143results about "Heterogenous reactors" patented technology

The present invention relates to a reactor construction comprising an upper compartment above a reactor compartment, said reactor compartment being separated from the upper compartment by a radiation shield, wherein a molten salt reactor (MSR) is situated in the reactor compartment, the MSR comprising a reactor vessel with a reactor vessel lid assembly comprising a reactor vessel lid shield and a reactor vessel lid wherein a lifting element is situated in the reactor construction, the lifting element comprising: at least one bar being substantially vertical and axially adjustable along its longitudinal axis by a drive mechanism situated in the upper compartment, and wherein at least one through-hole for the at least one bar is provided in the radiation shield and wherein the reactor vessel lid shield has at least one receiving structure configured to engage with and lock to the at least one bar. The invention also relates to a method of performing a maintenance operation in the reactor compartment and a power barge comprising at least one reactor construction.

The invention relates to a device and a method for controlling a pressure stabilizer spray. The device comprises: a valve regulating control loop and a valve opening-closing control loop, wherein the valve regulating control loop comprises a main control module and a signal conditioning module, the valve opening-closing control loop comprises a priority control module and a switch control module. The switch control module (21) and the priority control module (24) are respectively used for executing switch control and priority control: the signal conditioning module (13) is used for converting a position detection signal of a valve and then transmitting same to the main control module (12), and the main control module (12) is used for performing closed-loop adjustment control according to the position detection signal and a closed-loop adjustment instruction signal, By means of providing the valve switch control loop (20) and the valve adjustment control loop (10), full-opening control, full-closing control and closed-loop adjustment control can be performed on a electromagnetic drive valve, and stable control over the valve under various working conditions can be ensured, thereby ensuring the safe and stable operation of a nuclear power station.

An end fitting for a pressure tube in a nuclear reactor system includes an inboard end configured to surround and be joined to the end of the pressure tube in a fluid tight manner by a rolled joint. The rolled joint includes a grooved region having at least one annular groove. The end fitting also incudes a wettable inner surface that extends axially outboard from an outboard-most annular groove of the grooved region. The end fitting also incudes a corrosion barrier applied to form a coated portion of the wettable inner surface. The coated portion extends from a first position that is at or outboard of a coating reference position at the outside edge of the outboard-most annular groove, to a second position. An axial extent of the coated portion, measured axially from the coating reference position to the second position, is at least about 20 mm.

A fuel assembly (400) for a nuclear fission reactor system (100). The fuel assembly (400) may extend along a longitudinal axis (420). The fuel assembly (400) may comprise a pressure tube (402) and a single sleeve member (430) provided radially inward of a radially inner surface (432) of the pressure tube (402). The sleeve member (430) may define a chamber (434) which extends along a length of the sleeve member (430). The chamber (434) may comprise, in series, a first sub-chamber (436), a second sub-chamber (438) and a third sub-chamber (440). The first sub-chamber (436) may define a chamber flow inlet (442) and the third sub-chamber (440) defines chamber flow outlet (444). There may be provided a fuel compact unit (500) located in the second sub-chamber (438) within the sleeve member (430) such that a clearance gap (522) is maintained between a radially inner surface (450) of the sleeve member (430) and the fuel compact unit (500) to define a sleeve member flow passage (452).

Cogeneration installation with a nuclear reactor site and an industrial site with a chemical production unit, the industrial and nuclear sites being separated and connected by a closed-loop thermal energy storage system. The invention essentially consists of physically relocating and functionally separating a nuclear site from an industrial site comprising at least one chemical production unit that utilizes the heat produced by a reactor block at the nuclear site. This physical relocation is achieved by a thermal storage loop with hot and cold tanks located both at the nuclear site and at the industrial site. Figure for the abstract: Fig. 1

A buoyancy driven fluid system coupled to a reactor system configured to achieve free convection operable to cool the reactor system is disclosed. The buoyancy driven fluid system of the present disclosure generates natural circulation by designing the reactor system to have a large vertical offset between the heat exchanger and the reactor core thereby generating a large buoyancy force between a thermal center of the reactor core and a thermal center of the heat exchanger. By ensuring that the sum pressure drop of the components connected to the primary fluid loop is no greater than the buoyancy force of the system, the fluid may circulate throughout the reactor system without the aid of pumps or other forced flow mechanism. The reactor system may be designed within certain size constraints to maintain a compact form while still providing free convection.

