WESTINGHOUSE ELECTRIC COMPANY LLC Patent applications |
Patent application number | Title | Published |
20160141845 | ELECTRICAL SYSTEM, AND CONNECTION DEVICE AND METHOD OF POWERING A SWITCHGEAR BUS IN AN ELECTRICAL SYSTEM - A connection device is for an electrical system. The electrical system has a powering apparatus, an electrical switching apparatus, and a switchgear bus. The electrical switching apparatus is coupled to the powering apparatus. The connection device includes: a mounting assembly having a panel and a support wall opposite the panel; an electrical transfer assembly including: a number of interconnect assemblies each having a load interconnect member, the load interconnect member being coupled to the panel and electrically connected to the powering apparatus; and a number of base assemblies each including: a number of stud members each coupled to the support wall. At least one of the number of stud members is electrically connected to the load interconnect member and electrically connected to the switchgear bus. | 05-19-2016 |
20160123310 | Thermal Retracting Actuator - A thermal actuator for a rotating shaft shutdown seal that has a piston with a portion of its axial length enclosed within a chamber shell with a material that expands upon a rise in temperature. The portion of the actual length of the piston within the chamber has at least two different diameters with the larger diameter leading in the direction of travel of the piston. Upon a rise in temperature, expansion of the material surrounding the piston within the chamber creates a force on the piston in the desired direction of travel. Below a preselected temperature the piston is positively locked with a passive release when the preselected temperature is reached. | 05-05-2016 |
20160118150 | Method and Apparatus for the Shielded Relocation of a Nuclear Component - A nuclear component transfer device that incorporates a shielded canister into the mast design of a conventional nuclear refueling machine. A moveable mast telescopes within a stationary mast which is attached to a bridge for lateral positioning. The canister allows for the addition of shielding that is positioned with the movement of the moveable mast without additional motorized components to deploy the shielding during nuclear component movement. The nuclear component is drawn up into the shielded canister as the moveable mast lifts the nuclear component. The nuclear component is then placed into a transfer cart that is also fitted with a shielded canister. The transfer is made without exposing the nuclear components resulting in completely shield movement. | 04-28-2016 |
20160118147 | Passive System for Cooling the Core of a Nuclear Reactor - A system for passively cooling nuclear fuel in a pressurized water reactor during refueling that employs gravity and alignment of valves using battery reserves or fail in a safe position configurations to maintain the water above the reactor core during reactor disassembly and refueling. A large reserve of water is maintained above the elevation of and in fluid communication with the spent fuel pool and is used to remove decay heat from the reactor core after the reaction within the core has been successfully stopped. Decay heat is removed by boiling this large reserve of water, which will enable the plant to maintain a safe shutdown condition without outside support for many days. | 04-28-2016 |
20160111174 | ADVANCED FIRST CORE FUEL ASSEMBLY CONFIGURATION - An advanced initial core fuel configuration is for improving the fuel management efficiency and thus economics for a nuclear reactor. The advanced initial core fuel configuration includes a plurality of fuel assemblies having different average enrichments of uranium 235 and arranging the fuel assemblies in an initial core configuration structured to emulate a known equilibrium reload cycle core at least in terms of spatial reactivity distribution. The resulting average enrichment within the initial core ranges from below about 1.0 percent weight of uranium 235 to about 5.0 percent weight of uranium 235. An advanced lattice design is also disclosed. | 04-21-2016 |
20160099080 | NUCLEAR FUEL ELEMENT CORRUGATED PLENUM HOLDDOWN DEVICE - A fuel rod having a gas plenum holddown device that occupies less volume than a conventional spiral spring and is formed from a bellows-like, resilient tubular member having a hollow interior volume open to the gas plenum and an exterior sheath formed from a plurality of alternating ridges and troughs stacked in tandem and interconnected. | 04-07-2016 |
20160097552 | NUCLEAR REACTOR CONTROL ROOM HABITABILITY SYSTEM - A cooling system for a nuclear plant main Control Room Habitability System that employs a vortex tube interposed between a compressed air supply and the control room. The relatively cold air output from the vortex tube is fed into the control room or the electrical equipment room while the relatively hot air exiting the vortex tube is exhausted to the atmosphere. | 04-07-2016 |
20160093407 | PRESSURIZED WATER REACTOR FUEL ASSEMBLY - An elongated control rod guide thimble for a nuclear reactor having a tube-in-tube dashpot design that has circumferential slots in the dashpot walls that align with spaced openings in the guide thimble sheath. The dashpot tube has an end plug with a threaded opening extending axially therethrough which is captured by a thimble screw that extend through an opening in the bottom nozzle and sandwiches an end plug attached to the guide thimble sheath between the dashpot tube end plug and the bottom nozzle. | 03-31-2016 |
20160049212 | Method for In-Core Instrumentation Withdrawal From the Core of a Pressurized Water Reactor - A method of removing the upper internals assembly from a nuclear reactor pressure vessel for refueling that simultaneously disconnects two or more of the in-core instrument thimble assemblies from the reactor vessel penetrations through which their signal cables extend. The signal cables are connected to the penetrations with an electrical connector that supports the two or more in-core instrumentation thimble assembly signal leads. Before the electrical connector is disconnected, water in the vessel is lowered below the connection so that the process is performed in a dry environment. | 02-18-2016 |
20160040681 | THERMAL RETRACTING ACTUATOR - A thermal actuator for a rotating shaft shutdown seal that has a piston with a portion of its axial length enclosed within a chamber shell with a material that expands upon a rise in temperature. The portion of the actual length of the piston within the chamber has at least two different diameters with the larger diameter leading in the direction of travel of the piston. Upon a rise in temperature, expansion of the material surrounding the piston within the chamber creates a force on the piston in the desired direction of travel. Below a preselected temperature the piston is positively locked with a passive release when the preselected temperature is reached. | 02-11-2016 |
20160035442 | CHEMICAL PROCESS FOR PRIMARY SYSTEM MATERIAL PASSIVATION DURING HOT FUNCTIONAL TESTING OF NUCLEAR POWER PLANTS - The present invention relates to a pre-core hot functional testing (HFT) preconditioning process, which includes the introduction of chemical additives, e.g., zinc, into coolant water that circulates through the primary system of a new nuclear power plant, at various temperatures. The chemical additives contact the primary system surfaces, which results in the formation of a protective zinc-containing oxide film on the fresh surfaces to control corrosion release and deposition during subsequent normal operation of the nuclear power plant. The method includes a series of three chemistry phases to optimize the passivation process: 1) an alkaline-reducing phase, 2) an acid-reducing phase and 3) an acid-oxidizing phase. | 02-04-2016 |
20160033216 | METHOD OF SERVICING MODULAR PLATE AND SHELL HEAT EXCHANGER - A modular plate and shell heat exchanger in which welded pairs of heat transfer plates are placed in the shell in order to transfer heat from a secondary fluid to a primary fluid. The heat transfer plates are removably connected using gaskets to header pipes which are connected to a primary fluid inlet and a primary fluid outlet nozzle. The header pipes are supported by a structure which rests on an internal track which is attached to the shell and facilitates removal of the heat transfer plates. The modular plate and shell heat exchanger has a removable head integral to the shell for removal of the heat transfer plates for inspection and replacement. | 02-04-2016 |
20160019991 | SOURCE OF ELECTRICITY DERIVED FROM A SPENT FUEL CASK - Apparatus for extracting useful electric or mechanical power in significant quantities from the decay heat that is produced within spent nuclear fuel storage casks. The power is used for either powering an active forced air heat removal system for the nuclear fuel storage casks, thereby increasing the thermal capacity of the casks, or for emergency nuclear plant power in the event of a station blackout. Thermoelectric generators or other heat engines are employed using the thermal gradient that exists between the spent nuclear fuel container surface and the environment surrounding the cask's components housing the nuclear fuel to produce the power. | 01-21-2016 |
20160018362 | TOTAL INTEGRATED TUBE ANALYSIS - The invention relates to improved systems and methods for inspecting the tubes of a steam generator of a nuclear reactor that involves modeling the steam generator, comparing signals of a tube from an eddy current sensor with aspects of the model to determine whether further analysis is required, employing primary and secondary analysis processes, and producing a combined report of the primary and secondary analyses results. | 01-21-2016 |
20150357055 | CONTROL ROD GUIDE TUBE WITH AN EXTENDED INTERMEDIATE GUIDE ASSEMBLY - A nuclear reactor having an upper internals control rod assembly guide tube formed from upper and lower sections that are connected along a central axial region of the guide tube at an intermediate coupling. An extended control rod axial support is provided for at least some of the control rods over a finite distance within at least one of the interiors of the lower guide tube section or the upper guide tube section. | 12-10-2015 |
20150357053 | CRUSH RESISTANT NUCLEAR FUEL ASSEMBLY SUPPORT GRID - A spacer grid design for a nuclear fuel assemblies that exhibits increased crush strength. The walls of the grid straps that surround the fuel elements have a number of dimples and/or springs with the flat surfaces of those walls formed with a plurality of emboss geometries that are formed in a symmetrical pattern with the pattern covering substantially an entire area of the wall except for the contact surfaces of the dimples and springs that interface with the fuel rods. | 12-10-2015 |
