521results about "Heterogenous reactors" patented technology

A molten salt breeder reactor that has fuel conduit surrounded by a fertile blanket. The fuel salt conduit has an elongated core section that allows for the generation of electrical power on a scale comparable to commercially available nuclear reactors. The geometry of the fuel conduit is such that sub-critical conditions exist near the input and output of the fuel salt conduit and the fertile blanket surrounds the input and output of the fuel salt conduit, thereby minimizing neutron losses.

A nuclear reactor in which a secondary or tertiary coolant system of a nuclear steam supply system is simplified, comprising: a reactor vessel (2) which integrates a reactor core (1); a first coolant (3) which is stored in the reactor vessel (2) and heated by the reactor core (1) to convect; a first heat transfer tube (4) which is arranged in the reactor vessel (2) and comes into contact with the first coolant (3); and a second coolant (5) which is supplied from the outside of the reactor vessel (2) to the first heat transfer tube (4), cools the first coolant (3) and led to the outside of the reactor vessel (2).

The invention discloses an integral low-temperature nuclear heat reactor with the circuit equipment adopting integral arrangement and belonging to the low- and medium-parameter pressurized-water reactor, wherein, a reactor core adopts the mature nuclear power plant fuel component and control rod component; a main heat exchanger is of integral coil type; a voltage stabilizer is a built-in nitrogen partial pressure control voltage stabilizer; coolant circulation is completed by a built-in jet apparatus and the equipment of an external drive circuit; the drive circuit and the equipment and a main circuit auxiliary system are arranged at the circumference of a reactor pressure vessel; a containment vessel consists of a reactor body containment vessel and a reactor top containment vessel; the reactor body containment vessel which is a structure combined by a reactor vault of reinforced concrete structure and a casing of steel structure is connected with a sealed refueling water storage pool through a pipe and a valve. The thermal power of the reactor can be selected between 50MW and 500MW at will and the outlet temperature of the reactor can be selected between 100 DEG C to 200 DEG C according to application, requirement and power.

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 UClxFy 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.

An offshore, floating, moored, nuclear power generating plant and multi-purpose platform is disclosed herein. In a preferred embodiment, the invention consists of a spar or cell spar platform with multi-purpose, all weather topside decks, attached to a submerged “dry tank” that further includes: reactor generator deck(s), power plant main control deck, and central plant deck, that are all integrated within a watertight ballast hull. The invention design further includes cells that are modular for facilitating factory assembly and ultimate construction in a shipyard environment. Reactor vessels are typical naval nuclear reactor having a time tested outstanding safety record. A plurality of reactor generator modules are interconnected and operate independently and collectively and are transmitted by a plurality of high voltage direct current (HVDC) submarine cables through a transformer to the electrical grid. Multipurpose topside decks house vessel commend, 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 clamp apparatus is provided which is designed to structurally replace a weld that attaches a riser brace assembly to a reactor vessel wall, and to stiffen the entire riser brace assembly, thereby increasing the natural vibration frequency of the riser brace assembly. Installation of the clamp apparatus should not require removal of any installed reactor hardware, such as jet pump assembly hardware or shroud repair hardware. In an embodiment, the clamp apparatus may include a first plate arranged so as to mate with a first groove on the riser brace, and a second plate arranged so as to mate with a second groove on the riser brace. The grooves provide alignment for the clamp apparatus, so that the first and second plates can be fixedly secured to the riser brace.

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.

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".

The present invention is to provide a reactor core that allows a nuclear plant to continuously operate for a long term period, for example 15 years or longer, without requiring any fuel exchange, reduces the duration and number of maintenance steps involved in regular plant inspections, markedly improves plant availability and economic efficiency, and is effective in terms of nuclear nonproliferation.A plurality of fuel assemblies 103, themselves obtained by arranging fuel rods 100 and water rods 107 in square lattices, are arranged in a square lattice at a certain pitch. The blades 102a of a cross-shaped (cruciform) control rod 102 in a cross section are inserted into four adjacent spaces formed by four fuel assemblies 100 facing each other. A value of 0.06 cm−1 or greater is selected for the ratio (B / S) of the width (B) of each blade on the cruciform control rod 102 and the surface (S) of the fuel lattice defined by the surface area of a square whose side is equal to the pitch between the fuel assemblies 103.

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 compact pressurised water nuclear reactor comprises a primary circuit fully integrated into the reactor vessel (10). Thus, a single steam generator (12) forms the cover of the vessel (10) and the pressuriser (30) and the primary pumps (28) are housed in the vessel (10). The same is true for the control mechanisms of the control rods (40). Finally, a venturi system (44) is also provided in the vessel (10) to create water circulation if there is a failure of the primary pumps (28).

A reactor vessel includes a plenum and a reactor core with first and second sets of channels. A blanket salt flows through the first set of channels, and a fuel salt flows through the second set of channels. The plenum receives the blanket salt from the first set of channels. The blanket salt provides a breed-stock for a fission reaction in the fuel salt and transfers heat generated by the fission reaction without mixing with the fuel salt.

