667results about "Reactor fuel susbtances" patented technology

A set of alloy formulations is disclosed to use with thorium based nuclear fuels in a fast spectrum reactor; with thorium based nuclear fuels in existing thermal spectrum power reactors; for medical isotope production in the epithermal, the fast, the fission spectrum and the thermal spectra; and to use as fuel in test and experimental reactors that are non proliferative. The alloys form inert metal matrixes to hold fine particles of dispersed thorium containing fuel. The formulations also are useful for the production of medical and commercial isotopes in the high energy, fast and epithermal neutron spectra.

Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

A nuclear fuel element includes a core formed from a high density solid solution fissile material that is substantially free of carbon and void space. A cladding substantially surrounds the core.

This disclosure describes nuclear fuel salts usable in certain molten salt reactor designs and related systems and methods. Binary, ternary and quaternary chloride fuel salts of uranium, as well as other fissionable elements, are described. In addition, fuel salts of UCl_xF_y are disclosed as well as bromide fuel salts. This disclosure also presents methods and systems for manufacturing such fuel salts, for creating salts that reduce corrosion of the reactor components and for creating fuel salts that are not suitable for weapons applications.

The present invention relates to a fast reactor type coupling nuclear reaction implementation method and a nuclear reactor for same. The main contents comprise: a fast reactor type coupling nuclear reaction implementation method, a reactor modular design approach, a fast reactor type coupling nuclear reactor, a reactor core, a fuel element, a nuclear control system, and a proliferation fuel system. The fast reactor type coupling nuclear reactor mainly combusts thorium and nuclear waste, and has inherent security. The reactor main container is composed of a fission pool and a moderating pool that are completely isolated from each other but coupling to each other. A primary coolant is separated from a moderator. A thermal insulation layer is disposed between the fission pool and the moderating pool so that both can perform neutron exchange but heat exchange is blocked. Fast neutrons produced by the fission pool and moderated neutrons reflected by the moderating pool may enable the reactor core to simultaneously perform coupling nuclear reaction of the two types of neutrons. The moderating pool may be provided with the nuclear control system, and ex-core coupling core control may be implemented. The moderating pool is provided with a thorium purification fuel system, and on-line extraction of the purification fuel can be performed, and separation of nuclide is safe and simple, thereby providing a solution to the technical bottleneck of "thorium reactor".

A molten salt reactor is described that includes a containment vessel, a reactor core housed within the containment vessel, a neutron reflector spaced from the containment vessel and positioned between the core and the containment vessel, a liquid fuel comprised of a nuclear fission material dissolved in a molten salt enclosed within the core, a plurality of heat transfer pipes, each pipe having a first and a second end, wherein the first end is positioned within the reactor core for absorbing heat from the fuel, a heat exchanger external to the containment vessel for receiving the second end of each heat transfer pipe for transferring heat from the core to the heat exchanger, and at least one and preferably two or more reactor shut down systems, where at least one may be a passive system and at least one or both may be an active or a manually operated system. The liquid fuel in the core is kept within the core and heat pipes are used to carry only the heat from the liquid core to the heat exchanger.

A novel containment system for encapsulating nuclear fuel particles is disclosed. The containment system has a gas-impervious ceramic composite hollow shell having a spheroidal or ovoidal shape. The shell has a pair of longitudinally aligned round openings that are sealed with a gas-impervious ceramic composite tube to define a cavity between the shell inner surface and the tube outer surface. A ceramic composite matrix containing the nuclear fuel particles is enclosed within the cavity. The ceramic composite matrix has a controlled porosity, and can contain moderators or neutron absorbing material. The tube and shell are composed of a ceramic matrix composite material composed of ceramic reinforcement material that is bound together by a polymer-derived ceramic material.

The invention discloses a method for preparing metal reinforced uranium dioxide nuclear fuel pellets. The method mainly comprises two steps: firstly, preparing core-shell structure granules, namely performing low-temperature rapid pre-sintering on UO2 powder by using a Spark Plasma Sintering SPS technique, pelletizing, balling to obtain UO2 pellets, performing physical mixing on the UO2 pellets with metal (one of Mo, Cr, W and the like) micro powder to coat surfaces of the UO2 pellets by the metal micro powder, thereby obtaining metal coated uranium dioxide core-shell structure granules; secondly, preparing a nuclear fuel pellets, namely performing high-temperature liquidation on the metal powder on the surfaces of the UO2 pellets, thereby forming a micro cell structure continuous phase similar to a cytomembrane structure around the UO2 pellets, and obtaining the special metal reinforced UO2 nuclear fuel pellets with a UO2 substrate.

A nuclear-powered plant for systems of up to about 100 MWs with a confinement section where the reaction takes place in a core having a reactive thorium/uranium-233 composition, and where an external neutron source is used as a modulated neutron multiplier for the reactor core output. The core is housed in a containment structure that radiates thermal energy captured in a multiple-paths heat exchanger. The exchanger heat energy output is put to use in a conventional gas-to-water heat exchanger to produce commercial quality steam.

A composite nuclear fuel pellet comprises a composite body including a UO2 matrix and a plurality of high aspect ratio particies dispersed therein, where the high aspect ratio particies have a thermal conductivity higher than that of the UO2 matrix. A method of making a composite nuclear fuel pellet includes combining UO2 powder with a predetermined amount of high aspect ratio particles to form a combined powder, the high aspect ratio particles having a thermal conductivity higher than that of the UO2 powder; mixing the combined powder in a solvent to disperse the high aspect ratio particles in the UO2 powder; evaporating the solvent to form a dry mixture comprising the high aspect ratio particles dispersed in the UO2 powder; pressing the dry mixture to form a green body; and sintering the green body to form the composite fuel pellet.