20150348652 | DEPOSITION OF A PROTECTIVE COATING INCLUDING METAL-CONTAINING AND CHROMIUM-CONTAINING LAYERS ON ZIRCONIUM ALLOY FOR NUCLEAR POWER APPLICATIONS - The invention relates to compositions and methods for coating a zirconium alloy cladding of a fuel element for a nuclear water reactor. The coating includes a first tier or layer and a second tier or layer. The first layer includes an elemental metal and the second layer is an oxidation-resistant layer that includes elemental chromium. The first layer serves as an intermediate layer between the zirconium alloy substrate and the second layer. This intermediate layer can be effective to improve adhesion of the second layer to the zirconium alloy substrate. The multilayer coating forms a protective layer which provides improved capability for the zirconium alloy cladding to withstand normal and accident conditions to which it is exposed in the nuclear reactor. | 12-03-2015 |
20150337425 | ZIRCONIUM ALLOYS WITH IMPROVED CORROSION/CREEP RESISTANCE DUE TO FINAL HEAT TREATMENTS - Articles, such as tubing or strips, which have excellent corrosion resistance to water or steam at elevated temperatures, are produced from alloys having 0.2 to 1.5 weight percent niobium, 0.01 to 0.6 weight percent iron, and optionally additional alloy elements selected from the group consisting of tin, chromium, copper, vanadium, and nickel with the balance at least 97 weight percent zirconium, including impurities, where a necessary final heat treatment includes one of i) a SRA or PRXA (15-20% RXA) final heat treatment, or ii) a PRXA (80-95% RXA) or RXA final heat treatment. | 11-26-2015 |
20150310940 | NUCLEAR FUEL ELEMENT - A top end plug design for a nuclear fuel rod or control rod that maximizes the fuel rod length and internal volume for high burn-up, but limits plenum spring melting for eutectic formation margin. The press fit length of the top end plug is increased to increase the distance from the center of heat from the TIG welding process that seals the end plug to the cladding, to the back face of the end plug. A hole in the back of the end plug is enlarged to recover the volume loss from the press fit length increase. | 10-29-2015 |
20150307976 | ZIRCONIUM ALLOYS WITH IMPROVED CORROSION/CREEP RESISTANCE DUE TO FINAL HEAT TREATMENTS - The invention relates to zirconium-based alloys and articles produced therefrom, such as tubing or strips, which have at least one of excellent corrosion resistance to water or steam and creep resistance at elevated temperatures in a nuclear reactor. The alloys include from about 0.2 to 1.5 weight percent niobium, from about 0.01 to 0.6 weight percent iron, from about 0.0 to 0.8 weight percent tin, from about 0.0 to 0.5 weight percent chromium, from about 0.0 to 0.3 weight percent copper, from about 0,0 to 0.3 weight percent vanadium, and from about 0.0 to 0.1 weight percent nickel with the balance at least 97 weight percent zirconium, including impurities. Further, the articles are formed by processes that include final heat treatment of (i) SRA or PRXA (0-33% RXA), or (ii) RXA or PRXA (80-100% RXA). | 10-29-2015 |
20150287484 | METHOD FOR REFUELING A NUCLEAR REACTOR HAVING AN INSTRUMENTATION PENETRATION FLANGE - A method for retracting in-core instrument thimble tubes from the reactor core prior to refueling a nuclear reactor with top mounted instrumentation. The apparatus includes a penetration flange interposed between the head flange and the reactor vessel flange through which the instrumentation cabling passes. The penetration flange is connected to the upper internals and is raised relative thereto to retract instrumentation thimbles from the core prior to removal of the upper internals from the reactor vessel for refueling. The penetration flange is removed from the vessel with the upper internals. | 10-08-2015 |
20150270021 | Upper Internals Arrangement for a Pressurized Water Reactor - A telescoping guide for extraction and reinsertion support handling of in-core instrument thimble assemblies in the area above the upper support plate in the upper internals of a pressurized water reactor. The telescoping guides extend between the upper ends of the upper internals support columns and an axially movable instrumentation grid assembly which is operable to simultaneously raise the telescoping guides and extract the in-core instrument thimble assemblies from the reactor fuel assemblies. | 09-24-2015 |
20150233869 | ULTRASONIC PHASED ARRAY TRANSDUCER FOR THE NDE INSPECTION OF THE JET PUMP RISER WELDS AND WELDED ATTACHMENTS - An ultrasonic phased array transducer assembly having a single housing in which a plurality of phased array transducer subassemblies are mounted at a skewed angle relative to a leading face of the housing and to each other, with each transducer mounted on composite wedge(s) at different orientations within the housing. | 08-20-2015 |
20150221401 | APPARATUS AND METHOD TO REMOTELY INSPECT PIPING AND PIPING ATTACHMENT WELDS - An apparatus and method to remotely perform automated piping and piping attachment weld inspections. The apparatus has two spaced positioning arms that rotate out from one side of a frame structure and a kicker arm that rotates out from an opposite side of the frame structure at a location between the two positioning arms. The positioning arms and the kicker arm wedge the frame structure between an object to be scanned and an opposing structure. A scanning subassembly supported on the frame structure is configured to pivot and move in an appropriate direction and to pilot a transducer around the surface of the object to be scanned. | 08-06-2015 |
20150206612 | SOLID STATE ELECTRICAL GENERATOR - A solid state electrical generator that is responsive to a relatively low level radiation field to power emergency equipment in a nuclear powered generating facility. The electricity is generated from materials, that are not initially radioactive, that are able to produce electrical power when placed inside a relatively low neutron and/or gamma radiation field and will essentially breed material to enhance the power produced by the device sufficiently to allow the device to provide sufficient power to the emergency equipment, even though the reactor or other source of neutron and/or gamma radiation has shut down. | 07-23-2015 |
20150187448 | MOBILE BORATION SYSTEM - A mobile boration system ( | 07-02-2015 |
20150176837 | STEAM GENERATOR SLUDGE LANCE APPARATUS - A sludge lance for a tube and shell steam generator that has a central divider plate that extends substantially the length of a central tube lane substantially bisecting a hand hole through which the tube lane can be accessed. The sludge lance has a nozzle with a spring biased, reciprocally movable plunger that extends against the divider plate and is locked in position by a stream of high pressure cleaning fluid that traverses the nozzle and exits through jets to clean sludge from between the tubes. An alignment tool with a swing arm indexes the jets to assure they are properly aligned with the tube rows and spaced from the divider plate. | 06-25-2015 |
20150125271 | PUMP SEAL WITH THERMAL RETRACTING ACTUATOR - A thermal actuator for a rotating shaft shutdown seal that has a piston with a portion of its axial length enclosed within a cylinder shell with a material, such as a fusible link that changes state or deforms above a given temperature, interposed between a closed end of the cylinder and one end of the piston. The piston is spring biased toward the material and moves toward the closed end of the cylinder when the given temperature is reached and the deformation or change of state of the material makes room for the piston to move toward the closed end. Movement of the piston is transferred through a piston rod to activate the seal. | 05-07-2015 |
20150117587 | AMBIENT TEMPERATURE DECONTAMINATION OF NUCLEAR POWER PLANT COMPONENT SURFACES CONTAINING RADIONUCLIDES IN A METAL OXIDE - This invention generally concerns radioactive decontamination of deposits on components in a nuclear power plant and is specifically concerned with improved systems and methods for disrupting, dissolving, removing and reducing at ambient temperature radionuclides formed on the primary side surfaces of components in a pressurized water reactor and the internal components of a boiling water reactor. The methods include identifying the structure, taking the structure out of operational service, contacting the structure with an aqueous solution (e.g., a recirculating flow or static immersion), and adding an effective amount of elemental metal in solid form to the aqueous solution. | 04-30-2015 |
20150114845 | TARGETED HEAT EXCHANGER DEPOSIT REMOVAL BY COMBINED DISSOLUTION AND MECHANICAL REMOVAL - This invention relates to compositions and methods for the at least partial dissolution, disruption and/or removal of deposit, such as scale and other deposit, from heat exchanger components. The heat exchanger components can include pressurized water reactor steam generators. In accordance with the invention, elemental metal is added locally to the surface of the deposit and/or anodic or cathodic current is applied locally to the deposit surface to destabilize or weaken the deposit. Subsequently, mechanical stress is applied to the weakened deposit to disrupt and remove the deposit from the surface of the heat exchanger component. | 04-30-2015 |
20150110235 | METHOD FOR MONITORING BORON DILUTION DURING A REACTOR OUTAGE - A method for monitoring changes in the boron concentration in the coolant of a reactor during a nuclear plant outage that applies temperature compensation to the source range detector output. The method then monitors the compensated output signal to identify changes in the detector count rate above a preselected value. | 04-23-2015 |
20150098546 | HIGH TEMPERATURE STRENGTH, CORROSION RESISTANT, ACCIDENT TOLERANT NUCLEAR FUEL ASSEMBLY GRID - The invention pertains to a nuclear fuel assembly grid or a portion or a part of the grid, such as a grid strap and/or an integral flow mixer that is at least partially constructed of a composition containing one or more ternary compounds of the general formula I: | 04-09-2015 |
20150098545 | Deposition of Integrated Protective Material Into Zirconium Cladding for Nuclear Reactors by High-Velocity Thermal Application - A zirconium alloy nuclear reactor cylindrical cladding has an inner Zr substrate surface ( | 04-09-2015 |
