A method for cavitation peening of an internal surface of a hollow metal part including an internal surface is provided. The method includes installing a first seal in sealing contact with the internal surface; installing a second seal axially offset from the first seal, the first seal and the second seal defining a chamber within the hollow part axially therebetween; providing a cavitation peening nozzle of a cavitation peening tool within the chamber; pressurizing the chamber; and cavitation peening the internal surface via the cavitation peening nozzle while actuating the cavitation peening nozzle within the pressurized chamber. An apparatus for cavitation peening of an internal surface of a hollow metal part, the hollow part including an internal surface is also provided.
C21D 7/04 - Modifying the physical properties of iron or steel by deformation by cold working of the surface
C21D 8/10 - Modifying the physical properties by deformation combined with, or followed by, heat treatment during manufacturing of tubular bodies
C21D 9/08 - Heat treatment, e.g. annealing, hardening, quenching or tempering, adapted for particular articlesFurnaces therefor for tubular bodies or pipes
C21D 9/50 - Heat treatment, e.g. annealing, hardening, quenching or tempering, adapted for particular articlesFurnaces therefor for welded joints
C21D 10/00 - Modifying the physical properties by methods other than heat treatment or deformation
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
2.
METHOD OF PEENING AN OBSTRUCTED REGION OF A METAL ASSEMBLY
A method for peening an obstructed region of a metal assembly that is obstructed by an obstructing part of the metal assembly is provided. The method includes determining an optimal peening path for treating the obstructed region irrespective of the obstructing part; identifying a portion of the obstructing part within the optimal peening path; determining a section of the portion of the obstructing part that is removable without affecting a mechanical integrity and functionality of the obstructing part; removing, by machining, the section so as to create additional space along the optimal peening path; and peening the obstructed region, a path of the peening at least partially crossing through the additional space. A method for peening a nuclear reactor pressure vessel is also provided.
G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
3.
APPARATUS FOR CAVITATION PEENING OF A METAL PART AND METHOD THEREFORE
A sealing device is provided to form a sealed region about one or more surfaces to be treated. The sealing device has an open end with a rim configured to matingly engage a treatment surface. The sealing device is braced both vertically and laterally, and the sealed region is flooded and pressurized. A peening nozzle and manipulating tooling are positioned within an interior volume of the sealing device. Pressurized fluid is ejected from the nozzle causing the formation of cavitation bubbles. The nozzle flow causes the cavitation bubbles to settle on the surfaces to be treated. The collapsing impact of the cavitation bubbles imparts compressive stress in the materials of the treatment surfaces.
C21D 7/06 - Modifying the physical properties of iron or steel by deformation by cold working of the surface by shot-peening or the like
C21D 10/00 - Modifying the physical properties by methods other than heat treatment or deformation
B23P 9/04 - Treating or finishing by hammering or applying repeated pressure
G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
4.
CONTROL ROD DRIVE MECHANISM INNER DIAMETER ANNULUS ULTRA HIGH PRESSURE CAVITATION PEENING
A sealing member is provided to create a sealed region about an annulus formed between an inner body, such as a thermal sleeve, and an outer body, such as a control rod drive housing nozzle. Liquid is introduced into the sealed region to create a flooded region, which is pressurized to a desired level. A nozzle is provided into the flooded region, the nozzle being configured to fit within the annulus. Pressurized fluid is ejected from the nozzle, causing the formation of cavitation bubbles. The nozzle flow causes the cavitation bubbles to settle on the surfaces forming the annulus. The collapsing impact of the cavitation bubbles imparts compressive stress in the materials of the surfaces forming the annulus.
C21D 7/04 - Modifying the physical properties of iron or steel by deformation by cold working of the surface
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
B23P 9/02 - Treating or finishing by applying pressure, e.g. knurling
5.
