86results about "Facility decommissioning" patented technology

The apparatus for reverse cutting of a structure, according to the present invention, comprises a pair of reverse pulley devices, the pair of pulley devices comprising: main tubes inserted and mounted so as to penetrate two holes; a pair of first support pulleys mounted on the inner sides of the front ends of the main tubes to rotatably support a wire; a pair of second support pulleys mounted on the inner sides of the rear ends of the main tubes to rotatably support the wire; openings which are cut in portions where the two main tubes face one another and allow the wire to move; and cap covers which are coupled to the front ends of the main tubes to seal same.

A system for servicing a nuclear reactor module comprises a crane operable to attach to the nuclear reactor module, wherein the crane includes provisions for routing signals from one or more sensors of the nuclear reactor module to one or more sensor receivers.

There is provided a method of dismantling a dousing tank of a heavy water reactor structure. The method of dismantling a dousing tank of a heavy water reactor structure includes: selecting a dousing tank of a heavy water reactor structure for dismantlement; dismantling a 1st structure inside the dousing tank; fixing the dousing tank to a dome structure portion provided above the dousing tank through a mounting port; positioning a cutting unit on a 2nd structure including a concrete structure inside the dousing tank; and performing, by the cutting unit, cutting based on a forward or reverse rotation manner on an inner peripheral surface of the dousing tank.

The invention relates to a robot for the dismantling of nuclear facilities. The robot comprises an upper carriage and an undercarriage rotatably connected to the upper carriage. At least three support legs are arranged on the undercarriage for lateral or horizontal support of the robot. The undercarriage has at least one support system for vertical support and / or for vertical or height adjustment of the robot. The upper carriage has at least one arm, at the end of which at least one attachment is arranged and / or can be arranged.

Provided is a method and apparatus for dismantling a nuclear power plant. The method for dismantling a nuclear power plant includes the steps of selecting a nuclear reactor to be dismantled and performing a dismantling operation of the nuclear power plant on the reactor by a nuclear reactor dismantling device, the nuclear reactor dismantling device including a frame unit for shielding the reactor to prevent external release of contaminants from the reactor, a cutting module for performing a cutting operation on the reactor, a bar-shaped interlocking and fastening module having a hollow area formed therein and entering the internal area of ​​the reactor to interlock the frame unit and the reactor, and a purge module for entering the hollow area of ​​the interlocking and fastening module to purge the inside of the reactor to ensure safety during the cutting operation.

A cutting device and method for cutting a plurality of parallel fingers of at least one core component in a plane orthogonal to the longitudinal axis of the fingers, the fingers being adapted to be introduced into the fuel assembly of a nuclear reactor core. The cutting device includes: a first bearing plate including a plurality of abutments respectively for a finger to be cut. The cutting device also includes a rotatable cutting plate including a corresponding cutting blade for each abutment in the first bearing plate, the rotatable cutting plate being rotatable about a rotating axis between a starting position and a final position in a circumferential direction, the first bearing plate and the cutting plate being parallel to each other, the cutting blades including a plurality of groups of cutting blades. For each group of cutting blades the angle between the cutting blades and the corresponding abutments is different to an angle between the cutting blades and the corresponding abutments of any other group of cutting blades.

Method for renovating a nuclear power plant initially comprising at least one light water nuclear reactor (LWR), in particular a pressurized water reactor (PWR) or boiling water reactor (BWR), replaced by at least one integrated modular nuclear reactor (SMR). The invention essentially consists of a method for retrofitting a nuclear power plant which consists of removing and removing all the components of the primary circuit with the exception of the LWR reactor vessel, which is emptied of all material and neutralized, then replacing in place of a part of these components with sub-assemblies each consisting of an integrated SMR reactor and a mixed concrete / metal structure which acts both as a reactor vessel for the SMR reactor, advantageously filled with water, as an anchor for the SMR inside the reactor building and advantageously as a contribution to the third containment barrier, while minimally modifying the infrastructure of the reactor building. Figure for abstract: Fig. 14D

A structural panel (1) for a modular containment system (20). The structural panel (1) comprises a flat rectangular panel (2) having a front side (3), a rear side (4), and four edges (5); four side panels (6), each extending rearwardly from a different one of the four edges (5) perpendicular to the rear side (4) of the rectangular flat panel (2), each having a front edge affixed to the rectangular flat panel (2) and a rear edge; and four rear tabs (8), each extending inwardly from a rear edge of a different one of the four side panels (6), and being parallel to the rear side (4) of the flat panel (2). A modular containment system (20) constructed of a plurality of structural panels (1) is also provided. The modular containment system (20) is formed by affixing the structural panels (1) to one another.