20150083365 | STEAM GENERATOR AND METHOD OF SECURING TUBES WITHIN A STEAM GENERATOR AGAINST VIBRATION - A steam generator includes a tube bundle having a plurality of tubes, arranged in rows and columns. The first column of tubes includes a first tube having a curved center line disposed in a first plane. The second column of tubes includes a second tube having a curved center line disposed in a second plane, the second plane being parallel to and spaced a distance from the first plane. The steam generator further includes a first number of solid anti-vibration bars disposed between the first column of tubes and the second column of tubes; wherein each of the tubes has a tube outer diameter; and wherein each of the first number of anti-vibration bars has a thickness generally transverse to the first and second planes, the thickness being greater than the distance between the first and second planes minus the tube outer diameter. | 03-26-2015 |
20150078505 | SIC MATRIX FUEL CLADDING TUBE WITH SPARK PLASMA SINTERED END PLUGS - A method of providing an end-capped tubular ceramic composite for containing nuclear fuel ( | 03-19-2015 |
20150055742 | Ion Chamber Radiation Detector - An in-core nuclear detector for detecting the neutron population surrounding the detector. The detector is an ion chamber having a cylindrical outer electrode that is insulated from a central electrode and capped to contain an Argon gas. An electron radiator that produces prompt neutron capture gamma radiation that is substantially, directly proportional to the local neutron population is disposed between the outer tubular electrode and the central electrode. | 02-26-2015 |
20140376681 | NUCLEAR FUEL ASSEMBLY HAVING A SPACER GRID WITH ONE OR MORE SEAMLESS CORNERS - A nuclear fuel assembly grid that has fuel rod support features that take up a substantial portion of the width of the corner fuel rod support cells. The nuclear fuel assembly grid has an outer strap that is joined around a corner of the grid to another outer strap segment and a mating inner strap end at the intersection with the inner strap. The juncture accommodates the width of the rod support feature, enables grid-to-grid anti-snag capabilities and facilitates the use of longitudinal feed materials and dies. | 12-25-2014 |
20140362965 | THERMO-ACOUSTIC NUCLEAR POWER DISTRIBUTION MEASUREMENT ASSEMBLY - A nuclear power distribution measurement assembly that is sized to fit within an instrumentation thimble of a nuclear fuel assembly, that employs a spaced tandem arrangement of thermo-acoustic engines, each of which has a heat source side that is insulated from the reactor coolant traversing the nuclear core in which the fuel assembly is to be placed and a cold side housing a resonator chamber with enhanced thermal conductance to the coolant. The resonator chamber of each of the thermo-acoustic engines is of a different length to generate a different frequency whose amplitude is proportional to the neutron activity at the axial and radial position of the thermo-acoustic engine. The frequency identifies the measurement assembly's position. Acoustic telemetry is employed to monitor the acoustic waves generated by the individual thermo-acoustic engines to provide a remote reading of the axial and radial power distribution of a reactor core. | 12-11-2014 |
20140360443 | METHOD AND APPARATUS FOR MANIPULATING EQUIPMENT INSIDE A STEAM GENERATOR - A method and apparatus for manipulating a tool within the secondary side of a steam generator having a tube sheet with a tube bundle having a plurality of heat exchange tubes extending from the tube sheet in rows with an annulus extending around the heat exchange tubes on a periphery of the tube bundle, between the tubes and a wrapper which surrounds the tube bundle. A robot is introduced into the annulus and extends a probe with a tool across selected lanes between the rows of tubes. A method and apparatus is also disclosed for cleaning sludge from the top of a tube sheet that includes introducing a moveable suction apparatus having attached vacuum inlets into either the no tube lane or the circumferential annulus and sludge vacuuming the top of the tube sheet. | 12-11-2014 |
20140332178 | METHOD AND APPARATUS FOR DELIVERING A TOOL TO THE INTERIOR OF A HEAT EXCHANGE TUBE - A delivery system for remotely driving an eddy current probe through the tubing of a heat exchanger. The system uses a flexible shaft and air pressure to move an inspection probe through the heat exchanger tubes. The flexible shaft initially drives the probe through a sealed conduit to deliver the probe to the tube end at which point a seal on the shaft near the probe head contacts the tube inner surface allowing a buildup of air pressure behind the seal, thus driving the probe through the tube. | 11-13-2014 |
20140321592 | SELF POWERED NUCLEAR DETECTOR - A self-powered neutron detector having an emitter with a slightly negative bias voltage that assures that an increase in the electrons that enter the insulator are counted and decreases or eliminates the gamma induced prompt signal. Variations in the size of the bias is used as a diagnostic tool to estimate the gamma induced prompt signal. | 10-30-2014 |
20140321591 | THERMO-ACOUSTIC NUCLEAR POWER DISTRIBUTION MEASUREMENT ASSEMBLY - A nuclear power distribution measurement assembly that is sized to fit within an instrumentation thimble of a nuclear fuel assembly, that employs a spaced tandem arrangement of thermo-acoustic engines, each of which has a heat source side that is insulated from the reactor coolant traversing the nuclear core in which the fuel assembly is to be placed and a cold side housing a resonator chamber with enhanced thermal conductance to the coolant. The resonator chamber of each of the thermo-acoustic engines is of a different length to generate a different frequency whose amplitude is proportional to the neutron activity at the axial and radial position of the thermo-acoustic engine. The frequency identifies the measurement assembly's position. Acoustic telemetry is employed to monitor the acoustic waves generated by the individual thermo-acoustic engines to provide a remote reading of the axial and radial power distribution of a reactor core. | 10-30-2014 |
20140307843 | REACTOR IN-CORE INSTRUMENT HANDLING SYSTEM - A reactor in-core instrument handling system in which the signal leads are routed from the instrument sensors through an outer sheath through the upper reactor internals and out of and around the sheath in a substantially tightly wound spiral before exiting the reactor vessel. | 10-16-2014 |
20140278293 | METHODOLOGY FOR THE ANALYSIS OF MSLB AND TSV ACOUSTIC TRANSIENTS IN BWRS - This invention relates to a new methodology to analyze the effects of the acoustic waves generated during accident or operational transients occurring in boiling water reactors (BWRs). These transients include the main steam line break (MSLB) event and the turbine stop valve (TSV) operational transient. Accordingly, the invention is utilized for calculating the dynamic loads on steam dryers of a boiling water reactor resulting from the main steam line break event or the turbine stop valve event. | 09-18-2014 |
20140271294 | PUMP SEAL WITH THERMAL RETRACTING ACTUATOR - A thermal actuator for a rotating shaft shutdown seal that has a piston with a portion of its axial length enclosed within a chamber shell with a material that expands upon a rise in temperature. The portion of the actual length of the piston within the chamber has at least two different diameters with the larger diameter leading in the direction of travel of the piston. Upon a rise in temperature, expansion of the material surrounding the piston within the chamber creates a force on the piston in the desired direction of travel. | 09-18-2014 |
20140271288 | PUMP SEAL WITH THERMAL RETRACTING ACTUATOR - A thermal actuator for a rotating shaft shutdown seal that has a piston with a portion of its axial length enclosed within a chamber shell with a material that expands upon a rise in temperature. The portion of the actual length of the piston within the chamber has at least two different diameters with the larger diameter leading in the direction of travel of the piston. Upon a rise in temperature, expansion of the material surrounding the piston within the chamber creates a force on the piston in the desired direction of travel. Below a preselected temperature the piston is positively locked with a passive release when the preselected temperature is reached. | 09-18-2014 |
20140270047 | RIB-TYPE ROUGHNESS DESIGN FOR IMPROVED HEAT TRANSFER IN PWR ROD BUNDLES - The invention pertains to a nuclear fuel assembly in a nuclear water reactor. The fuel assembly includes an array of a plurality of axially extending elongated tubular nuclear fuel elements having first and second closed ends and encapsulating a fissionable material axially along at least a portion of an interior volume thereof and an exterior of at least one of the fuel elements. The fuel elements include a cladding that extends substantially axially between the first and second closed ends. The exterior surface of the cladding is modified to include a surface texture varying axially in a prescribed pattern along at least a portion of an axial length of the cladding. The surface texture includes a plurality of ribs placed parallel to one another and circumferentially around one or more of the fuel elements. Each of the plurality of ribs has a specified height, width, and pitch between each of the plurality of ribs. | 09-18-2014 |
20140270042 | SOURCE OF ELECTRICITY DERIVED FROM A SPENT FUEL CASK - Apparatus for extracting useful electric or mechanical power in significant quantities from the decay heat that is produced within spent nuclear fuel casks. The power is used for either powering an active forced air heat removal system for the nuclear casks, thereby increasing the thermal capacity of the casks, or for emergency nuclear plant power in the event of a station blackout. Thermoelectric generators or other heat engines are employed using the thermal gradient that exists between the spent nuclear fuel and the environment surrounding the cask's components housing the nuclear fuel to produce the power. | 09-18-2014 |
20140270040 | SYSTEMS AND METHODS FOR SPENT FUEL POOL SUBCRITICALITY MEASUREMENT AND MONITORING - A system and method for measuring and monitoring axial flux to determine subcriticality in a spent fuel pool of a nuclear power plant. In certain embodiments of this invention, one or more neutron detectors are operable to generate signals resulting from neutron interactions in the spent fuel pool, a counting device counts the signals which are generated by the one or more neutron detectors, a connecting means electrically connects the one or more neutron detectors to the counting device, a signal analyzer is used to determine reactivity of the fuel assemblies in the spent fuel pool based on the counted signals, a power supply provides power for the neutron detectors, the counting device and the system analyzer, and a software code containing an axial flux curve index is used to correlate the counted signals to determine the subcriticality of the spent fuel pool. | 09-18-2014 |
