323results about "Heterogenous reactors" patented technology

A control method and system for controlling a nuclear power plant includes, in the absence of detection of an imbalance between a primary power signal (S1) and a secondary power signal (S2), the implementation of a setpoint-following mode. The nuclear power plant is controlled as a function of an operational power setpoint (COP), and in the event of detection of an imbalance, automatically implementing a power-limiting mode, including the calculation of a target equilibrium power (PEC) less than or equal to the primary power (P1) and less than or equal to the secondary power (P2), and controlling the nuclear power plant (2) as a function of the target equilibrium power (PEC).

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.

An integrated energy system including a power plant is discussed herein. In some examples, the integrated energy system may include a power plant configured to generate steam, a hydrothermal decomposition reactor configured to receive at least a portion of the steam (H2O) from the power plant to react with Methane (CH4) within the hydrothermal decomposition reactor to produce Hydrogen (H2) and Carbon Dioxide (CO2), a first separation unit configured to separate the Hydrogen (H2) and the Carbon Dioxide (CO2), a Solid Oxide Stack configured to receive at least a portion of the Carbon Dioxide (CO2) and to produce Carbon Monoxide (CO), a second separation unit configured to separate the Carbon Dioxide (CO2) from the Carbon Monoxide (CO), and a methanol synthesis reactor configured to receive at least a portion of the Hydrogen (H2) and at least a portion of the Carbon Monoxide (CO) to produce Methanol (CH3OH).

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.

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.

Advances in creating carbon-free emissions continue with nuclear power. This invention aims to achieve an environment with less radioactive pollution from pebble bed reactors. This invention uses stainless steel, niobium, or ferroniobium which does not form dust like carbon pebbles. The pebble bed reactor coolant will pass over the coated pebbles and not pick up radioactive dust as in old pebble models. This new design is more complicated and costs more than current designs. However, the cleanup of a used pebble bed reactor will be far less, and the health of the workers will be greatly improved through use of this invention. This practical design will enable pebble bed reactors to become approved at a faster rate than old designs.

An integrated energy system including a power plant is discussed herein. In some examples, the integrated energy system may include at least one nuclear reactor and electrical power generation system configured to generate steam and electricity, a water treatment plant configured to produce Sodium Hydroxide (NaOH) from salt water, a Sodium Formate (HCOONa) production plant configured to receive the Sodium Hydroxide (NaOH) to produce Sodium Formate (HCOONa), a Thermal Decomposition reactor configured to receive the Sodium Formate (HCOONa) and configured to receive at least a first portion of the steam or at least a second portion of the electricity from the power plant to indirectly heat the Thermal Decomposition reactor to produce Hydrogen (H2), Carbon Dioxide (CO2), and Carbon Monoxide (CO) from the Sodium Formate (HCOONa), and a Methanol (CH3OH) reaction chamber configured to receive the Hydrogen (H2), the Carbon Dioxide (CO2), and the Carbon Monoxide (CO) to produce Methanol (CH3OH).

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 nuclear engine system includes: a pump configured to pump a propellant; a fuel element including a set of nuclear fuel particles; a moderator configured to surround the fuel element and defining a set of moderator coolant channels configured to cool the moderator; a reflector including a neutron-reflecting material and a reflector coolant channel arranged within the reflector to cool the reflector, the reflector configured to at a first time, operate in a closed configuration to reflect neutrons to the fuel element to increase an energy flux, and at a second time, operate in an open configured to leak neutrons out of the engine system to decrease the energy flux; and a thrust nozzle configured to outlet propellant from the fuel element to produce thrust including a nozzle coolant channel arranged within a wall of the thrust nozzle configured to cool the nozzle.

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 invention relates to a nuclear power plant with improved steam explosion mitigation performance, the nuclear power plant comprising: an integrated reactor including a hemispherical first portion inside which a core and a steam generator are positioned and a cylindrical second portion positioned above the first portion; a containment vessel for accommodating the integrated reactor, having a cooling space surrounding the integrated reactor, and including a first containment portion facing the first portion and a second containment portion facing the second portion; and a lower structure which is located in the cooling space and is located under the integrated reactor, wherein the lower structure comprises a first plate and a second plate disposed below the first plate, and the first plate and the second plate are arranged in a horizontal direction and face each other.