Vehicle (104) for a device (116) for removing contaminated material from a wall (100), comprising wheels or tracks (106), a drive mechanism, a chassis (105) and a height-adjustable support system (114) for the device (116) arranged on the chassis (105), characterized in that the support system (114) is displaceable and pivotable relative to the chassis (105).

This dismantling system (1) for a nuclear installation comprises a fixed part (2), configured to be connected to the nuclear installation, a mast (4) supported by the fixed part and extending downwards along a vertical axis from the fixed part, the mast having at least two modules (41, 43) assembled together, a tool (8), carried by the lowest module (41) of the mast, and a handling system (6), adapted to add an additional module (43) to the top of the mast and then lower the mast, so as to lengthen the mast, and to raise and then remove the highest module (43) of the mast, so as to shorten the mast. The handling system comprises a lifting beam intended to be attached to the highest module of the mast, and two actuators on which the lifting beam is intended to rest. The deployment of the actuators causes the lifting beam to move between an upper and a lower position.

A method of decommissioning a nuclear facility including a plurality of sandboxes that cover a plurality of upper penetration holes, includes: exposing the plurality of pipes through the plurality of upper penetration holes by removing the plurality of sandboxes; and cutting the plurality of pipes through the plurality of upper penetration holes.

The application belongs to the field of coolant purification of nuclear power plants, and particularly relates to a coolant circulating purification system and method applied to nuclear power plants. The system comprises: a primary coolant storage system cation exchanger, a primary coolant storage system anion exchanger, a primary coolant storage system resin collector, a primary coolant treatment system cation exchanger, a primary coolant treatment system anion exchanger, and a primary coolant treatment system resin collector. A third bypass is arranged in front of the inlet of the primary coolant treatment system cation exchanger, and a valve is arranged in the third bypass to connect to a position after the inlet of the primary coolant storage system cation exchanger. The application can effectively improve the reliability of the primary coolant treatment system in purifying the coolant, and can switch the primary coolant storage system ion exchanger to operate in the case of failure of the primary coolant treatment system ion exchanger.

A reactor vessel cutting device, including a hoist system; a mechanical cutter; and a cutter carriage secured to the hoist system to be suspended by the hoist system. The cutter is secured to a carriage body of the cutter carriage to be carried by the carriage body. The cutter carriage includes a stabilization system secured to the carriage body and operable to hold the cutter carriage against a vessel wall of the reactor vessel. The cutter carriage is operable to apply the cutter to the vessel wall in a mechanical cutting operation when the cutter carriage is held against the vessel wall by the stabilization system. The hoist system is configured to bear the weight of the cutter and the cutter carriage during the mechanical cutting operation and the stabilization system is configured to carry cutting forces generated during the mechanical cutting operation.

A dismantling and decontamination system of biodegradable concrete of a nuclear power plant according to an exemplary embodiment includes: a dismantling device for dismantling an in-core instrument installed under biodegradable concrete to form a lower penetrated part of the biodegradable concrete; a decontamination device inserted inside the biodegradable concrete for decontaminating radioactive waste of the inner wall of the biodegradable concrete; a waste receiving device movable through the lower penetrated part of the biodegradable concrete; and a blocking device for blocking the upper opening of the biodegradable concrete to block an outflow of the radioactive dust.

Disclosed is a method for dismantling a D2O tank of a heavy water reactor installed on an inner wall of a containment building and surrounded by a concrete structure, the method including cutting and opening a front surface of the concrete structure so as to dismantle the D2O tank, descending the D2O Tank by using an dedicated dismantling apparatus, and taking the D2O tank out of the containment building by using a crane, and a dismantling apparatus.

An apparatus for cutting a curved wall of a structure, of the present invention, may comprise; a guide rail mounted on a curved wall of a structure; a chain mounted on the guide rail; a support plate supported by the guide rail so as to be movable by means of a plurality of rollers; a body case coupled to the support plate; a sprocket mounted at the body case so as to rotate while engaged with the chain; a driving motor mounted in the body case so as to rotate the sprocket; a cutting blade rotatably mounted at the front of the body case; and cutting motor mounted in the body case so as to rotate the cutting blade.

A module for use in a robot-deployable modular accessibility system. The module includes a block, an attachment portion, and a gripping portion. The attachment portion is configured to attach the module to another module. The gripping portion is configured to enable a robot to grip and handle the module. The robot-deployable modular accessibility system includes a plurality of modules. The plurality of modules is configured to be assembled by the robot to form a structure. Also, a method of assembling the robot-deployable modular accessibility system by a robot.