20140270039 | NUCLEAR RADIATION DOSIMETER USING STRESS INDUCED BIREFRINGENCE CHANGES IN FIBER OPTIC CABLES - The present invention relates to devices and methods for measuring neutron fluence at a pre-selected location which is positioned in a nuclear power plant. The devices and methods include passing neutrons through a fiber optic cable. The fiber optic cable has disposed therein a neutron sensitive material which is capable of absorbing the neutrons to produce a gas. The gas results in a build-up of pressure in the fiber optic cable which causes a change in the optical stress birefringence pattern. This change is measured and used to determine the amount of gas in the fiber optic cable, the number of neutrons absorbed by the neutron sensitive material and subsequently, the neutron fluence at the pre-selected location. | 09-18-2014 |
20140270038 | APPARATUS AND METHOD TO INSPECT NUCLEAR REACTOR COMPONENTS IN THE CORE ANNULUS, CORE SPRAY AND FEEDWATER SPARGER REGIONS IN A NUCLEAR REACTOR - This invention generally concerns robotic systems and is specifically concerned with an improved apparatus and method for inspecting nuclear reactor components in limited access areas, such as, the core annulus, core spray and feedwater sparger regions of a nuclear reactor. This invention includes an apparatus for remotely operating and positioning at least one inspection device for inspecting at least one component in an annulus region of a reactor pressure vessel of a nuclear power plant. The apparatus includes a track, a braking system and a frame assembly which has a frame movably connected to the track, at least one mast assembly and at least one mast positioning assembly. The at least one inspection device is attached to the at least one mast assembly. In certain embodiments, the at least one mast assembly includes a mast that is capable of becoming rigidly stable in both an extended tube form and a retracted rolled form. | 09-18-2014 |
20140261246 | LOCALIZED VACUUM REMOVAL OF STEAM GENERATOR DEPOSITS - A method of cleaning sludge from the top of tube sheet ( | 09-18-2014 |
20140260631 | ACCESS HOLE COVER ULTRASONIC INSPECTION TOOLING - This invention relates generally to ultrasonic inspection of welds and more particularly, to apparatus and methods for ultrasonic inspection of welds on access hole covers found in boiling water reactors having jet pumps. The apparatus includes a base, a center frame coupled to the base and projecting vertically relative to the access hole cover, a radial arm structured to rotate on the center frame and having attached thereto a first pneumatic linear thruster and a second pneumatic linear thruster, and a skew motor assembly and transducer are attached to the first pneumatic linear thruster for scanning the access hole cover weld. The skew motor assembly is structured to control the angle of the transducer and the first pneumatic linear thruster is structured to raise and lower the transducer. | 09-18-2014 |
20140260628 | ULTRASONIC EXAMINATION OF COMPONENTS WITH UNKNOWN SURFACE GEOMETRIES - A method of nondestructively providing a volumetric examination of a bolt through a recess in the bolt head where the recess may have a varying or unknown surface geometry. The method first performs a phased array ultrasonic scan of the bolt's socket surface geometry. The results of the first scan are employed to set the focal laws of a second scan to perform the volumetric examination. | 09-18-2014 |
20140241484 | PRESSURIZED WATER REACTOR DEPRESSURIZATION SYSTEM - A passive cooling system of a pressurized water reactor that relies on a depressurization system to reduce the pressure in the reactor vessel in the event of a loss of coolant accident and vent the steam generated by the decay heat of the reactor core in a post loss of coolant accident stage. The depressurization results in a low pressure difference between the reactor vessel and the containment and enables gravity driven cooling system injection into the reactor vessel. The depressurization system includes a flow restrictor within an orifice in the reactor vessel wall that connects to a vent pipe which forms a flow path between the interior of the reactor vessel and the containment atmosphere when a valve within the vent pipe is in an open position. Preferably, the flow restrictor is a venturi that has a gradual contraction and a gradual expansion in the flow path area. | 08-28-2014 |
20140219411 | ALTERNATE PASSIVE SPENT FUEL POOL COOLING SYSTEMS AND METHODS - The present invention relates to passive cooling systems and methods for cooling a spent fuel pool in a nuclear power plant in the absence of onsite and offsite power, e.g., in a station blackout event. The systems include a gap formed along the periphery of the spent fuel pool, a heat sink, one or more thermal conductive members, a water supply system for delivering water to at least partially fill the gap and conduct heat generated from the spent fuel pool through the gap to at least one thermal conductive member for transporting heat to the heat sink, and a thermal switch mechanism for activating and deactivating the water supply system. | 08-07-2014 |
20140216021 | SYSTEMS AND METHODS FOR GENERATING POWER EMPLOYING VES AIR SUPPLY STORED ENERGY - The present invention relates to a generation system for converting compressed air in a passive main control room habitability system to energy when the main control room habitability system is activated during an accident scenario in a nuclear reactor power plant. The system includes a pressure regulator for reducing the pressure of the compressed air to produce reduced pressurized air, an eductor to deliver air to the control room, and piping to connect the tank to the pressure regulator and the eductor to allow the flow of compressed air therein. The generation system includes a mechanism positioned upstream of the eductor for receiving the reduced pressurized air from the pressure regulator and converting at least a portion of said reduced pressurized air into energy. | 08-07-2014 |
20140205051 | PASSIVE SYSTEM FOR COOLING THE CORE OF A NUCLEAR REACTOR - A system for passively cooling nuclear fuel in a pressurized water reactor during refueling that employs gravity and alignment of valves using battery reserves or fail in a safe position configurations to maintain the water above the reactor core during reactor disassembly and refueling. A large reserve of water is maintained above the elevation of and in fluid communication with the spent fuel pool and is used to remove decay heat from the reactor core after the reaction within the core has been successfully stopped. Decay heat is removed by boiling this large reserve of water, which will enable the plant to maintain a safe shutdown condition without outside support for many days. | 07-24-2014 |
20140205050 | NUCLEAR FUEL ASSEMBLY HANDLING APPARATUS - A fuel assembly handling tool that can be lowered onto the top nozzle of a fuel assembly, positively latch the top nozzle, unlatch from the top nozzle, and be raised off the top nozzle of the fuel assembly. The tool head, that interfaces with the top nozzle has load bearing grippers that latch onto the fuel assembly, that are located in a storage position up within the tool while the tool is lowered onto the fuel assembly. The gripper fingers are then lowered into position during the latching process, and are raised back to the storage position during the unlatching process. In the storage position, the gripping fingers are spaced above the fuel assembly top nozzle when the tool head is resting on the nozzle. | 07-24-2014 |
20140203460 | LASER SINTERING SYSTEMS AND METHODS FOR REMOTE MANUFACTURE OF HIGH DENSITY PELLETS CONTAINING HIGHLY RADIOACTIVE ELEMENTS - The invention relates to remotely operated laser sintering systems and methods for manufacturing pellets containing highly radioactive elements. The highly radioactive elements can be recovered from used nuclear fuels. The systems and methods of the invention employ a feed composition including one or more highly radioactive elements and a ceramic matrix material. The feed composition is distributed in the form of a layer and sintered by directing at least one laser beam to form a pattern in the layer of the feed composition. The pattern corresponds to the shape of the pellet. Further, the sintering process can be repeated as necessary to achieve a pre-determined pellet height. | 07-24-2014 |
20140198890 | APPARATUS AND METHOD FOR REMOVING THE UPPER INTERNALS FROM A NUCLEAR REACTOR PRESSURIZED VESSEL - A lifting fixture for removing the upper internals from a reactor to provide access to the core for refueling that does not require flooding of a refueling canal. The invention provides a means of shielding and ventilation that is integral to a lifting rig used to remove the upper internals. The lifting rig provides a means for personnel to decouple the drive rods from the rod cluster control assemblies so the drive rods can be lifted from the core with the upper internals while shielding maintenance personnel without flooding the area above the reactor. | 07-17-2014 |
20140198889 | METHOD AND APPARATUS FOR THE SHIELDED RELOCATION OF A NUCLEAR COMPONENT - A nuclear component transfer device that incorporates a shielded canister into the mast design of a conventional nuclear refueling machine. A moveable mast telescopes within a stationary mast which is attached to a bridge for lateral positioning. The canister allows for the addition of shielding that is positioned with the movement of the moveable mast without additional motorized components to deploy the shielding during nuclear component movement. The nuclear component is drawn up into the shielded canister as the moveable mast lifts the nuclear component. The nuclear component is then placed into a transfer cart that is also fitted with a shielded canister. The transfer is made without exposing the nuclear components resulting in completely shield movement. | 07-17-2014 |
20140177779 | HEAVY RADIAL NEUTRON REFLECTOR FOR PRESSURIZED WATER REACTORS - A heavy radial neutron reflector for a pressurized water reactor that employs elongated lengths of round bar stock closely packed in either a triangular or rectangular array extending between former plates of a core shroud between the core barrel and the baffle plates which outline the periphery of the reactor core and are formed in axial and circumferential modules. Flow channels are formed in the long gaps between the adjacent round bar stock that communicates cooling water that enters through the core barrel at the top of the shroud and flows down through openings in the former plates to the bottom of the neutron reflector where it exits through a lower baffle orifice to join other cooling water flowing up through the lower core support plate. | 06-26-2014 |