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.

The present invention relates to a nuclear power plant considering defense-in-depth levels, comprising: a plurality of reactors; common facilities that are jointly used by the plurality of reactors; and a measurement control system including a plurality of measurement control platforms corresponding to various operating conditions, wherein the measurement control system separately measures and controls the respective reactors and the common facilities.

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.

To solve such a problem that fuel balls are blocked when a fuel element runs in a dust and chipping environment is not fully considered.SOLUTION: A pressure-bearing unit comprises a pressure-bearing box body and an end flange, an allocation space is formed in the pressure-bearing box body, and the end flange is connected to the opening side of the pressure-bearing box body; a rotor unit comprises a rotating shaft, a material taking disc and a material distributing disc, and the material taking disc and the material distributing disc are connected to the rotating shaft in a sleeving mode and installed in the allocation space; a bush unit comprises a first bush and a second bush, and the first bush and the second bush are connected to the material taking disc and the material distributing disc in a sleeving mode. A non-return unit is sleeved on the rotating shaft and is connected with the end flange; the bearing assembly comprises a first bearing and a second bearing, and the first bearing and the second bearing are arranged in the end flange and the pressure bearing box body; a driving unit comprises a servo motor, a speed reducer, a coupler and a magnetic synchronizer which are connected in sequence.SELECTED DRAWING: Figure 1

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) comprising: a nuclear reactor core (20) comprising a fuel system (30) and a moderator (35); a plurality of heat pipes (40) that each at least partially extend within the core; a neutron reflector (70) disposed around the core; a plurality of primary neutronic control elements (50) disposed around the core; a secondary neutronic control element (60) configured to be inserted into a void in the core; a fluid circuit (90) comprising a first heat exchanger (92), a second heat exchanger (96), and a pump (94); and an open Brayton power system comprising: a compressor (102); one or more turbines (104, 106); and a generator (108), and being arranged such that air flows through the compressor, the second heat exchanger and the one or more turbines, wherein the second heat-exchanger is configured to permit heat from the core to be transferred to the compressed air from the compressor, and the one or more turbines drive the compressor and the generator.

The invention belongs to the technical field of nuclear power station voltage stabilizer spray heads, and particularly relates to a voltage stabilizer spray head. The device is of a centrifugal structure and comprises a shell and a rotary water core, the shell and the rotary water core are integrally formed through 3D printing, the shell is of an irregular structure and is divided into a shell upper portion, a shell middle portion and a shell lower portion, the shell upper portion is of an internal thread boss structure, the shell middle portion is of a hollow cylindrical structure, the rotary water core is located above the shell middle portion, and the shell lower portion is of a hollow cylindrical structure. And the lower part of the shell is provided with a vortex chamber and a spray head. When the optimal lead, the optimal groove width, the optimal groove depth, the optimal shell taper and the optimal shell nozzle diameter are guaranteed, the distribution uniformity of the mist amount can be guaranteed, and the working condition use requirements are met.

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.

Anti-debris filter for lower nozzle of nuclear fuel assembly with variable thickness The anti-debris filter is formed of a grid (32) having a first face (32A) and a second face (32B) opposite each other, the grid (36) having passages (36) extending between an inlet (36A) located on the first face (32A) and an outlet (36B) located on the second face (32B) for the flow of a coolant through the grid (32), the grid (32) having a first zone (Z1) and at least one second zone (Z2) in which the grid (32) has a thickness greater than that of the grid (32) in the first zone (Z1), passages (36) being present in the first zone (Z1) and in each second zone (Z2). Figure for abstract: Figure 7

A curvilinear electromagnetic pump is configured to follow a curve, such as by joining together multiple linear pump segments offset at an angle relative to one another. A curvilinear electromagnetic pump can be curved in two dimensions or in three dimensions. A curvilinear electromagnetic pump allows for more efficient placement of components and systems within a reactor vessel, significantly reducing the height of the reactor vessel compared to vertically positioned linear pumps. A curvilinear electromagnetic pump can follow the curvature of the reactor vessel wall and can be positioned generally near the bottom of the reactor vessel.

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).