METHOD OF PLATINUM INJECTION INTO A NUCLEAR REACTOR
A method of injecting platinum into a Boiling Water Reactor is provided. The Boiling Water Reactor includes a reactor core including a plurality of fuel rods having a zirconium alloy cladding. The method includes injecting a first platinate compound and a second platinate compound into a nuclear reactor. The first platinate compound is non- alkalizing and the second platinate compound includes an alkalizing element. The first platinate compound and the second platinate compound are injected such that the second platinate compound injected does not exceed a predetermined threshold or such that a value related to an alkalizing element of the second platinate compound does not exceed a predetermined threshold.
G21C 19/30 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products
6.
CONVECTIVE DRY FILTERED CONTAINMENT VENTING SYSTEM
A dry FCVS for a nuclear reactor containment is provided. The dry FCVS includes a housing and a round and/or elongated aerosol filter inside the housing for removing contaminant aerosols from gas passing through the housing during venting of the containment. The housing includes at least one inlet portion configured for directing gas into the aerosol filter during the venting of the containment and an outlet portion for gas filtered by the aerosol filter during the venting of the containment. The dry filtered containment venting system is arranged and configured such that when a flow of gas through the outlet portion is closed off at least one of convective, radiant and conductive heat transfer removes decay heat of aerosols captured in the aerosol filter.
G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details
G21C 19/303 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for gases
A dry FCVS for a nuclear reactor containment is provided. The dry FCVS includes a housing and a round and/or elongated aerosol filter inside the housing for removing contaminant aerosols from gas passing through the housing during venting of the containment. The housing includes at least one inlet portion configured for directing gas into the aerosol filter during the venting of the containment and an outlet portion for gas filtered by the aerosol filter during the venting of the containment. The dry filtered containment venting system is arranged and configured such that when a flow of gas through the outlet portion is closed off at least one of convective, radiant and conductive heat transfer removes decay heat of aerosols captured in the aerosol filter.
A dry FCVS for a nuclear reactor containment is provided. The dry FCVS includes a housing and a round and/or elongated aerosol filter inside the housing for removing contaminant aerosols from gas passing through the housing during venting of the containment. The housing includes at least one inlet portion configured for directing gas into the aerosol filter during the venting of the containment and an outlet portion for gas filtered by the aerosol filter during the venting of the containment. The dry filtered containment venting system is arranged and configured such that when a flow of gas through the outlet portion is closed off at least one of convective, radiant and conductive heat transfer removes decay heat of aerosols captured in the aerosol filter.
This invention provides an improved replacement nozzle configuration. The nozzle is provided in two pieces. A first nozzle piece is a relatively thin-walled cylinder that is inserted into the counterbored opening of a pressure vessel after the protruding portion of the original nozzle has been removed. A weld pad is attached to the pressure vessel, to which the first replacement nozzle piece is welded. Due to the reduced dimensions of the thin-walled first nozzle portion, a much smaller weld pad and pad-to-nozzle weld can be used. A second nozzle piece is positioned within the first nozzle piece and welded thereto. The existing instrumentation piping is connected to the free end of the second nozzle piece. This two-piece replacement nozzle can be implemented in approximately half the time required for the one-piece design.
B23K 31/02 - Processes relevant to this subclass, specially adapted for particular articles or purposes, but not covered by any single one of main groups relating to soldering or welding
A clamp for a boiling water reactor feedwater sparger is disclosed and claimed. The clamp assembly includes a base and cooperating bolt, a reaction arm, and a cross pin. With the exception of the cross pin, the parts are pre-assembled and then installed over the sparger pin head. The cross pin is then installed through the clamp base and through the sparger pin. The clamp bolt is then tightened, locking the clamp and sparger pin. The clamp provides an increased surface area that extends 360° around the sparger pin. This increased surface area reduces the likelihood of wear of the clamp or bracket during plant operation, and the clamp restores the position of the sparger pin head relative to the contact surface of the end bracket if the original pin head or bracket is worn.
