182results 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.

The present invention is related to a device and a method for producing a radioisotope of interest from a target fluid irradiated with a beam of accelerated charged particles, the device includes in a circulation circuit (17): an irradiation cell (1) having a metallic insert (2) able to form a cavity (8) designed to house the target fluid and closed by an irradiation window (7), the cavity (8) including at least one inlet (4) and at least one outlet (5); a pump (16) for circulating the target fluid inside the circulation circuit (17); an external heat exchanger (15); the pump (16) and the external heat exchanger (15) forming external cooling means of the target fluid; the device means for pressurizing (14) of the circulation circuit (17) and the external cooling means of the target fluid are arranged in such a way that the target fluid remains inside the cavity (8) essentially in the liquid state during the irradiation.

A system and method for employing the in-core movable detectors of a commercial nuclear powered electric generating facility to transmute a user-specified target material into a desired isotope. The process is conducted remotely resulting in a shielded end product available for shipment for further 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.

A mixed waste mediated electrochemical oxidation process (MEO) process and apparatus for the dissolution of transuranic elements, and / or compounds thereof in transuranic waste, low level waste (LLW), low level mixed waste, special case waste, and greater than class C LLWS, and also the destruction of the non-fluorocarbon organic component in the waste. The MEO process and apparatus operates in three different modes: dissolution, destruction, and decontamination. In the first mode, dissolution, the process runs until the transuranics such as a mixed oxide or carbide, and / or mixture of oxides or carbides of uranium and plutonium are totally dissolved into solution. The second mode, destruction, the process is operated such that the mixed waste materials are reduced to CO2, water and small amounts of inorganic salts. The third mode, decontamination, involves contaminated equipment. In the decontamination mode the MEO process destroys the mixed wastes that have contaminated the equipment.

A mixed waste mediated electrochemical oxidation process (MEO) process and apparatus for the dissolution of transuranic elements, and / or compounds thereof in transuranic waste, low level waste (LLW), low level mixed waste, special case waste, and greater than class C LLWS, and also the destruction of the non-fluorocarbon organic component in the waste. The MEO process and apparatus operates in three different modes: dissolution, destruction, and decontamination. In the first mode, dissolution, the process runs until the transuranics such as a mixed oxide or carbide, and / or mixture of oxides or carbides of uranium and plutonium are totally dissolved into solution. The second mode, destruction, the process is operated such that the mixed waste materials are reduced to CO2, water and small amounts of inorganic salts. The third mode, decontamination, involves contaminated equipment. In the decontamination mode the MEO process destroys the mixed wastes that have contaminated the equipment.

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.

A laminar component (1) for a fusion reactor, which can accommodate high thermal stresses, comprises tungsten or a tungsten alloy with a tungsten content greater than 90 wt% in the plasma region (2), and copper or a copper alloy in a heat removal region (4). A metallic component (5) is against the copper region. Preferred Features: The latter has a conductivity greater than 250W / mK and a mean granular size in excess of 100 micron. A refractory metal-copper laminate is between them in an intermediate region (3). The laminate is 0.1-4 mm thick, and the refractory metal content is 10-40 Vol%. The refractory metal phase forms a skeleton.

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 and system are provided for propelling an aerodynamic vehicle into space. The aerodynamic vehicle uses a nuclear-based thermal rocket (NTR) propulsion system capable of producing a hydrogen exhaust. A flow of air is introduced into the hydrogen exhaust to augment the thrust force at speeds of the vehicle up to approximately Mach 6. When the speed of the vehicle is approximately Mach 6 and the altitude of the vehicle is approximately 40 kilometers, the flow of air is stopped and the vehicle is propelled into space using only the NTR.

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.

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.

Device to irradiate a sample in the core or at the periphery of the core of a nuclear reactor comprising, a double wall containment (6) delimiting a chamber (7), a receptacle (4) contained in said chamber (7), said receptacle being held at a distance from an inside wall (8) of the containment, said receptacle (4) being designed to contain a coolant, a sample holder (2), a free end of which will be located in the receptacle, in which the inside of the receptacle is in fluid communication with the outside of the receptacle and in which a volume between the inside wall (8) of the containment (6) and the receptacle (4) will be filled with a gas or a mix of gases called the coolant cushion gas.

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

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.