The invention relates to a modular ventilation system for a building, comprising: - at least one first extraction module comprising a first casing and at least one air extraction unit disposed inside the first casing, the air extraction unit being adapted to draw air from outside the first casing via the first air inlet opening and to inject the drawn air towards the first air outlet opening, and - at least one second filtration module comprising a second casing and at least one filtration unit disposed inside the second casing, the air filtration unit being adapted to filter air coming from the second air inlet opening, the filtered air being discharged towards the second air outlet opening, the modular ventilation system being configured to be able to selectively adopt a functional configuration and a storage and / or transport configuration. Figure for the abstract: Fig. 6

The present invention relates to a cutting device (3) for cutting a plurality of parallel fingers (7) of at least one core component (5) in a plane orthogonal to the longitudinal axis of the fingers (7), the fingers (7) being adapted to be introduced into the fuel assembly of a nuclear reactor core, the cutting device (3) comprising: a first bearing plate (30, 46) comprising a plurality of abutments (54, 56) respectively for a finger (7) to be cut, wherein the cutting device (3) further comprises: a rotatable cutting plate (34) comprising a corresponding cutting blade (42a, 42b, 42c) for each abutment (54, 56) in the first bearing plate, the rotatable cutting plate (34) being rotatable about a rotating axis (Y) between a starting position and a final position in a circumferential direction (C), the first bearing plate (30, 46) and the cutting plate (34) being parallel to each other, the cutting blades (42a, 42b, 42c) comprising a plurality of groups of cutting blades (42a, 42b, 42c), wherein for each group of cutting blades (42a, 42b, 42c) the angle between the cutting blades (42a, 42b, 42c) and the corresponding abutments (54, 56) is different to an angle between the cutting blades (42a, 42b, 42c) and the corresponding abutments (54, 56) of any other group of cutting blades

A method of handling spent nuclear fuel assemblies immerses the spent nuclear fuel assemblies in water in a relatively short time period when compared to traditional methods. A spent nuclear fuel assembly is removed from a nuclear reactor, an inert gas is applied to the fuel assembly, moisture content in the inert gas is gradually increased as it is applied to the fuel assembly, and the fuel assembly is immersed in water. The fuel assembly is immersed relatively quickly, within about 2 hours or less, which improves safety and allows normal processing and handling equipment to care for the fuel assembly. The fuel assembly may then be loaded into a cask for long-term storage and / or disposal.

Disclosed is a method for dismantling metal equipment of a heat transport system in a containment building at a nuclear power plant, the metal equipment including a heat transport pump, a steam generator, a pressurizer, a condenser, and a D2O tank, including: a first dismantling step of dismantling the heat transport pump; a second dismantling step of dismantling the steam generator after the first dismantling step; a third dismantling step of dismantling the pressurizer after the second dismantling step; and a fourth dismantling step of dismantling the condenser and the D2O tank after the third dismantling step.

The large concrete container cutting, dismantling, and disposal suitability evaluation system according to an embodiment of the present invention comprises: a vertical and horizontal cutting device for a large concrete container and a solidified material, the device being capable of vertically and horizontally cutting a concrete container and a solidified material; a cut block separation device for separating a solidified material block, a grounding concrete block, and an external container block from the cut container and the solidified material cut blocks produced by the vertical and horizontal cutting device for a container and a solidified material; and a suitable disposal device for determining self-disposal or disposal suitability for each of the solidified block, the grounding concrete block, and the external container block individually separated by the cut block separation device, the suitable disposal device enabling self-disposal or disposal when the self-disposal or disposal suitability is satisfied..

This disclosure belongs to the field of nuclear power technology, specifically to a method for measuring and controlling oxidation operations during nuclear power plant overhauls. This disclosure provides a complete chemical control and analysis method for nuclear power plant overhaul oxidation operations, encompassing preparation for overhaul oxidation operations, dosing, control and measurement of the oxidation operation platform, and purification. This method clarifies the parameters and control and analysis methods at each stage, improving oxidation operation performance and measurement capabilities. Furthermore, this method is simple to operate, significantly reduces costs, and is highly safe, making it suitable for widespread adoption.