A clamp for a boiling water reactor feedwater sparger is disclosed and claimed. The clamp assembly includes abuse and cooperating bolt, a reaction arm, and a cross pin. With the exception of the cross pin, the parts are pre-assembled and then installed over the sparger pin head. The cross pin is then installed through the clamp base and through the sparger pin. The clamp bolt is then tightened, locking the clamp and sparger pin. The clamp provides an increased surface area that extends 360° around the sparger pin. This increased surface area reduces the likelihood of wear of the clamp or bracket during plant operation, and the clamp restores the position of the sparger pin head relative to the contact surface of the end bracket if the original pin head or bracket is worn.
G21C 15/02 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements
G21C 13/032 - Joints between tubes and vessel walls, e.g. taking into account thermal stresses
G21C 15/22 - Structural association of coolant tubes with headers or other pipes, e.g. in pressure tube reactors
G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
12.
NUCLEAR FUEL ROD CLADDING INCLUDING A METAL NANOMATERIAL LAYER
A nuclear fuel cladding (16) is provided. The nuclear fuel cladding (16) includes a base cladding (18) and at least one nanomaterial layer (20) deposited on a surface of the base cladding (18), the nanomaterial layer (20) having an average grain size of between 5 to 400 nanometers. A method of manufacturing nuclear fuel cladding (16) is also provided. The method includes depositing nanoparticles on a base cladding (18) to form at least one nanomaterial layer (18), the nanoparticles having an average grain size of between 5 to 400 nanometers.
A method of operating a nuclear reactor is provided. The method includes defining a layer increment (24) of a deposit layer (10) modeling a deposit on a heat transfer surface (12) of the nuclear reactor (200); periodically updating a thickness of the deposit layer (10) by adding the layer increment (24) to the deposit layer (10); recalculating properties of the deposit layer (10) after each layer increment (24) is added to the deposit layer (10); determining a temperature related variable of the heat transfer surface (12) as a function of the recalculated properties of the deposit layer (10); and altering operation of the nuclear reactor (200) when the temperature related variable of the heat transfer surface (12) reaches a predetermined value. A method of modeling a deposit on a heat transfer surface (12) of a nuclear reactor (200) is also provided.
F01K 13/02 - Controlling, e.g. stopping or starting
G21D 3/10 - Regulation of any parameters in the plant by a combination of a variable derived from neutron flux with other controlling variables, e.g. derived from temperature, cooling flow, pressure
A steam generator anti-vibration bar (A.VB) that extends across multiple rows of tubes and contains a string of sensors is disclosed and claimed. The instrumented anti-vibration bar includes a housing and cover that cooperate to enclose the sensors. The sensors capture data resulting from the fluid flow passing through the steam generator during plant operation. Exemplary sensors that may include: strain gauges to measure loads imposed on the AVBs by the tubes; temperature sensors to sample thermal-hydraulic conditions within the tube bundle; pressure and flow detectors to sample hydraulic conditions within the tube bundle; inductive coils to measure in-plane and out-of-plane motion; and accelerometers to measure AVB motions imposed by surrounding tubes. The captured data is transmitted out of the steam generator in known fashion.
F22B 37/20 - Supporting arrangements, e.g. for securing water-tube sets
F28F 9/013 - Auxiliary supports for elements for tubes or tube-assemblies
F28D 7/06 - Heat-exchange apparatus having stationary tubular conduit assemblies for both heat-exchange media, the media being in contact with different sides of a conduit wall the conduits having a single U-bend
The present invention provides a drain plug assembly that prevents significant quantities of corium from entering the drain line. By protecting the drain line (101), essentially no high-activity fission products would be released to the reactor building or the environment during a severe accident. The ceramic drain plug assembly includes a drain plug base (10) and a drain plug (20) supported by a steel pedestal (30). The lower surface of the plug has a spherical shape such that the plug can be positioned within the base to block access to the drain opening provided in a central portion of the base. During normal operation conditions, the plug is retained above the base by the pedestal. During a severe accident, when corium comes into contact with the pedestal, it will melt rapidly and the drain plug will drop by gravity, effectively closing the sump drain opening and preventing the flow of corium into the drain line.