A hybrid power generation system which includes a nuclear facility comprising a nuclear reactor and an exclusion zone. A thermal energy storage, a nuclear steam supply system located, and a solar energy collection system are all located within the exclusion zone. The thermal energy storage vessel contains a thermal mass composition operable to store thermal energy. The nuclear steam supply system has a nuclear reactor and a working fluid loop. The working fluid loop is configured to circulate a working fluid from a steam generator through the thermal energy storage vessel to absorb thermal energy and heat the working fluid for introduction to an electricity generating system. The solar energy collection system includes a heat transfer loop heated via a solar collector. The heat transfer loop is configured to circulate a heated heat transfer fluid to add thermal energy to the thermal mass composition in the thermal energy storage vessel.

A fuel assembly (100) for a nuclear fission reactor system (110). The fuel assembly (100) may comprise a pressure tube (120) comprising an aluminium or aluminium alloy. The fuel assembly (100) may comprise a sleeve member (132), the sleeve member (132) extending along a radially inner surface (134) of the pressure tube (120). The fuel assembly (100) may comprise a fuel compact unit (136). The fuel compact unit (136) may be located within the sleeve member (132)) such that a clearance is maintained between a radially inner surface (139) of the sleeve member (132) and the fuel compact unit (136) to define a sleeve member flow passage (140).

Regarding molten salts used in contact with graphite in nuclear reactors, there is a need for them to be easily maintained under strong reducing conditions without causing graphite degradation. [Solution] Use of a molten salt containing aluminum trifluoride and sodium fluoride as a primary coolant for a nuclear fission reactor, wherein the molten salt comes into contact with graphite and aluminum metals during the operation of the nuclear fission reactor.

One embodiment of the present invention provides a nuclear reactor structure comprising: a nuclear fuel storage part in which molten nuclear fuel is stored; a heat pipe filled with a working fluid having a temperature lower than the temperature of the molten nuclear fuel; a heat pipe jacket which forms a space in which the heat pipe can be disposed, and in which a heat transfer fluid is filled in the space; and a housing in which the nuclear fuel storage part and the heat pipe jacket are formed, wherein a nuclear fuel solidification layer is formed between the molten nuclear fuel and the heat pipe.

An apparatus and a method for separating a shielding ball are disclosed. An apparatus for separating a shielding ball comprises: an end shield which is installed in a spinning structure of a heavy water reactor facility and includes a fueling tubesheet and a calandria tubesheet facing each other and multiple shielding balls for filling a space between the calandria tubesheet and the fueling tubesheet; as a device for removing the shielding balls from a lattice tube connected to the calandria tube by being connected to the calandria tubesheet through the fueling tubesheet, a magnetic force applying unit inserted into the lattice tube in a direction of the fueling tubesheet to apply a magnetic force for fixing positions of the multiple shielding balls positioned within a first distance from the calandria tubesheet; a mover for moving the magnetic force applying unit in a direction of the lattice tube; and a storage rack for separately storing, inside the end shield, the shielding balls positioned within the first distance from the calandria tubesheet and shielding balls positioned at a distance greater than the first distance.

The present disclosure discloses a method and an apparatus for determining a rod position of a control rod of a pressurized water reactor, a rod position measurement system and a non-volatile computer-readable storage medium, and belongs to the technical field of nuclear power. In the method for determining a rod position of a control rod of a pressurized water reactor, a rod position determination condition of each rod position of a full stroke is determined according to a Gray code bit signal voltage of the rod position of the full stroke, and Gray code bits and corresponding Gray code bit signal voltage values are enabled to correspond to rod positions, so that according to a real-time signal and by means of a rod position interval and the rod position determination condition, each Gray code bit signal, which is currently obtained in real time, can be accurately positioned to each rod position point, so that rapid and accurate full-stroke rod position measurement is realized, accurate full-core real-time rod position information is provided for an operator to perform monitoring, and precise full-core and full-fuel-cycle rod position measurement data is provided for use in reactor core operation tracking, reactor core safety calculation and fine fuel management.