A telescoping guide for extraction and reinsertion support handling of in-core instrument thimble assemblies in the area above the upper support plate in the upper internals of a pressurized water reactor. The telescoping guides extend between the upper ends of the upper internals support columns and an axially movable instrumentation grid assembly which is operable to simultaneously raise the telescoping guides and extract the in-core instrument thimble assemblies from the reactor fuel assemblies.

The invention relates to a novel architecture for a nuclear reactor of the integrated type.The invention comprises:realising the hot area and cold area separation device for the primary sodium flow in the form of two walls with cuts,providing two pumping groups hydraulically in series, one for the flow of sodium from the hot area to the cold area through the intermediate exchangers and the other in the cold area;providing outlet windows of the intermediate exchangers below the lower wall;providing outlet windows of the removal exchangers of the decay heat above the cold area, wherein all of clearances between the walls with cuts and the heat removal exchangers and the height between the two walls with cuts are previously determined so as to, during normal operation, take up differential movements between the walls, exchangers and vessel and to make it possible to establish during normal operation a thermal stratification of the primary sodium in the space defined between the horizontal portions of the two walls and so as to reduce, in case of an unexpected stop of a single pumping group, the mechanical stress applied to the walls and due to the portion of the primary sodium flow passing between said clearances.

The invention belongs to the technical field of nuclear reactors and discloses a heating pipe-type dual-mode space nuclear reactor core. The reactor core comprises a reactor core active zone, a reactor core cylinder, a radial reflection layer, an axial reflection layer and a control drum. The reactor core active zone is located in the reactor core cylinder. The axial reflection layer is located above the reactor core active zone. The axial reflection layer is a hollow cylindrical structure. The reactor core active zone, the axial reflection layer and a heating pipe are located in the cavity ofthe radial reflection layer. The reactor core can effectively prevent the heat pipe overheating and has the beneficial effects of safety and reliability such as passive and non-single point failure.

A process for producing technetium-99m from a molybdenum-100 metal powder, comprising the steps of:(i) irradiating in a substantially oxygen-free environment, a hardened sintered target plate coated with a Mo-100 metal, with protons produced by a cyclotron;(ii) dissolving molybdenum ions and technetium ions from the irradiated target plate with an H2O2 solution to form an oxide solution;(iv) raising the pH of the oxide solution to about 14;(v) flowing the pH-adjusted oxide solution through a resin column to immobilize K[TcO4] ions thereon and to elute K2[MoO4] ions therefrom;(vi) eluting the bound K[TcO4] ions from the resin column;(vii) flowing the eluted K[TcO4] ions through an alumina column to immobilize K[TcO4] ions thereon;(viii) washing the immobilized K[TcO4] ions with water;(ix) eluting the immobilized K[TcO4] ions with a saline solution; and(x) recovering the eluted Na[TcO4] ions.

A molten salt breeder reactor that has fuel conduit surrounded by a fertile blanket. The fuel salt conduit has an elongated core section that allows for the generation of electrical power on a scale comparable to commercially available nuclear reactors. The geometry of the fuel conduit is such that sub-critical conditions exist near the input and output of the fuel salt conduit and the fertile blanket surrounds the input and output of the fuel salt conduit, thereby minimizing neutron losses.

A portable nuclear fuel cartridge comprising a unitary support structure and a plurality of nuclear fuel assemblies that collectively form a nuclear fuel core. The nuclear fuel core is integrated into the unitary support structure to collectively form a self-supporting assemblage than can be lifted as a single unit. In another aspect, the invention is a method of fueling and / or defueling a nuclear reactor utilizing a nuclear fuel cartridge that is loaded and / or unloaded from the nuclear reactor as a single unit. In another aspect, a nuclear reactor core is provided that comprises a nuclear fuel core comprising; a plurality of first nuclear fuel assemblies, each of the plurality of first nuclear fuel assemblies having a first transverse cross-sectional configuration; and a plurality of second nuclear fuel assemblies, each of the plurality of second nuclear fuel assemblies having a second transverse cross-sectional configuration that is different than the first transverse cross-sectional configuration.

The invention provides a module type pressurized water reactor. A reactor module is composed of a reactor body, a cooling agent system, and a safety system, the reactor body, the cooling agent system, and the safety system are all enclosed in a steel safety shell, and the whole reactor module is immersed in a reactor safety water pool, taking the pool water as the final hot trap for heat discharging of accidents. Emergency core cooling and emergency excess heat discharging are achieved by adopting a complete passive way, and the pressure inhibition and long-term heat discharging of the safety shell are achieved through repressing the water in the water pool and the reactor safety water pool; and during the processes of reloading and maintenance of reactor, through arranging a reloading water jacket between the safety shell reloading channel flange and the reactor container flange, direct-moving underwater fuel lifting operation is achieved between the safety shell and the spent fuel pool. A reactor factory can be provided with a plurality of reactor modules, all reactor modules co-use material loading equipment and other nuclear auxiliary systems, and the safety and economy of the reactor are improved through modularized combination arrangement. The module type pressurized water reactor is suitable for being applied to the fields of nuclear electricity generation, gas application, heat application, sea water desalination, and the like.