A fuel rod (1) for a nuclear fission reactor is disclosed and claimed. The fuel rod includes an elongate hollow cladding (6) configured to retain a nuclear fuel (20) therein. The cladding (6) includes an elongate hollow tube (10). Fiber layers (12) are positioned around the outside surface of the tube (10) or within the tube forming an integral part thereof. Both the tube (10) and the fibers (8) are formed of a ceramic material. A fuel assembly (2) including a plurality of such fuel rods (1) is also disclosed and claimed.
A permanent seal for a refueling canal of a nuclear power plant. The seal includes a support structure and a membrane. The support structure includes a first annular plate with ribs connected to and extending from a lower surface of the annular plate. The support structure is positioned atop the shield wall on the refueling canal floor, encircling and positioned near the annulus. The membrane includes a first end that is connected to the seal ledge and a second end that is connected to the refueling canal floor. The membrane has a stepped profile, with side walls extending substantially perpendicularly from a central annular plate to form a pocket configured to fit over the support structure. Loads imparted to the membrane are transferred through the support structure annular plate and ribs to the refueling canal floor.
A method of operating a power generator is provided. The method includes determining an amount of oxides on a heat transfer surface of the power generator as a function of a concentration of a noble metal substance in the oxides; and altering operation of the power generator when the amount of oxides on the heat transfer surface reaches a predetermined value. A method of operating a nuclear reactor is also provided.
A composite basket assembly (10) for radioactive fuel containment includes a plurality of cast disks (20) disposed together in face-to-face relationship by connecting rods (40) extending through the disks. The disks are formed with interior cells (124) and a grid structure (122) to form discrete compartments for individual fuel assemblies. A neutron-absorbing material (34) is incorporated into the disks to absorb radioactive emissions from the stored fuel.
A horizontal storage module (HSM) assembly generally includes a base portion including a plurality of segments that are vertically layered on top of each other and a lid portion. A method of constructing an HSM assembly generally includes forming a plurality of segments for the base portion of the HSM assembly, and vertically stacking the segments.
A travel system (20) for a canister storage, transfer, or transport system generally includes a support structure (22), at least one traveling device (24) for preparing, inspecting, and/or repairing the canister, and a base ring (26) for supporting the traveling device and providing for rotational movement of the traveling device relative to the support structure.
G01N 27/90 - Investigating or analysing materials by the use of electric, electrochemical, or magnetic means by investigating magnetic variables for investigating the presence of flaws using eddy currents
G01N 21/90 - Investigating the presence of flaws, defects or contamination in a container or its contents
22.
Eddy current method for the characterization of broached tube support plate blockage
A method for determining tube support plate blockage of a steam generator includes the following steps: measuring at least five different eddy current values per tube support intersection; calculating a nominal clean fit radius of flow hole; determining a center signal response; converting the center signal response to a deposit thickness; determine an edge reduction; converting the edge reduction to an edge thickness; calculating the resulting flow hole radius; verifying the reasonableness of the resulting flow hole radius; and determining a virtual calibration range.
A method for recycling AglnCd control rod absorber bar material from a used control rod from a nuclear power plant includes sectioning AglnCd absorber bar from a used control rod into a first section and a second section, the first section having a higher radioactivity than the second section; and recycling the material of the second section of the AglnCd absorber bar.
A tool for inspecting the integrity of fasteners in their environment of use and methods of performing such inspections are disclosed and claimed. The tool includes a probe that matches the internal socket by which the fastener is coupled to the workpiece. The probe contains ultrasonic transducers on flat portions corresponding to flat portions of the socket. The transducers induce angled ultrasonic beams into the fastener to detect flaws therein. The beams are angled so they can be directed to the areas of interest at the head to shank region of the fastener. The presence of a defect such as a crack is determined based on the reply/echoes of the imparted ultrasonic beams.