The present application relates to a nuclear facility decommissioning model processing method and apparatus, a computer device, a storage medium, and a computer program product. The method includes: acquiring a three-dimensional model of a processed object; extracting basic attributes of the processed object and a decommissioning tool behavior attributes of the decommissioning tool, and engineering attributes of the decommissioning tool from pre-established nuclear facility decommissioning information modeling (DIM); performing action simulation on the decommissioning tool according to the action constraints of the decommissioning tool and the radiation field dose meshes of the processed object; simulating action simulation results of the three-dimensional models of the processed object and the decommissioning tool based on the preset interactive mode between the processed object and the decommissioning tool, to obtain a target simulation result of the processed object and the decommissioning tool; and traversing at least two processed objects and the decommissioning tool, and generating a simulated decommissioning processing plan for the processed objects. With the method, a backend developer is not required to program various models, which improves efficiency of decommissioning.

According to an aspect, there is provided a lattice sleeve and shield plug assembly. The assembly includes a shield plug comprising a flange and a plug and a lattice sleeve. The lattice sleeve includes a sleeve to engage a fuel channel after removal of an end fitting, a face with an opening configured to receive the shield plug through the face and into the fuel channel, and a stud configured to insert through a stop tube through to an end shield. According to an aspect, there is provided a shield plug including a flange, a plug, and one or more barbs configured to engage the interior of a lattice site. According to an aspect, there is provided a tool to handle a payload. The tool including an arm equipped with a gripping member.

Disclosed is a method for dismantling a D20 tank of a heavy water reactor installed on an inner wall of a containment building and surrounded by a concrete structure, the method including cutting and opening a front surface of the concrete structure so as to dismantle the D20 tank, descending the D20 Tank by using an dedicated dismantling apparatus, and taking the D20 tank out of the containment building by using a crane, and a dismantling apparatus.

Method (S100) for measurements on surfaces (11, 12) in a nuclear facility (10), comprising Creating (S110) a digital model (21) of a first surface (11) of the nuclear facility (10); Overlaying (S120) the digital model (21) of the first surface (11) with a first digital raster grid (31) comprising a plurality of raster elements (35, 36); Assigning (S130) a first raster element (35) of the plurality of raster elements (35, 36) of measurement information for measuring a part of the first surface (11) corresponding to the first raster element (35) with a detector for measuring radioactive radiation (40); and Displaying (S140) the first raster element (35) in the first raster grid (31) together with the associated measurement information on a digital display device (51), the method further comprising Recording (S210) the first surface (11) with an optical sensor unit (52); and superimposing display (S220) of the recorded first area (11) together with the first grid (31) and the first grid element (35) with the associated measurement information on the digital display device (51).

A nuclear instrument removal system (NIRS) for use with processing a nuclear instrument is provided. The NIRS includes a control system and an instrument processing system configured for use with the control system. The instrument processing system includes a counter module configured to receive the nuclear instrument, a feed and cut module removably coupled to the counter module and configured to receive the nuclear instrument after passing through the counter module, and a spool module removably coupled to the feed and cut module and configured to be mounted to a cask assembly. The spool module includes a spool canister and a drive shaft associated with the canister. The drive shaft is configured to rotate about an axis in order to spool a cold portion of the nuclear instrument within the canister.

Method (100) for carrying out a release process of a, in particular at least partially radioactive, measured material (10), comprising the following steps: - Recording (111) preliminary activity data (11) of the sample (10) depending on a preliminary investigation (110), - Preparation of the preliminary evaluation (112) of the preliminary investigation (110) of the measured material (10), - Planning (113) a subsequent part of the release process depending on the preliminary evaluation (112) and on a method of carrying out a decision measurement (140) for the release (150) of the measured item (10), - Performing the decision measurement (140) in a clearance measurement facility (1), wherein the decision measurement (140) is carried out using the method in which a detection (141) and an evaluation (142) of a radioactive activity distribution in the measured material (10) is carried out, wherein the preparation of the preliminary evaluation (112) of the measured material (10) is carried out taking into account the preliminary activity data (11) with regard to a specification criterion (111.1) for specifying at least a part of the release process by the measured material (10), wherein the release process begins with the dismantling of the measured material (10) in a nuclear facility (8), wherein the preliminary investigation (110) takes place in the nuclear facility (8) to record (112) the preliminary activity data (11).

In a method for handling spent nuclear fuel assemblies, the spent nuclear fuel assemblies are immersed in water for a relatively short period of time compared to conventional methods. The spent nuclear fuel assemblies are removed from a nuclear reactor, an inert gas is applied to the fuel assemblies, the water content of the inert gas is gradually increased as the inert gas is applied to the fuel assemblies, and the fuel assemblies are immersed in water. The fuel assemblies are immersed relatively quickly, for approximately two hours or less. This improves safety and allows the fuel assemblies to be cared for using conventional processing and handling equipment. The fuel assemblies may then be loaded into casks for long-term storage and / or disposal.