238results 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.

The present disclosure relates to a method for monitoring failure of coated particles in fuel elements in a core of a pebble-bed high-temperature gas-cooled reactor, which is related to the technical field of nuclear reactor engineering and includes the following steps: S11, calculating an inventory of a short-lived noble gas fission nuclide; S12, obtaining a ratio of a release rate to a birth rate of the short-lived noble gas fission nuclide based on a temperature of the fuel elements using a Booth diffusion and release model; S13, deriving a theoretical expression for an activity concentration of the short-lived noble gas fission nuclide in a primary circuit using a migration model of the nuclide in the primary circuit; S14, obtaining an experimental measurement value of the activity concentration of the short-lived noble gas fission nuclide in the primary circuit at a sampling moment by gas sampling; S15, optimally calculating a failure fraction of the coated particles in the fuel elements and a share of uranium contamination in the matrix graphite in the core based on the theoretical expression and the experimental measurement value. The present disclosure can provide key parameters for the performance and status of the fuel elements in the core, which are required for radiation safety studies, source term calculations and accident analysis of the pebble-bed high-temperature gas-cooled reactor.

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 nuclear reactor is designed to couple the load path of the control elements with the reactor core, thus reducing the opportunity for differential movement between the control elements and the reactor core. A cartridge core barrel can be fabricated in a manufacturing facility to include the reactor core, control element supports, and control element drive system. The cartridge core barrel can be mounted to a reactor vessel head. Thus, any movement, such as through seismic forces, transmits an equal direction and magnitude to the control elements and the reactor core. This arrangement reduces the opportunity for differential movement between the control elements and the reactor core.

The application discloses a kind of efficient supercritical carbon dioxide high temperature gas cooled reactor systems and its operating method, the system includes pre-cooler, main compressor, re-compressor, high temperature turbine, low temperature regenerator, high temperature regenerator, high temperature helium-supercritical carbon dioxide heat exchanger, low temperature helium-supercritical carbon dioxide heat exchanger and reactor.The application proposes the power generation system of two-stage split supercritical carbon dioxide cycle comprising high-temperature split re-compression cycle loop and low-temperature re-split reheat expansion cycle loop, by arranging high-temperature helium-supercritical carbon dioxide heat exchanger and low-temperature helium-supercritical carbon dioxide heat exchanger in power generation system, ensure the good matching of helium loop heat release temperature window and supercritical carbon dioxide power cycle loop heat absorption temperature window, realize the efficient cascade utilization of high-temperature gas cooled reactor helium loop large-span temperature interval heat.

In many embodiments, a plurality of neutron moderators may be arranged to define a plurality of flow channels within a reactor vessel, wherein each neutron moderator of the plurality of neutron moderators includes at least one surface face, and each neutron moderator is not machined. In some embodiments, the flow channel may be defined by the at least one surface face of each neutron moderator of the plurality of neutron moderators. In at least one embodiment, the plurality of neutron moderators may include at least three neutron moderators. In several embodiments, the plurality of neutron moderators may include a first pair of neutron moderators and a second pair of neutron moderators. In one or more embodiments, the first pair of neutron moderators and the second pair of neutron moderators may be arranged to form a lattice configuration in a reactor vessel.

A nuclear installation comprising at least one modular nuclear reactor (SMR) and a reactor vessel well delimiting a water basin in which the SMR reactor unit is immersed. The invention relates to a nuclear installation comprising: at least one SMR reactor unit (4) for heat generation, delimited by a reactor vessel (40, 41, 42); a reactor vessel well (100) delimiting a water-filled basin (B) forming part of a secondary circuit, in which the reactor vessel is immersed; at least one heat exchanger (49) between the primary circuit of the SMR reactor and the secondary circuit, arranged outside the reactor vessel and in the water basin. Figure for the abstract: Fig. 6

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).

A control rod drive mechanism having a torque tube, a control rod assembly including a connecting rod disposed within the torque tube and including an annular collar defining a key slot, an elongated key that is slidably receivable within the key slot, a lock cam assembly rotatably secured to a top end of the connecting rod and including a locking cam, and a holdout collar disposed non-rotatably within the torque tube, the holdout collar including a locking recess with an entry slot, wherein in a first axial position the lock cam assembly is rotatable with respect to the torque tube, and a second axial position the lock cam assembly is non-rotatable with respect to the torque tube.