The invention belongs to the technical field of reactor material irradiation performances and discloses a metallic material irradiation device used in a reactor. The device comprises a packaging casing, a sample casing, a core shaft, a base and a pressing cover, wherein the packaging casing is positioned above the sample casing and is connected with the sample casing by a clamping groove; the sample casing comprises one or more sub-casings; the inner side of each sub-casing is provided with one or more isolating devices which can be used for placing standard samples. The device disclosed by the invention has the characteristics that the temperature distribution is uniform and a plurality of samples can be placed.

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.

The present invention relates to a treatment of high-level waste of radiochemical production containing radionuclides and macro-admixtures including sodium. The method of extraction of radionuclides by processing acidic aqueous waste solutions by extractants containing macrocyclic compounds selected from the group of crown ethers having aromatic fragments containing alkyl and / or hydroxyalkyl substituents of a linear and / or branched structure, and / or cyclohexane fragments containing alkyl and / or hydroxyalkyl substituents of a linear and / or branched structure, and / or fragments of —O—CHR—CH2O—, where R is the normal or branched alkyl or hydroxyalkyl in organic solvents containing polyfluorinated telomeric alcohol 1,1,7-trihydrododecafluoroheptanol-1 having the formula H(CF2CF2)nCH2OH, where n=3, and a mixture of polyoxyethylene glycol ethers of synthetic primary higher aliphatic alcohols of a fraction C12-C14 of a general formula CnH2n+1O(C2H4O)mH, where n=12-14, m=2 is proposed. A possibility of selection of quantitative and qualitative composition of the extractant depending on the waste composition and the object to be achieved of joint or individual extraction of radionuclides results in universal technology. A method of regeneration of extractants containing crown ethers allowing one to carry out the operation of extraction of radionuclides in a commercial scale under countercurrent dynamic conditions, with regeneration of the extractant for reuse is also proposed.

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.

The invention belongs to the technical field of treatment of nuclear materials and relates to a method for recovering plutonium from a radioactive spent fuel. The method sequentially comprises the following steps: performing extraction: extracting tetravalent plutonium in an aqueous solution of the radioactive spent fuel with an organic solvent containing tributyl phosphate; performing reduction on plutonium with a reducing agent: adding an organic solvent containing dimethylhydroxylamine in an organic phase to reduce the tetravalent plutonium into trivalent plutonium; and performing back extraction: adding a dilute acid aqueous solution in the organic phase to perform back extraction on the plutonium into a water phase, and performing recovering. Through adoption of the method for recovering the plutonium from the radioactive spent fuel, disclosed by the invention, the efficiency of recovering the plutonium from the radioactive spent fuel can be obviously improved compared with HAN back extraction, and the problem caused by reduction and extraction with U (IV) can be solved at the same time.

Embodiments of a method and a system for recovering a metal, such as uranium, from a metal-containing material are disclosed. The metal-containing material is exposed to an extractant containing a liquid or supercritical-fluid solvent and an acid-base complex including an oxidizing agent and a complexing agent. Batches of the metal-containing material are moved through a series of stations while the extractant is moved through the stations in the opposite direction. After the extraction step, the metal is separated from the solvent, the complexing agent and / or other metals by exposing the extract to a stripping agent in a countercurrent stripping column. The complexing agent and the solvent exit the column and are separated from each other by reducing the pressure. The recovered complexing agent is recharged with fresh oxidizing agent and recombined with fresh or recovered solvent to form a recovered extractant, which is distributed through the extraction stations.

A nuclear power plant thermal efficiency diagnostic system 10 comprises: a feed-and-condensate-water-flow-rate-setting-means 13 for setting a flow rate of feedwater tentatively; a heat-exchange-on-heater-calculating-means 14 for calculating heat exchange quantities of the feedwater and the condensate water on a heater; a HP turbine power calculating means 15 for acquiring a calculated power value of a high pressure turbine; a HP turbine power correcting means 16 for making the calculated power value of the high pressure turbine corrected; a HP turbine internal efficiency calculating means 17 for calculating an internal efficiency of the high pressure turbine; a steam condition on LP turbine inlet calculating means 18 for setting a condition of a steam on an inlet of a low pressure turbine; a LP turbine power calculating means 21 for acquiring a calculated power value of the low pressure turbine; a LP turbine power correcting means 22 for making the calculated power value of the low pressure turbine corrected; a LP turbine internal efficiency calculating means 23 for calculating an internal efficiency of the low pressure turbine; and a performance-deteriorating-element-specifying-means 25 for specifying an element which causes on deterioration of performance of the nuclear power plant.