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 UClxFy 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 invention relates to a novel architecture for a nuclear reactor of the integrated type. The invention comprises: realising the hot area and cold area separation device for the primary sodium flow in the form of two walls with cuts, providing two pumping groups hydraulically in series, one for the flow of sodium from the hot area to the cold area through the intermediate exchangers and the other in the cold area; providing outlet windows of the intermediate exchangers below the lower wall; providing outlet windows of the removal exchangers of the decay heat above the cold area, wherein all of clearances between the walls with cuts and the heat removal exchangers and the height between the two walls with cuts are previously determined so as to, during normal operation, take up differential movements between the walls, exchangers and vessel and to make it possible to establish during normal operation a thermal stratification of the primary sodium in the space defined between the horizontal portions of the two walls and so as to reduce, in case of an unexpected stop of a single pumping group, the mechanical stress applied to the walls and due to the portion of the primary sodium flow passing between said clearances.

A reactor vessel includes a plenum and a reactor core with first and second sets of channels. A blanket salt flows through the first set of channels, and a fuel salt flows through the second set of channels. The plenum receives the blanket salt from the first set of channels. The blanket salt provides a breed-stock for a fission reaction in the fuel salt and transfers heat generated by the fission reaction without mixing with the fuel salt.

This disclosure describes various configurations and components of a molten fuel fast or thermal nuclear reactor in which one or more primary heat exchangers are located above the reactor core of the nuclear reactor.

A steam separator comprises an outer main swirler and an inner auxiliary swirler which is smaller than the main swirler. The swirlers are provided so as to be concentric on the inner wall at the lower side of the first stage inner cylinder. In the steam separator, when the gas-liquid two-phase flow which flows in the vicinity of the axial center of the first stage inner cylinder passes the auxiliary swirler, it is separated into steam and water by the centrifugal force. The separated water (droplets) is introduced into the main swirler. When the separated water (droplets) passes the main swirler, it is separated at the inner wall side of the first stage inner cylinder by the centrifugal force.Pressure loss in a steam separator is reduced and steam separation capability is increased without increasing the moisture from the steam separator.

The invention discloses a method and a device for fast multiplicating and transforming nuclear fuel with high inherent security and low cost, and the method and the device adopt 'moving solid nuclear fuel' which can flow and is similar to fluid, can be applied in a fast neutron reactor, and produce more nuclear fuel with less fissile fuel initial installation quantity and lower cost in a shorter time, thus satisfying the nuclear fuel demand at the current world nuclear power big development stage. The method and the device for fast multiplicating and transforming nuclear fuel are real nuclear reactors with inherent security, and can thoroughly eliminate the serious accident of massive radioactive release caused by element melting of the nuclear fuel which still possibly occurs in the current nuclear power station.

The present invention provides an inherent security pool-shell combination low temperature nuclear heating device and an operation program thereof, which belongs to nuclear energy centralized heat-supply field. The invention mainly comprises the following procedures: adopting separated layout, forced circulation and high effective small temperature difference heat exchange in primary loop of reactor; under the precondition that parameter requirements of centralized heat-supply by big cities heat supply network satisfied, reducing work temperature of shell-type heating reactor as far as possible; meanwhile optimized layout of equipment, and organic combination of normal pressure pool in large volume (47) and shell-style heating reactor(29) remove accident possibility of water loss and coldness loss of reactor core (1). Complete inherent safety principle is carried out in the invention. Thus, absolute security of nuclear heating can be guaranteed by using mature pressurized water reactor technology, and enterprises have develop nuclear station pressurized water reactor experience can possibly jump over research and develop stage and enter into nuclear heating market directly, and make a real contribution for energy-saving and emission-reduction and improving country energy structure.

A reactivity control rod adapted to be used in a reactor core of a fast reactor and disposed at a substantially central portion of the reactor core for controlling a reactivity therein. The reactivity control rod includes a wrapper tube surrounded by a plurality of fuel rods in a reactor core, and a plurality of neutron absorber rods arranged in the wrapper tube. At least one of the plurality of neutron absorber rods includes a cladding tube and a mixture filled in the cladding tube. The mixture is composed of a neutron absorber that absorbs a neutron and a neutron moderator that moderates the neutron.

An integral pressurized light water reactor having most of the components of a primary side of a pressurized water reactor nuclear steam supply system housed in a single pressure vessel with a pressurizer separated from the remaining reactor system by a surge separator having multiple layers of separated steel plates with a number of concentric baffles extending therebetween. A circuitous flow path is provided through and between the plates and concentric baffles and a relatively stagnant pool of coolant is maintained within an innermost zone between the plates to provide thermal isolation.