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 nuclear fission power plant (10) configured for extra-terrestrial use comprising a gas-cooled moderated nuclear reactor core (20) with a metallic fuel (30), a moderator (60), and a containment vessel (40) with a containment vessel inlet (42) and a containment vessel outlet (44). A neutron reflector (70) is disposed around a periphery of the gas-cooled moderated nuclear reactor core, and a plurality of neutronic control elements (50) comprising rotatable control drums disposed around the periphery of the gas-cooled moderated nuclear reactor core. A fluid circuit (90) is configured to receive hot gas from the containment vessel outlet and deliver cooled gas to the containment vessel inlet. An electricity generating system (91) is coupled to the fluid circuit so as to generate an electrical current.

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 high temperature control rod (22) for a nuclear fuel assembly (10) is described herein that includes a neutron absorbing material (24) having a melting point greater than 1500 °C that does not form a eutectic with a melting point less than 1500 °C, and may further include a cladding material (30) having a melting point greater than 1500 °C. The cladding material (30) is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material (24) is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding (30) or the absorbing material (24) may be coated with an anti-oxidation coating (26) of Cr with or without a Nb intermediate layer (28).

A fuel fragment screening device of a nuclear fuel loading and unloading system in a gas-cooled high-temperature reactor, comprising a box body, a spiral screen channel, a fuel ball inlet, a fuel ball outlet, and a waste outlet; the box body is arranged on the inlet line of the baffle, with the fuel ball inlet provided on the top of the box body and the fuel ball outlet and the waste outlet provided on the bottom; inside the box body, between the fuel ball inlet and the fuel ball outlet, the spiral screen channel is provided, and the spiral screen channel has a certain angle and is in an inclined state, with the waste outlet being located below the spiral screen channel.During the operation of the present application, when the fuel balls, fuel ball fragment, and the dust mixture pass through the device, the device can completely separate the intact fuel balls from the fuel ball fragment and the graphite dust from the graphite ball fragment, thus effectively preventing the fragment and dust from entering the baffle and affecting its normal operation, thereby improving the normal operation of the fuel loading and unloading system and increasing the reliability of the system operation.

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.

In order to ensure the structural integrity of a hood against a load such as pressure exerted from the outside of the hood, the present invention provides a steam dryer for a boiling water-type light-water reactor which ensures the stiffness of the hood without making the hood plate itself thicker, and which avoids lengthening the fabrication time or complicating the fabrication procedure. This steam dryer for a boiling water-type reactor comprises: a hood for guiding, to a steam dryer unit, steam separated by a gas-water separator installed above a reactor core made of a plurality of fuel assemblies disposed inside a nuclear reactor pressure vessel; a support ring; a skirt; and the steam dryer unit having a corrugated plate therein, the steam dryer being characterized in that the hood is made of a first hood plate disposed in the vertical direction, a reinforcing plate mounted on the first hood plate, a second hood plate disposed vertically above the first hood plate and the reinforcing plate, and a side plate disposed between the first hood plate and the steam dryer unit, and the reinforcing plate is formed such that the cross-section thereof is a concavo-convex shape.

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 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 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).

The present invention relates to a nuclear reactor, comprising a reactor vessel, a heat exchange device, a drive device, and a water supply device. The reactor vessel contains liquid coolant and an air cavity. The heat exchange device is disposed within the reactor vessel, wherein the inner wall forms a first cavity for accommodating a core, and the outer and inner walls together define a second cavity. One end of the drive device is connected to the air cavity, and the other end is disposed near the core. The water supply device is in communication with the second cavity. When the nuclear reactor is operating normally, the drive device causes the average density of the liquid coolant above the top of the core to be lower than the average density of the lower cavity below the core, thereby generating a gravity difference that drives the liquid coolant to circulate within the reactor vessel, transferring heat generated by the core. When the nuclear reactor malfunctions, the circulating cold source water in the water supply device circulates in the second cavity, continuously exchanging heat with the core to remove excess heat from the core. The present invention eliminates the need for a large main pump, reduces costs, and improves the safety and reliability of the nuclear reactor.

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.