Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: 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.

Abstract: A method for protecting the components of the primary system of a boiling water reactor, especially against stress cracking corrosion, includes feeding a reducing agent into the primary coolant in order to reduce the number of substances having a oxidizing effect or in order to modify the electrochemical potential of the component surfaces covered with an oxide layer to negative values. An alcohol that can be oxidized in the conditions of a reactor is fed in as a reducing agent, preferably in a liquid form, into the primary coolant. The component surfaces are provided with a blank coating or a single intrinsic oxide coating.

Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: A method of operating a nuclear reactor is provided. The method includes defining a layer increment of a deposit layer modeling a deposit on a heat transfer surface of the nuclear reactor; periodically updating a thickness of the deposit layer by adding the layer increment to the deposit layer; recalculating properties of the deposit layer after each layer increment is added to the deposit layer; determining a temperature related variable of the heat transfer surface as a function of the recalculated properties of the deposit layer; and altering operation of the nuclear reactor when the temperature related variable of the heat transfer surface reaches a predetermined value. A method of modeling a deposit on a heat transfer surface of a nuclear reactor is also provided.

Abstract: 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.

Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: A concrete storage module (26) is adapted to slideably receive a cylindrical canister assembly (12) therein. Heat dissipation fins (62) and a tubular heat shield (96) are disposed within the module to help dissipate heat emitted from the nuclear fuel assemblies stored in the canister to air flowing through the module. The canister assembly (12) is composed of a basket assembly (70) constructed from multi-layer structural plates disposed in cross-cross or egg carton configuration. A single port tool (106) is provided for draining water from the canister (12) and replacing the drain water with make-up gas. The single port tool is mounted in the cover (100) of the canister and is in fluid flow communication with the interior of the canister.

Abstract: The inventive tool is designed to ensure proper plant operation position of the core shroud head bolt assemblies of a boiling water nuclear reactor. The tool exerts an upward pulling force on stuck retainers to return them to the locking position around the nut. The tool includes a frame, an engagement member, and a collar. The engagement member is lowered toward and end of the bolt until it comes into contact therewith. Further movement of the engagement member causes the frame and collar to move upward, lifting the retainer and extending the spring.

Abstract: A concrete storage module (26) is adapted to slideably receive a cylindrical canister assembly (12) therein. Heat dissipation fins (62) and a tubular heat shield (96) are disposed within the module to help dissipate heat emitted from the nuclear fuel assemblies stored in the canister to air flowing through the module. The canister assembly (12) is composed of a basket assembly (70) constructed from multi-layer structural plates disposed in cross-cross or egg carton configuration. A single port tool (106) is provided for draining water from the canister (12) and replacing the drain water with make-up gas. The single port tool is mounted in the cover (100) of the canister and is in fluid flow communication with the interior of the canister.

Abstract: A method to assess and predict pressurized water stress corrosion cracking in operational nuclear power plants and the effect of adding zinc compounds into a reactor coolant system of the nuclear power plant.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.