A nuclear reactor core (1) for a molten salt nuclear reactor (100). The nuclear reactor core (1) has a tubular cylindrical center moderator vessel (10) for passage of a liquid moderator (11), a cylindrical fuel salt jacket surrounding the center moderator vessel (10), and a cylindrical neutron reflector jacket surrounding the cylindrical fuel salt jacket, wherein the cylindrical center moderator and neutron reflector vessel (10) has a largest inner cross-sectional area medially between a liquid moderator and neutron reflector inlet (12) of the center moderator and neutron reflector vessel (10) and a liquid moderator and neutron reflector outlet (13) of the moderator and neutron reflector vessel (10).

Heat pipe fuel element includes an evaporation section, a condensing section, a capillary section connecting the evaporation section to the condensing section, and a primary coolant. In a cross-section in a plane perpendicular to a longitudinal axis of the evaporation section, the heat pipe fuel element includes a cladding layer enclosing an interior area including a fuel body formed of a fissionable fuel composition and that has an outer surface oriented toward the cladding layer and an inner surface defining a periphery of a vaporization space of the evaporation section. The fuel body has a structure with a shape corresponding to a mathematically-based periodic solid, such as a triply periodic minimal surface (TPMS), and the evaporation sections of a plurality of heat pipe fuel elements are arranged in a phyllotaxis pattern (as seen in a cross-section in a plane perpendicular to a longitudinal axis of the active core region).

Disclosed is an energy-independent smart port system using SMR-based surplus power. When an SMR-powered vessel docks at a port while the reactor is in operation, the energy-independent smart port system uses relevant technology and equipment to supply surplus power from the SMR-powered vessel to the port in consideration of safety regulations and the characteristic that restarting the reactor of the SMR-powered vessel requires significant time and cost, stably increase the energy independence of a smart port, and can reduce marine air pollutant emissions.

Fission reactions within a molten salt reactor may produce gaseous fission products. These fission products may be removed through a variety of methods. However, proper removal of such fission products requires a thorough understanding of the proportion of gaseous space to fluid space within the component being interrogated, known as the void fraction. The present invention provides a means for determining the void fraction within such a component utilizing an enrich cadmium zinc telluride detector to produce collimated radiation data. The collimated radiation data may then be used to compute a mean void fraction across the domain being interrogated.

A nuclear fission power plant (10) comprising: a gas circuit (90) to circulate gas; a nuclear reactor core (20) comprising: a fuel system (30) and a moderator (35); the nuclear fission power plant further comprising a neutron reflector (70), a plurality of control drums (50), one or more control rods (60), and a containment vessel (40) substantially surrounding the nuclear reactor core and allowing gas in the gas circuit to flow through both the containment vessel and the core, and an open Brayton power system coupled to the gas circuit via a heat exchanger (92), the open Brayton power system comprising: a compressor (102); a turbine system (104); and a generator (108); the open Brayton power system being arranged such that the turbine system is configured to drive the compressor and the generator.

In some embodiments, a power generation system may include a first nuclear reactor, which may include a vessel having a wall defining a cavity. The nuclear reactor may include a core disposed within said cavity. Said core may include a solid matrix material defining one or more channels. A fuel material may be disposed in said one or more channels. Said solid matrix material may be a primary heat transfer medium. The power generation system may include an electric generating system having a secondary heat transfer medium. The power generation system may include one or more heat exchangers thermally coupled to said vessel wall. Said primary heat transfer medium may transfer heat generated by the fuel material to the one or more heat exchangers. Said secondary heat transfer medium may draw heat from said one or more heat exchangers.

An integrated system for indirect cycle steam heating comprising a nuclear power module configured to output first steam; a turbine generator configured to receive the first steam and output second steam; a heat exchanger configured to receive water, receive at least one of first steam, second steam, and transfer heat from the at least one of first steam and second steam into the water to create third steam a peaking heater configured to receive the third steam, transfer augmenting heat into the third steam, and heat, based at least in part on the transfer, the third steam to have a temperature above a threshold temperature an auxiliary heater configured to receive the third steam; and a chemical processing plant configured to receive the third steam and transfer heat from the third steam into a chemical.

Systems and methods use inserted controllers in assembly openings of nuclear fuel castings. The controllers can be secured entirely within the opening and condition coolant flow to a desired level for the associated assembly. The controllers can be installed or removed at any time through clamps or other securing structures accessible through the top of the opening. This may include during fuel removal or shuffling during a maintenance outage, or at casting construction and installation, or at decommissioning. Side orifices do not have to be accessed from below the core nor do the castings otherwise require movement or alteration. Any desired flow characteristics may be selected and achieved through use of the configured controllers at specific core locations. Size, shape, and flow surfaces can all be configured to achieve these characteristics where they permit the controller to fit in the casting.