The invention relates to a preparation method of uranium nitride fuel powder and a pellet, which comprises the following steps: (1) carrying out surface cleaning treatment on a pure metal uranium lump; (2) hydrogenating the pure metal uranium lump in hydrogen at 150-300 DEG C for 2-16 hours, dehydrogenating in a vacuum, and repeating the hydrogenation-dehydrogenation several times to obtain metal uranium powder; (3) nitridizing the metal uranium powder in nitrogen at 200-600 DEG C for 6-24 hours to obtain U2N3 powder; (4) denitrifying the U2N3 powder in a mold, keeping the system vacuum, pressurizing the mold to 30-60 MPa, sintering at 1450-1620 DEG C for 1-2 hours, and cooling to obtain the UN ceramic pellet. The method provided by the invention is simple in technique and easy to control; and the obtained intermediate product U2N3 powder has higher purity and higher sintering activity, and can be finally subjected to hot pressed sintering to obtain the UN fuel pellet of which the TD is up to 98.9%.

A steam generator system for a pressurized water reactor which employs an external to containment steam drum and recirculation loop piping. The steam generator system changes the arrangement of a typical pressurized water reactor recirculation steam generator by relocating the functions of steam separation and feedwater preheating outside of the reactor coolant system. The steam generator system and thermal hydraulic conditions are selected in order to minimize the size of the steam generator heat exchanger component volume inside of the containment. The external steam drum component can be isolated in accident conditions when desired and is used as a source of secondary fluid inventory for improved decay heat removal capability and tolerance for loss of feedwater events. Thus, the steam generator component volume inside of the containment is reduced and the amount of maintenance required for the reactor coolant system components are similarly reduced.

A sheathed, annular metal fuel system is described. A metal fuel pin system is described that includes an annular metal nuclear fuel alloy. A sheath may surround the metal nuclear fuel alloy, and a cladding may surround the sheath. A gas plenum may also be present. Mold arrangements and methods of fabrication of the sheathed, annular metal fuel are also described.

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

The invention discloses a preparation method of a porous-fuel-core inert matrix dispersion fuel pellet. The preparation method mainly comprises the steps of preparation of porous fuel core beads, preparation of inert matrix slurry, preparation of composite particles, pressing of a core, pressing of a core shell and preparation forming. According to the method, TRISO particles in an existing pellet is replaced with a porous fuel core, so that uraniumu abundance in the pellet is improved by four times or more, neutron economy of the fuel pellet can also be obviously improved, and power station operation cost and spent fuel reprocessing cost are reduced; moreover, a process period can also be greatly shortened, batch preparation of the pellet is implemented, and production cost is obviously reduced; the prepared inert matrix dispersion fuel pellet is higher in compactness, smaller in crystal grain size, fewer in structural defects and more excellent in high-temperature stability and irradiation resistance.

A clamp assembly connects a diffuser adapter or lower ring to a diffuser tail pipe of a jet pump diffuser in a boiling water nuclear reactor. The clamp assembly includes at least two clamp segments shaped generally corresponding to an exterior circumference of the diffuser, a swivel link affixed at each end of each of the clamp segments, and at least two connecting bands pivotably secured to the swivel links between the ends of the clamp segments. The clamp segments each includes a locking assembly engageable with the diffuser adapter or lower ring and the diffuser tail pipe. The clamp assembly structurally replaces/repairs the weld joining the diffuser adapter or lower ring and the diffuser tail pipe.

A method of manufacturing nuclear fuel elements comprising the steps of placing nuclear fuel balls in the container made from ultra-porous material, applying a CVI to the container and removing the container. The container for manufacturing fuel elements comprising balls, and is produced from at least one ultra-porous material, for example carbon foam.

The invention discloses an integrated small molten salt reactor. The reactor body of the integrated small molten salt reactor comprises a fuel salt, heat tubes, a core container, a side reflection layer, neutron absorbers, control drums, a shell and a lower reflection layer, wherein the heat tubes and the core container are integrally designed, the heat tubes are partly or fully inserted into thecore container, the core container is filled with the fuel salt, a cavity is formed between the core container and the liquid surface of the fuel salt, the outer wall of the core container is surrounded with the side reflection layer, the control drums are arranged in the side reflection layer, the neutron absorbers are arranged at one sides of the control drums, the lower reflection layer is arranged at the bottom of the core container, and the shell surrounds the external of the side reflection layer and the lower reflection layer. A heat tube technology is applied to the molten salt reactorin order to greatly simplify the structure of the above system, and natural circulation or natural convection is generated to realize the long-term efficient and stable transmission of heat and improve the safety and the reliability of the system.

The invention relates to a process for preparing a castable powder of uranium dioxide UO2, for use in the manufacture of MOX fuel.This process comprises the following stages:1) to prepare an aqueous suspension of a powder of UO2 obtained by dry process from uranium hexafluoride, said suspension comprising 50 to 80% by weight of UO2 and at least one additive chosen among deflocculation agents, organic binders, hydrogen peroxide H2O2 and a powder of U3O8, in such a quantity that the viscosity of the suspension does not exceed 250 mPa.sec, and2) to atomise this suspension and dry it in a hot gas, at a temperature of 150 to 300° C., to obtain a castable powder of UO2 with an average particle size of 20 to 100 mum.