G01N 29/00 - Investigating or analysing materials by the use of ultrasonic, sonic or infrasonic wavesVisualisation of the interior of objects by transmitting ultrasonic or sonic waves through the object
G01N 29/28 - Investigating or analysing materials by the use of ultrasonic, sonic or infrasonic wavesVisualisation of the interior of objects by transmitting ultrasonic or sonic waves through the object Details providing acoustic coupling
G01D 11/30 - Supports specially adapted for an instrumentSupports specially adapted for a set of instruments
G01N 29/07 - Analysing solids by measuring propagation velocity or propagation time of acoustic waves
G01N 29/22 - Investigating or analysing materials by the use of ultrasonic, sonic or infrasonic wavesVisualisation of the interior of objects by transmitting ultrasonic or sonic waves through the object Details
25.
Method for recycling nuclear control rods and recycled control rod section
A method for recycling AgInCd control rod absorber bar material from a used control rod from a nuclear power plant includes sectioning AgInCd absorber bar from a used control rod into a first section and a second section, the first section having a higher radioactivity than the second section; and recycling the material of the second section of the AgInCd absorber bar.
G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
G21C 21/18 - Manufacture of control elements covered by group
An integrated reactor missile shield and crane assembly (IRMSCA) is disclosed and claimed. The IRMSCA replaces the existing concrete missile shields and reactor services crane. The IRMSCA is moveable such that the missile shield can be moved away from the reactor head, allowing the integral crane to lift the control rod drive mechanism components and other routine loads at the refueling cavity.
G21C 11/00 - Shielding structurally associated with the reactor
G21C 7/20 - Disposition of shock-absorbing devices
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
A tool for inspecting a cell formed by grid beams of a top guide structure in a nuclear reactor is provided. The tool includes a camera; a support structure coupled to the camera for contacting at least one of the grid beams to support the camera within the cell; and at least one actuator moving the camera with respect to the support structure and along one of the grid beams, the at least one actuator coupling the camera to the support structure. A method for inspecting a cell formed by grid beams of a top guide structure in a nuclear reactor is also provided.
A method to perform an analysis of two types of CRUD on a nuclear fuel rod, including providing a nuclear fuel rod with a first and second layer of CRUD on an exterior of the fuel rod; brushing the first layer of CRUD from the fuel rod with a CRUD tool on a selected area; wherein the tool has a brushing device, a force applied to the brushing device on the fuel rod to remove the first layer of CRUD, the force being sufficient enough to perform such removal; collecting the first layer of CRUD from the brushing device, scraping the second layer of CRUD from the fuel rod in the selected area with the tool, wherein the tool has a scraping device and a second force is applied to the tool for scraping, collecting the second layer of CRUD from the scraping device, and analyzing the first layer and second layer of CRUD separately with a scanning electron microscope.
G01C 17/00 - CompassesDevices for ascertaining true or magnetic north for navigation or surveying purposes
G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations
G21C 17/06 - Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination
G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
G01N 23/225 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or by measuring secondary emission from the material using electron or ion microprobes
G01N 1/04 - Devices for withdrawing samples in the solid state, e.g. by cutting
A nuclear power plant is provided including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system and an alcohol injection system connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.
G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 19/307 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for liquids
30.
Method for prediction of light water reactor fuel defects using a fuel condition index
A method to assess light water reactor fuel integrity is presented having the steps of granting access in a nuclear reactor fuel pool to at least one of a discharged fuel rod and a nuclear fuel assembly, calculating an operating flux for the fuel rod, measuring a thickness of CRUD on the fuel rod, measuring a thickness of oxide on the fuel rod, calculating a maximized flux for the at least one fuel rod for a position of the one fuel rod in a nuclear reactor, calculating a maximized deposit for the fuel rod, calculating a maximized oxide thickness for the fuel rod, calculating a fuel condition index of the fuel rod, comparing the fuel condition index to an index constant, and removing the fuel rod from operation when the fuel condition index is greater than the index constant.
patents