Disclosed is a sampling device for a calandria end fitting (10) of a heavy water reactor including: a sampling tool (100) inserted into an interior of an end fitting and configured to process the end fitting which is extended in an axial direction; a tool driver (200) configured to rotate the sampling tool; and a tool moving part (300) configured to move the sampling tool or the tool driver in which the sampling tool is installed in a perpendicular direction perpendicular to the axial direction.

An integrated PWR nuclear reactor that requires no refueling and is cooled by natural convection. It is housed within a reactor containment unit which is transportable to a placement site and placed within an underground building that allows heat transfer from the reactor to the surrounding soil. Tire reactor containment unit encloses an integrated PWR nuclear reactor, a pressure and chemical control system, a safety system and a turbo generator. After its operational life, the reactor's water and air are replaced with inert gases, circulating by natural convection to remove decay heat to the surrounding soil. This enables the integrated PWR reactor, initially designed to generate electricity, to act as a dry storage facility for uranium fuel elements after the end of its operational life.

The present invention relates to a reactor construction (1) comprising an upper compartment (2) above a drain tank compartment (3), said upper compartment (2) comprising: a molten salt reactor (MSR) (4), comprising a reactor vessel (5) comprising a molten fuel salt (6); a molten salt drain system connected to the reactor vessel (5); said drain tank compartment (3) comprising: one or more drain tanks (10) in communication with the molten salt drain system; a buffer water tank (15) comprising an inner wall (16) and an outer wall (17) and buffer water (21) in the gap between the inner wall (16) and the outer wall (17) of the buffer water tank (15), said buffer water tank (15) surrounding the one or more drain tanks (10), wherein a first piping structure (22) is defining a circuit for at least part of the buffer water (21) and comprising a heat exchanger (23) in thermal contact with a reservoir water tank (30), said reservoir water tank (30) being at a level above the buffer water tank (15), said reservoir water being in thermal contact with an environment and / or a second piping structure (24) is defining a circuit for at least part of the buffer water (21) and comprising a heat exchanger (25) in thermal contact with an environment, said heat exchanger (25) being at a level above the buffer water tank (15) and / or a seawater heat exchanger (26) in thermal contact with the buffer water (21) and seawater, said seawater heat exchanger (26) being located below seawater level outside the outer wall (17) of the buffer water tank (15).

The invention relates to an advanced nuclear fuel channel design for fast-spectrum reactors, featuring an annular O-ring configuration that supports natural circulation of molten fuel. The channel's variable cross-section is optimized to enhance fission in the core zone and neutron capture in the breeding zone. Radial distribution ensures a compact arrangement in the core and wider spacing in the breeding zone for improved neutron management and criticality control. The design integrates optional collectors for circulation support and is compatible with liquid metal or high-temperature gas coolants. This system offers a high-efficiency solution for reactors requiring extended operation and in-situ breeding capabilities.

A nuclear fission reactor system 100 provided with a nuclear reactor unit 200 comprises a moderator fluid tank 202 with a fluid flow inlet 210 and a fluid flow outlet 212. The temperature of moderator

The invention provides a reactor core block (200), comprising: a plurality of fuel channels (202); a plurality of heat pipes (204); a primary moderator matrix (206), consisting for example of graphite, silicon carbide, aluminum nitride, stainless steel, and combinations thereof, configured to encompass the plurality of fuel channels (202) and the plurality of heat pipes (204); and a secondary moderator channel (208) configured to at least partially surround the plurality of fuel channels (202), the plurality of heat pipes (204), and the primary moderator matrix (206). The secondary moderator channel (208) is comprised of metal hydride.

Thorium-based fuel bundles are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300 / 600 / 900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% 235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt%. Reprocessing of spent fuel bundles is also avoided.

A molten salt reactor system includes a fuel salt system configured to circulate a molten salt through a reactor vessel. The molten salt reactor system further includes an inert gas system fluidically coupled with the fuel salt system and configured to maintain a pressurized volume in a head space of a drain tank by circulating an inert gas along a first inert gas flow path. The molten salt reactor system further includes an equalization system configured to equalize pressure between head spaces of the reactor vessel and the drain tank in response to a reactor shutdown event. The inert gas system is configured to cease maintenance of the pressurized volume in response to the reactor shutdown event.