiated fuel solution comprising iodine and molybdenum and other fission products, the extraction system comprising at least one sorbent column; passing the irradiated fuel solution upwards through the at least one sorbent-containing extraction column; directing the irradiated fuel solution towards a fuel management system by means of at least one discharge alignment valve; directing the extraction column eluate towards an iodine removal system; removing the iodine from the extraction column eluate; purifying the extraction column eluate; and collecting the purified eluate. Also disclosed is an apparatus for accomplishing the aforementioned method.

A system and method for removing spent fuel assemblies from a fuel building and transporting them to on-site facilities. A cask transporter is moved into the fuel building with an empty spent fuel storage cask, spent fuel assemblies are loaded into spent fuel storage cask, the cask is sealed, and the cask transporter moves the loaded spent fuel storage cask to a handling area for final disposal. Components of the system include a penetration cover, a lifting mechanism, a control system, a valve system, and the cask transporter.

A hydraulic drive type partial inter-tube lancing system for cleaning a steam generator in a nuclear power plant, which removes foreign matter, such as sludge, deposited around heat transfer tubes in the steam generator, is disclosed. More particularly, a hydraulic drive type partial inter-tube lancing system, which serves to clean a steam generator of a light-water reactor in a nuclear power plant by removing sludge, attached inside the steam generator, using high-pressure water, is disclosed.

Disclosed is an offshore, manned, scalable, modular, floating, moored, nuclear power generating plant and multipurpose platform. The present invention is comprised of a Main Plant Control Deck, Central Plant Deck, and submerged Reactor-Generator Deck(s) integrated into the structure of a Spar or Cell Spar Platform. The Reactor-Generator decks are comprised of a plurality of modular, Naval Nuclear Pressurized Water Reactor Modules. Electricity generated is transmitted via submarine High Voltage Direct Current Cables to shore. Ancillary, co-generated services, e.g. desalinated water, are transmitted to shore via submarine pipelines. Multipurpose Topside Decks house vessel command, crew, and any ancillary and co-generation equipment. The present invention, constructed in a multi-path manufacturing process, provides exceptional economic, environmental, sustainability, security, safety, and operational advantages over the current art of power generation.

A method of producing radioisotopes using a heavy water type nuclear power plant is provided. The method includes inserting targets (37) into a heavy water moderator of the heavy water reactor througha guide tube (30) in a port (55) in a reactivity mechanism deck (45) of the heavy water reactor. The heavy water reactor operates to irradiate the target to convert the target into a radioisotope. The method then includes removing the radioisotope from the moderator of the heavy water reactor via the reactivity mechanism deck. A heavy water nuclear reactor is also provided.

A concrete storage module (26) is adapted to slideably receive a cylindrical canister assembly (12) therein. Heat dissipation fins (62) and a tubular heat shield (96) are disposed within the module to help dissipate heat emitted from nuclear fuel assemblies stored in the canister to air flowing through the module. The canister assembly (12) is composed of a basket assembly (70) constructed from multi-layer structural plates disposed in cross-cross or egg carton configuration. A single port tool (106) is provided for draining water from the canister (12) and replacing the drain water with make-up gas. The single port tool is mounted in the cover (100) of the canister and is in fluid flow communication with the interior of the canister.

Process for treating spent nuclear fuel (SNF) to make mixed metal oxides of UO₃ and PuO₂ to prevent nuclear proliferation. The process comprises the steps of (1) dissolving spent nuclear fuel in an acidic solution in the presence of an agent that reduces Pu⁺⁶ to Pu⁺⁴ and an agent that oxidizes Pu⁺³ to Pu⁺⁴; (2) extracting U⁺⁶ and Pu⁺⁴ from acidic solution with an organic solvent comprising a ligand that binds U⁺⁶ and Pu⁺⁴ and which is soluble in the organic solvent; (3) jointly extracting U⁺⁶ and Pu⁺⁴ from the organic solvent with an acidic aqueous solution; and (4) precipitating a mixture of U⁺⁶ and Pu⁺⁴ by adding a carboxylic acid to the acidic aqueous solution. The U⁺⁶ and Pu⁺⁴ precipitate is then be calcined to form a mixed metal oxide of UO₃ and PuO₂. Additional steps can result in the formation of a mixed metal oxide of UO₃, PuO₂ and NpO₂ as well as the removal and isolation of technecium from the SNF.