20140161219 | STEAM GENERATOR DUAL HEAD SLUDGE LANCE - A moveable sludge lance ( | 06-12-2014 |
20140140464 | WIRELESS TRANSMISSION OF NUCLEAR INSTRUMENTATION SIGNALS - A system for monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment includes an instrument structured to monitor a condition of the nuclear reactor pressure vessel; a powered wireless transmitting modem disposed in the radioactive environment, the wireless transmitting modem being electrically coupled to the instrument; a receiving modem disposed in the line of sight of the transmitting modem, the receiving modem being in wireless communication with the transmitting modem; and a signal processing unit electrically coupled to the receiving modem, the signal processing unit being structured to determine the condition of the nuclear reactor pressure vessel from the instrument. The transmitting modem is powered by a thermocouple disposed in or on the reactor pressure vessel. | 05-22-2014 |
20140133620 | METHOD OF VALIDATING NUCLEAR REACTOR IN-VESSEL DETECTOR OUTPUT SIGNALS - A method to perform signal validation for either reactor fixed incore detectors and/or core exit thermocouples to enhance core monitoring systems. The method uses a combination of both measured sensor signals and expected signal responses to develop a ratio of measured to expected signals. The ratios are evaluated by determining the expected ratios for each detector based on the behavior of the remaining collection of detectors, taking into account the geometry/location of the other detectors. The method also provides for automatic removal of invalid detectors from the core power distribution determination if sufficient detectors remain on line to adequately characterize the core's power distribution. | 05-15-2014 |
20140126683 | DEPOSITION OF INTEGRATED PROTECTIVE MATERIAL INTO ZIRCONIUM CLADDING FOR NUCLEAR REACTORS BY HIGH-VELOCITY THERMAL APPLICATION - A zirconium alloy nuclear reactor cylindrical cladding has an inner Zr substrate surface ( | 05-08-2014 |
20140123456 | NUCLEAR REACTOR BOTTOM-MOUNTED INSTRUMENTATION NOZZLE REPAIR METHOD - A method for removing and replacing a bottom-mounted instrumentation nozzle on a nuclear reactor pressure vessel. The method (i) caps or plugs the existing bottom-mounted instrumentation nozzle; (ii) cuts the nozzle at or near the nozzle to in-core instrument tube weld; (iii) installs a water-tight sealing enclosure outside the vessel over the bottom of the bottom-mounted instrumentation nozzle creating a water-tight seal with the underside of the reactor vessel; (iv) cuts to sever the existing bottom nozzle from the reactor vessel; (v) extracts the existing nozzle; (vi) installs a replacement alloy 690 nozzle or plug from inside the vessel; and (vii) welds the replacement nozzle or plug in place. The replacement bottom-mounted instrumentation nozzle incorporates an integral shoulder that prevents ejection during operation and facilitates installation, and the entire method is performed while the reactor pressure vessel is filled with water. | 05-08-2014 |
20140109406 | JET PUMP DIFFUSER STACK REPAIR - A method of repairing a slip joint on a jet pump assembly between an inlet mixer and a diffuser, with the diffuser having an opening that receives the inlet mixer with a given spacing between an outside diameter of the inlet mixer and an inside diameter of the opening in the diffuser forming an annulus whose spacing is a product of manufacture and vibration wear. The method comprises the steps of remotely accessing the annulus and narrowing a radial dimension of the annulus. | 04-24-2014 |
20140098924 | APPARATUS AND METHOD TO CONTROL SENSOR POSITION IN LIMITED ACCESS AREAS WITHIN A NUCLEAR REACTOR - This invention concerns robotic systems and is specifically concerned with an improved apparatus and method for remotely positioning a sensor, such as an ultrasonic probe, in limited access areas within a nuclear reactor. The apparatus includes a bottom frame and a top cover which is substantially aligned with and positioned above the bottom frame. A sensor is connected to the top cover and linear rails are connected to the bottom frame in a parallel relationship. There is a mechanism movably connected to the first and second linear rails in order to allow horizontal travel of the top cover. Further, there is at least one cable connected to the sensor and a power source, signal source or receiver. | 04-10-2014 |
20140098923 | APPARATUS AND METHOD TO SWITCH ULTRASONIC SIGNAL PATHS IN A MODERATELY HIGH RADIATION AREA - The invention relates to an apparatus and methods for operation in relatively high radiation fields to remotely switch signal devices through a shared single main umbilical signal cable. The invention is particularly suitable for use in a nuclear reactor, such as a boiling water reactor, and in difficult to access areas in the reactor pressure vessel. One or more main umbilical cables connect a control station to an enclosure housing a signal switching device. The signal switching device allows several signal generating/receiving devices, such as cameras and ultrasonic probes, to be controlled by the one or more main umbilical cables. | 04-10-2014 |
20140098922 | APPARATUS AND METHOD TO INSPECT, MODIFY, OR REPAIR NUCLEAR REACTOR CORE SHROUDS - This invention generally concerns robotic systems and is particularly concerned with improved apparatus and methods for remotely inspecting, modifying or repairing a core shroud in a nuclear reactor. The apparatus of the invention includes a partial upper track which horizontally movable along the core shroud, a head and frame assembly which is horizontally movable along the partial upper track, a lower track which is connected to the head and frame assembly and is horizontally movable along the core shroud, and a carriage and arm assembly which extends downward into an annulus formed by the reactor pressure vessel and the core shroud, wherein the arm includes at least one sensor for inspecting the core shroud. | 04-10-2014 |
20140097834 | SYSTEMS AND METHODS FOR STEAM GENERATOR TUBE ANALYSIS FOR DETECTION OF TUBE DEGRADATION - The systems and methods of the invention pertain to analyzing steam generator tube data for the detection of wear. Further, the invention is capable of performing a comparison of current tube signal data to baseline or historic tube signal data, e.g., from previous and/or the first, in-service inspection of the steam generator. The systems and methods are automated and can generate results to show potential tube-to-tube contact wear areas as well as the progression of tube-to-tube gap reduction within a steam generator tube bundle. In certain embodiments, the invention is capable of comparing current and historical eddy current data to determine the difference that may be related to degradation or other interested phenomena, and of processing and trending historical comparison results to establish normal variance and detect abnormal variances. | 04-10-2014 |
20140091570 | SLIDING DUCT CONNECTION - A duct connection including a first flange configured to couple with a first duct and a slip fit attached to the first flange. The slip fit has a first portion extending from the first flange substantially perpendicular with respect to the first flange and a second portion extending from the first portion of the slip fit substantially parallel with respect to the first flange. The slip fit and the first flange form a receiving area structured such that a portion of a corresponding flange included on a second duct can slide into the receiving area. | 04-03-2014 |
20140069515 | MOBILE BORATION SYSTEM - A mobile boration system ( | 03-13-2014 |
20140054429 | PIPELINE CLAMP FOR VIBRATION MEASUREMENT - A pipeline clamp including a clamp assembly adapted to attach to an outer surface of a pipeline and a protruding member having a first end portion attached to a surface of the clamp assembly and a second end portion extended away from the clamp assembly. The second end portion is configured to accommodate a sensor and the pipeline clamp has a natural frequency equal to or greater than a maximum expected vibration frequency of the pipeline. | 02-27-2014 |
20140029711 | PASSIVE POWER PRODUCTION DURING A NUCLEAR STATION BLACKOUT - Apparatus for passively generating electric power during a nuclear power station blackout by utilizing the temperature difference between the hot inlet of a residual heat removal circuit and the surrounding containment environment. A heat engine, such as a thermoelectric generator, a Sterling Cycle Engine or Rankine Cycle Engine, is coupled in heat exchange relationship with an uninsulated portion of the inlet to a passive residual heat removal heat exchanger and/or passive residual heat removal heat exchanger channel head to generate the power required to operate essential equipment needed to maintain the nuclear power station in a safe condition during a loss of normal onsite and offsite power. | 01-30-2014 |
20140026687 | CONDUIT LENGTH ADJUSTMENT APPARATUS AND METHOD - A conduit length adjustment apparatus having first and second housings each having a member structured to engage the respective housing with a first conduit. A housing movement device is connected between the first and second housings and configured to change a distance between the first and second housings. Selective operation of the members and housing movement device causes the first conduit to move into or out of a second conduit. | 01-30-2014 |
20140014048 | AXIAL FLOW STEAM GENERATOR FEEDWATER DISPERSION APPARATUS - A feedring for use with an axial flow preheat steam generator which utilizes a double wrapper to direct feedwater flow to the cold leg tube bundle region. The feedring is positioned directly over the double wrapper and includes a plurality of standpipes spaced circumferentially along the feedring. The standpipes respectively extend vertically from a lower portion of an interior of the feedring upward through the interior of the feedring. The standpipes have a feedwater intake in the upper portion of the feedring to minimize the potential for vapor formation and bubble collapse water hammer. The components of the standpipe are arranged to minimize the transmission of entrained loose parts from traveling with the feedwater to the tube bundle. A feedwater discharge is provided at the exit of the standpipe at or below the bottom of the feedring, for evenly distributing the feedwater into the double wrapper downcomer. | 01-16-2014 |
20140010340 | FILTER FOR A NUCLEAR REACTOR CONTAINMENT VENTILATION SYSTEM - A wet filter for a nuclear reactor primary containment vent that employs an inclined manifold having a plurality of outlets that communicate through a first set of metal fiber filters submerged in a pool of water enclosed within a pressure vessel. A demister suspended above the pool of water to remove any entrained moisture in the filtered effluent before being passed through a second stage of higher density, dry, metal fiber filters connected to a second manifold that communicates with an outlet on the pressure vessel that is connected to an exhaust passage to the atmosphere. | 01-09-2014 |
20130336442 | PRESSURIZED WATER REACTOR COMPACT STEAM GENERATOR - A steam generator system for a pressurized water reactor which employs an external to containment steam drum and recirculation loop piping. The steam generator system changes the arrangement of a typical pressurized water reactor recirculation steam generator by relocating the functions of steam separation and feedwater preheating outside of the reactor coolant system. The steam generator system and thermal hydraulic conditions are selected in order to minimize the size of the steam generator heat exchanger component volume inside of the containment. The external steam drum component can be isolated in accident conditions when desired and is used as a source of secondary fluid inventory for improved decay heat removal capability and tolerance for loss of feedwater events. Thus, the steam generator component volume inside of the containment is reduced and the amount of maintenance required for the reactor coolant system components are similarly reduced. | 12-19-2013 |
20130336441 | SMALL MODULAR REACTOR SAFETY SYSTEMS - An integral pressurized water reactor that combines all of the components typically associated with a nuclear steam supply system, such as the steam generator, reactor coolant pumps, pressurizer and the reactor, into a single reactor pressure vessel. The reactor pressure vessel is itself enclosed in a containment pressure vessel that also houses a number of safety systems, such as the core make-up tanks, the primary side of residual heat removal heat exchangers, an automatic depressurization system and a recirculation system that enables continuous core cooling through natural circulation over an extended period of time. Actuation of the passive systems is done by single actuation of valves, powered from redundant batteries. | 12-19-2013 |
20130336440 | COMBINED CORE MAKEUP TANK AND HEAT REMOVAL SYSTEM FOR A SMALL MODULAR PRESSURIZED WATER REACTOR - A combined makeup tank and passive residual heat removal system that places a tube and shell heat exchanger within the core makeup tank. An intake to the tube side of the heat exchanger is connected to the hot leg of the reactor core and the outlet of the tube side is connected to the cold leg of the reactor core. The shell side of the heat exchanger is connected to a separate heat sink through a second heat exchanger. | 12-19-2013 |
20130335111 | EDDY CURRENT INSPECTION PROBE - A probe for transporting a nondestructive inspection sensor through a tube, that employs wheels to reduce friction. The radial travel of the wheels are mechanically linked through a cam and axially reciprocal plunger arrangement that centers the probe at tube diameter transitions. Internal wire bending is minimized and a dynamic seal is provided to facilitate an insertion force at the probe and reduce or eliminate compressive load buckling of the flexible cable carried by the probe. Like the wheel arrangement, radial travel of the seal segments are mechanically linked to provide probe centering. | 12-19-2013 |
20130334182 | CORE SHROUD CORNER JOINTS - A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud. | 12-19-2013 |
20130330147 | CAPTIVE SCREW - A captive screw assembly with a knurled end that rotates the captive screw in a hole in which it is mounted as the captive screw is pressed into the hole. The lower end of a ferrule of the captive screw assembly terminates within the hole and is flared against the side wall of the hole. | 12-12-2013 |
20130329847 | NUCLEAR CONTROL ROD WITH FLEXURE JOINT - A rod cluster control assembly having a control rod upper end plug formed in two parts and connected together with a flexible joint coupling located at or below a connecting finger on a spider vane. The upper portion of the upper control rod end plug is screwed into the connecting finger on the spider vane and lock welded between a lower portion of the finger and a shoulder on an abutting ledge of the upper portion of the end plug. | 12-12-2013 |
20130312243 | METHOD OF FASTENING A CAPTIVE SCREW TO A PANEL - A method for securing a captive screw assembly within a hole in a panel cover with a press. The press is centered over a hole in the panel and clamped to the panel. A force arm on the press is lowered to drive the captive screw into the hole and then the press is released from the panel. | 11-28-2013 |
20130308740 | PRESSURIZER SURGE-LINE SEPARATOR FOR INTEGRAL PRESSURIZED WATER REACTORS - An integral pressurized light water reactor having most of the components of a primary side of a pressurized water reactor nuclear steam supply system housed in a single pressure vessel with a pressurizer separated from the remaining reactor system by a surge separator having multiple layers of separated steel plates with a number of concentric baffles extending therebetween. A circuitous flow path is provided through and between the plates and concentric baffles and a relatively stagnant pool of coolant is maintained within an innermost zone between the plates to provide thermal isolation. | 11-21-2013 |
20130299670 | TUBESHEET GRIPPING MECHANISM AND METHOD - A tubesheet anchor for suspending a tool from the underside of a heat exchanger tubesheet that inserts one end of two fingers into corresponding openings in the tubesheet and leverages one finger off the other to apply a frictional force to the sides of the tubesheet openings in which the fingers are inserted to clamp the fingers to the tubesheet. | 11-14-2013 |
20130299122 | TUBESHEET WALKER FOR HEAT EXCHANGER INSPECTIONS - A robotic tubesheet walker having two rails connected by a central hinge, wherein the central hinge can be opened or closed by an actuation device. Upon each rail is mounted a carriage, wherein each carriage can move along its respective rail toward or away from the central hinge by means of a drive mechanism. Each carriage further contains at least two “gripper” attachment mechanisms, such as camlocks, to grip the tubesheet. The grippers either insert into tube holes within the tubesheet to fasten the respective carriage to the tubesheet, or retract to disengage. Further attached to the central hinge is a tool support fixture, and attached to the tool support fixture is a coupler that holds maintenance or inspection tools. | 11-14-2013 |
20130294565 | METHOD OF REFUELING A NUCLEAR REACTOR - A method of refueling a nuclear reactor that includes the steps of removing the reactor vessel head and upper internals to a storage location and installing a cylindrical tank having open upper and lower ends, on the reactor vessel flange. The cylindrical tank is sealed to the reactor vessel and a penetration on the side of the cylindrical tank is sealed to a refueling canal that is connected to a spent fuel pool. The level of reactor coolant within the reactor vessel is then raised to at least partially fill the cylindrical tank to a level equal to that of the spent fuel pool. The refueling canal is then opened and a refueling machine supported on the reactor vessel is employed to transfer fuel assemblies between the core and the spent fuel pool. | 11-07-2013 |
20130287157 | INSTRUMENTATION AND CONTROL PENETRATION FLANGE FOR PRESSURIZED WATER REACTOR - A nuclear reactor having a penetration seal ring interposed between the reactor vessel flange and a mating flange on the reactor vessel head. Radial ports through the flange provide passage into the interior of the reactor vessel for utility conduits that can be used to convey signal cables, power cables or hydraulic lines to the components within the interior of the pressure vessel. A double o-ring seal is provided on both sides of the penetration flange and partial J-welds on the inside diameter of the flange between the flange and the utility conduits secure the pressure boundary. | 10-31-2013 |
20130285068 | SOLID STATE RADIATION DETECTOR WITH ENHANCED GAMMA RADIATION SENSITIVITY - A silicon carbide Schottky diode solid state radiation detector that has an electron donor layer such as platinum placed over and spaced above the Schottky contact to contribute high energy Compton and photoelectrical electrons from the platinum layer to the active region of the detector to enhance charged particle collection from incident gamma radiation. | 10-31-2013 |
20130281341 | ADDITIVES FOR HEAT EXCHANGER DEPOSIT REMOVAL IN A WET LAYUP CONDITION - This invention relates to compositions and methods for the at least partial dissolution, disruption and/or removal of deposits, such as scale and other deposits, from heat exchanger components. The heat exchanger components can include pressurized water reactor steam generators. The pressurized water reactor steam generators can be in a wet layup condition. The compositions include elemental metal and complexing agent selected from the group consisting of sequestering agent, chelating agent, dispersant, and mixtures thereof. The methods include introducing the compositions into the heat exchanger components. | 10-24-2013 |
20130279641 | METHOD TO EXTRACT TRITIUM FROM IRRADIATED BOILING WATER REACTOR CONTROL ROD BLADES - A method for extracting tritium from irradiated boiling water reactor control rods that have cruciform-shaped. Bands of a malleable metal are wrapped around the flat portions of the blades, one band near the top of each blade panel and a second band near the bottom. The bands are crimped and an inlet penetration is formed through one of the bands and the panel and an outlet penetration is formed through the second band and the panel. A termination of each end of a closed loop conduit is sealably connected to the inlet and outlet for transporting a carrier gas through the interior of the panel. The carrier gas passing through the interior transports the tritium out of the panel to a tritium getter filter to capture the tritium. The carrier gas then recirculates through the system. | 10-24-2013 |
20130272482 | PRESSURIZED WATER REACTOR FUEL ASSEMBLY GRID - A nuclear pressurized water reactor fuel assembly grid that has a plurality of spring-like, resilient, cantilevered members that extend, substantially adjacent each other, from a wall of a grid cell that supports the fuel rods, into the support cell, with a distal end of each member being compressed by the fuel rod passing through the cell so as to exert a lateral force on the fuel rod. The plurality of cantilevered members replace conventional dimples employed in fuel assembly grid support cells. | 10-17-2013 |
20130272475 | PASSIVE CONTAINMENT AIR COOLING FOR NUCLEAR POWER PLANTS - A passive containment air cooling system for a nuclear power plant that enhances air flow over a metal containment that houses the reactor system to improve heat transfer out of the containment. The heat transfer is improved by employing swirl vanes to mix the air as it rises over the walls of the containment due to natural circulation and a vortex engine proximate an exit along the cooling air path to increase the quantity of air drawn along the containment. | 10-17-2013 |
20130272474 | PASSIVE CONTAINMENT AIR COOLING FOR NUCLEAR POWER PLANTS - An enhanced passive containment air cooling system for a nuclear power plant that increases the heat transfer surface on the exterior of the nuclear plant's containment vessel. The increased surface area is created by forming a tortuous path in or on at least a substantial part of the exterior surface of the containment vessel over which a cooling fluid can flow and follow the tortuous path. The tortuous path is formed from a series of indentations and protrusions in or on the exterior surface that form a circuitous path for the cooling fluid. | 10-17-2013 |
20130266107 | METHODS FOR PROTECTION OF NUCLEAR REACTORS FROM THERMAL HYDRAULIC/NEUTRONIC CORE INSTABILITY - The invention relates to methods for protecting a nuclear reactor core, such as a boiling water reactor core, from fuel and cladding damage due to thermal hydraulic instability in extended operating power flow conditions and, in particular, when an extended power uprate is implemented. The methods employ existing licensed stability methodologies and incorporated minor changes, e.g., to the Average Power Range Monitor (APRM)-based trip system to preclude operation inside the stability vulnerable region of the power/flow map. The APRM-based trip system is modified to set down the APRM flow-biased scram line when core flow is less than a predetermined core flow to prevent the core from entering an unstable region of operation. | 10-10-2013 |
20130223579 | CONTROL ROD DRIVE MECHANISM ("CRDM") ASSEMBLY FOR A NUCLEAR REACTOR - The present invention relates to a control rod drive mechanism assembly for a nuclear reactor having a nuclear reactor vessel, a nuclear reactor core, a reactor vessel head, a latch housing nozzle, a latch housing, a rod travel housing, a latch assembly, a drive rod assembly and a rod control cluster assembly. The latch housing is integrated with the latch housing nozzle, the rod travel housing is welded to the latch housing, and the latch assembly is connected to the rod travel housing. The latch assembly includes the drive rod assembly and the rod control cluster assembly which is attached to the drive rod assembly. | 08-29-2013 |
20130180337 | SIGNAL PROCESSING OF LAMB WAVE DATA FOR PIPE INSPECTION - The invention relates to systems and methods for conducting an ultrasonic, nondestructive evaluation and inspection of a pipe using Lamb-type wave transducers to detect the presence of defects, flaws, discontinuities and the like. The transducers are positioned within the interior space of the pipe. Two transducers are positioned facing each other such that the ultrasonic beam emitted from each of the transducers is directed toward the other transducer and the portion of pipe to be inspected. The coverage of the transducers is verified and the results are processed using a combination of a cross correlation filter and a signal processing tool. | 07-18-2013 |
20130177127 | OPTIMIZED FLOWER TUBES AND OPTIMIZED ADVANCED GRID CONFIGURATIONS - A support grid for a nuclear fuel assembly, the nuclear fuel assembly including a generally cylindrical fuel rod with a diameter, wherein the support grid includes a frame assembly having a plurality of generally circular cells and a plurality of helical frame members. The helical frame members are disposed in the cells and are structured to contact the cell as well as a fuel rod. The helical fuel rod contact portion may have a variable pitch. | 07-11-2013 |
20130153715 | Support Apparatus for Supporting Dosimetry System in Proximity to Reactor Apparatus - An improved support apparatus is structured to support a dosimetry system within an interior region of a containment apparatus. The dosimetry system is supported in a region between an interior surface of the containment apparatus and an exterior surface of a reactor apparatus that is disposed within the interior of the containment apparatus. The support apparatus includes a retention apparatus having a plurality of brace elements that each have a receptacle formed therein. The receptacles are aligned with one another when installed within the interior of the containment apparatus. The support apparatus further includes a tube apparatus that includes a plurality of tube segments that are connectable together. The dosimetry system is situated within an elongated cavity formed in the tube apparatus. The tube apparatus is situated in the receptacles and thereby supports the dosimetry system on the containment apparatus. | 06-20-2013 |
20130129032 | METHOD TO REDUCE THE VOLUME OF BOILING WATER REACTOR FUEL CHANNELS FOR STORAGE - A method of reducing the volume of a boiling water reactor fuel channel for storage in which the fuel channel is sealed with an outer sleeve that is closed at its upper and lower ends. The sleeve, which is made of a malleable metal is then laterally compacted and sheared into segments suitable for transport and/or storage. | 05-23-2013 |
20130129031 | APPARATUS FOR VERTICALLY SEGMENTING A BOILING WATER REACTOR CONTROL ROD BLADE - Apparatus for longitudinally segmenting a cruciform shaped irradiated boiling water reactor control rod having four elongated blades radially extending from a central spline, into four flat panels. The apparatus employs a double bladed band saw with the band saw blades orthogonally oriented at two different elevations and having one side of each band saw blade crossing over the center of the spline of the control rod in between the control rod blades. | 05-23-2013 |
20130129029 | NUCLEAR FUEL ROD PELLET STACK INSPECTION - A method of detecting defects in nuclear fuel within a fuel rod that first heats the fuel rod to a temperature substantially above the ambient temperature. The surface temperature of the fuel rod cladding is then monitored as the fuel rod is allowed to cool. Variations in the temperature measured over the surface is then noted as an indication of defects. | 05-23-2013 |
20130125363 | Method of Replacing Shroud of Boiling Water Nuclear Reactor, and Associated Apparatus - An improved method of replacing at least a portion of a shroud apparatus of a BWR includes forming a cut in the shroud and removing at least a portion of the shroud apparatus that is adjacent the cut from a remaining portion of the shroud apparatus. The method also includes positioning a replacement shroud system adjacent the remaining portion of the original shroud apparatus and connecting a plurality of removable fastening devices between the new replacement shroud system and the remaining portion of the original shroud apparatus. The replacement shroud system thus can, in the future, be readily replaced by detaching the removable fastening devices, removing the replacement shroud system, and installing a new replacement shroud system with less effort than was required in cutting through the original shroud and replacing it with the replacement shroud system. An improved replacement shroud system is also disclosed. | 05-23-2013 |
20130121454 | SEMI-PORTABLE EMERGENCY COOLING SYSTEM FOR REMOVING DECAY HEAT FROM A NUCLEAR REACTOR - An emergency temporary spent fuel pool cooling system for a nuclear power generating facility that has a permanently installed primary loop within the nuclear containment and a mobile temporary secondary loop. The secondary loop is housed in transport vehicles that can be stored off site and is connectable in heat exchange relationship with the primary loop through quick disconnect couplings that are accessible on the outside of the reactor containment. The transport vehicles also include self-contained power and compressed air sources for powering and controlling the entire emergency cooling system. The system also has a make-up water injection capability for refueling the spent fuel pool and secondary loop. | 05-16-2013 |
20130114780 | NUCLEAR CORE COMPONENT - A fuel rod or control rod for a nuclear reactor that has a spacer interposed between an upper end plug and a plenum spring which extends between the spacer and the fissile or absorber material. Preferably, the spacer is a relatively thin sleeve with a radially extending lip that sits above the coil spring wound at least in part around the sleeve. | 05-09-2013 |
20130101077 | METHOD OF ACHIEVING AUTOMATIC AXIAL POWER DISTRIBUTION CONTROL - A control strategy for a pressurized water nuclear reactor that employs separate, independent control rod banks for respectively controlling T | 04-25-2013 |
20130092106 | ANTI-CLOGGING STEAM GENERATOR TUBE BUNDLE - A tube and shell steam generator having an anti-clogging heat exchange tube bundle wherein the tube support plates within the tube bundle are designed with varying degrees of porosity thereby regulating local secondary side fluid conditions (velocity, quality, superheat, void fraction, etc.) in a manner to reduce the potential for clogging of the tube support plate lobes that are more prone to clogging. | 04-18-2013 |
20130085684 | AUTOMATED ANALYSIS COVERAGE VERIFICATION (AACV) - The invention relates to systems and methods for verifying complete analysis coverage in a steam generator tube inspection. The analysis is conducted by an automated analysis process. The process includes setup and analysis functions. Information is entered into the setup function to identify abnormalities to be inspected and to model the steam generator tubes. The verification includes employing a software verification program to detect and identify a gap within analysis coverage for a particular abnormality or set of abnormalities for which the steam generator tube is inspected, in a particular portion of the steam generator tube. | 04-04-2013 |
20130083883 | POOL LEVEL INDICATION SYSTEM - A liquid level indication system that employs a plurality of heated thermocouples staggered at discrete elevations along a height of a liquid pool, whose outputs are respectively compared to the output of an unheated thermocouple positioned at one of the lower discrete elevations. A significant difference in the outputs of the heated and unheated thermocouples provides an indication of the liquid level. | 04-04-2013 |
20130083879 | IN-CORE INSTRUMENT THIMBLE ASSEMBLY - A self-powered integral in-core instrument thimble assembly for monitoring the temperature and radiation levels surrounding a nuclear fuel assembly, that transmits output signals wirelessly to a remote location. The in-core instrument thimble assembly is activated by a short exposure within a reactor core and remains active after the fuel assembly is removed from the reactor core to continuously provide a remote monitoring capability for the fuel assembly as it is transported or stored at a remote location, without an external power source. | 04-04-2013 |
20130077729 | METHOD OF SEGMENTING AND PACKAGING IRRADIATED COMPONENTS - A method of segmenting and packaging an irradiated hardware component for storage or shipment. Radiological and physical characteristics of the components are first mapped over its surface. A segmenting plan and a loading plan is then determined that sets forth where over the surface of the components lateral segments are to be made from a map obtained from the mapping step taken into consideration any licensing restrictions and requirements of the facility in which the casks is to be stored with the view to maximizing the loading of the casks. The irradiated components are then segmented in accordance with the segmenting plan and loaded into the casks in accordance with the loading plan. | 03-28-2013 |
20130077727 | ROD POSITION DETECTION APPARATUS AND METHOD - An improved apparatus for determining the position of a drive rod within the interior of a drive rod housing includes a transmission antenna at one location on the housing and a receiving antenna at another location on the housing. An electromagnetic excitation signal sent to the transmission antenna is detected, at least in part, by the receiving antenna, and the received signal is processed with a vector network analyzer routine to model the drive rod housing as a wave guide having a filter response. A group delay is detected and is compared with a calibration data set which provides a current position of the drive rod that corresponds with the group delay. | 03-28-2013 |
20130074302 | METHOD OF SEGMENTING IRRADIATED BOILING WATER REACTOR CONTROL ROD BLADES - A method of reducing the volume of a blade section of a boiling water reactor control rod for transport or storage that cuts the control rod spline into four substantially equal longitudinal sections, with each longitudinal section including one control rod blade. Each longitudinal section is radiologically characterized and the locations of desired lateral segmentation are identified. A band of malleable metal is wrapped around each longitudinal section at each of the locations and the bands are sheared to separate segments of the longitudinal section and the ends of the bands are crimped at the point of shearing to seal the interior of the segments. | 03-28-2013 |
20130070890 | GROOVED NUCLEAR FUEL ASSEMBLY COMPONENT INSERT - A nuclear fuel assembly component such as a control rod that has a cylindrical insert such as neutron absorbing material that is closely received within a cladding that is sealed at either end with end caps. The cylindrical member has grooves formed in its side wall extending from an upper surface to a lower surface to permit air to escape from the cladding as the cylindrical member is loaded into the cladding during manufacture. | 03-21-2013 |
20130069602 | SQUIB CONTROL CIRCUIT - An improved control circuit that is structured to energize another device such as a squib. A first portion of the circuit includes a first transistor and is structured to discharge at a first rate a first portion of a charge stored by a capacitor. Another portion of the circuit includes a second transistor and is structured to discharge a second portion of the charge subsequent to the discharge of the first portion of the charge and at a second rate greater than the first rate. | 03-21-2013 |
20130061664 | INSPECTION VEHICLE FOR A TURBINE DISK - An inspection vehicle structured to inspect a portion of the turbine disk, preferably the blade attachment hubs, while the turbine disk is disposed within a turbine housing assembly is provided. A turbine disk is generally planar but includes a inner hub and an outer blade attachment hub. The inner hub is coupled to a shaft and the blade attachment hub provides a surface to which removable blades are attached. The area between the inner hub and outer blade attachment hub is substantially planar. The inner and blade attachment hubs are the “inspection areas” that the inspection vehicle is structured to inspect. The inspection vehicle travels over, and is magnetically coupled to the planar surface between the two hubs. | 03-14-2013 |
20130044851 | BACKUP NUCLEAR REACTOR AUXILIARY POWER USING DECAY HEAT - A nuclear plant auxiliary backup power system that uses decay heat following a plant shutdown to produce electrical power through a dedicated steam turbine/generator set. The decay heat produces a hot operating gaseous fluid which is used as a backup to run an appropriately sized turbine that powers an electrical generator. The turbine is configured to utilize a portion of the existing nuclear plant secondary system and exhausts the turbine exhaust to the ambient atmosphere. The system functions to both remove reactor decay heat and provide electrical power for plant systems to enable an orderly shutdown in the event traditional sources of electric power are unavailable. | 02-21-2013 |
20130032100 | NUCLEAR STEAM GENERATOR STEAM NOZZLE FLOW RESTRICTOR - A steam nozzle flow restrictor for a steam generator that restricts steam exit flow during a steam line break, but has a low pressure drop during normal operation. The flow restrictor has a support web suspended from the steam outlet nozzle with the web having a central opening concentric with the central axis of the nozzle. A shaft is slidably supported within the central opening of the web and has an orifice plate that is suspended within the steam generator at one end of the shaft, spaced from the steam nozzle. The orifice plate closes against the underside of the nozzle upon encountering increased steam exit flow as a result of a steam line break. | 02-07-2013 |
20130028365 | POWER GENERATION FROM DECAY HEAT FOR SPENT NUCLEAR FUEL POOL COOLING AND MONITORING - An auxiliary power source for continuously powering pumps for replenishing water in a spent fuel pool and sensors monitoring the pool, in the event of a station blackout at a nuclear plant. The power source uses waste heat from spent fuel within the pool to activate a thermoelectric module system or a waste heat engine, such as a Stirling cycle or organic Rankine cycle engine to generate power for the pump and sensors. The auxiliary power source can also power a cooling system to cool the spent fuel pool. | 01-31-2013 |
20120307957 | FUEL HANDLING AREA PASSIVE FILTRATION DESIGN - The present invention relates to a passive filtration system for a fuel handling area having a spent fuel pool in a nuclear reactor. The passive filtration system reduces a discharge into the atmosphere of particulates, such as radioactive particulates, generated in a spent fuel pool boiling event. The passive filtration system includes a discharge path, a vent mechanism positioned between the fuel handling area and the discharge path. The vent mechanism is structured to release a steam and air mixture from the fuel handling area to the discharge path. The steam and air mixture includes the particulates. The passive filtration system further includes an air filtration unit located in the discharge path and this unit has at least one passive filter. The steam and air mixture is forced through the at least one passive filter due to a differential pressure generated in the fuel handling area. The at least one passive filter traps particulates from the steam and air mixture to produce a filtered steam and air mixture that is released through a second vent mechanism into the atmosphere. | 12-06-2012 |
20120304466 | ANTI-VIBRATION BAR CLAMPING TOOL - An automated tool and method for spacing a gap between the anti-vibration bars and the flow tubing in the bend region of a tube bundle of a U-tube steam generator, prior to welding, during manufacture. The tool comprises two toggle clamp elements attached to a scissors assembly, which are actuated by a linear drive motor in line with a force gauge, and coupled to a distance measuring device. The tool is operated by first attaching the toggle clamps respectively to two adjacent anti-vibration bars. A linear drive motor on the tool is then cycled and readouts of force and distance are plotted on a curve to determine the point of contact between the anti-vibration bar and the tube. The computer then automatically cycles the drive motor to adjust the gap to a desired range of position coordinates adjacent the point at which the curve markedly changes slope. | 12-06-2012 |
20120294406 | ADVANCED FIRST CORE FUEL ASSEMBLY CONFIGURATION AND METHOD OF IMPLEMENTING THE SAME - An advanced initial core fuel configuration is for improving the fuel management efficiency and thus economics for a nuclear reactor. A method of implementing such an initial core involves providing a plurality of fuel assemblies having different average enrichments of uranium 235 and arranging the fuel assemblies in an initial core configuration structured to emulate a known equilibrium reload cycle core at least in terms of spatial reactivity distribution. The resulting average enrichment within the initial core ranges from below about 1.0 percent weight of uranium 235 to about 5.0 percent weight of uranium 235. An advanced lattice design is also disclosed. | 11-22-2012 |
20120288050 | PROCESS FOR ADDING AN ORGANIC COMPOUND TO COOLANT WATER IN A PRESSURIZED WATER REACTOR - The present invention relates generally to a process for a pressurized water reactor. The pressurized water reactor includes a primary circuit and a reactor core. The process includes adding a sufficient amount of an organic compound to coolant water passing through the primary circuit of the pressurized water reactor. The organic compound includes elements of carbon and hydrogen for producing elemental carbon. | 11-15-2012 |
20120284995 | CAPTIVE SCREW DEVICE AND METHOD - A press for securing a captive screw assembly within a hole in a panel cover. The press is supported from the panel cover as it is operated and includes a tool for extracting the captive screw assembly from the panel cover hole when the captive screw assembly is no longer functional. The extraction tool engages a bottom end of the captive screw and severs the flare on the ferrule of the captive screw assembly that anchors the assembly to the hole in the panel cover. A captive screw assembly is also described that has a knurled end that anchors the captive screw assembly to a hole in the panel cover. | 11-15-2012 |
20120269313 | STEAM GENERATOR DUAL HEAD SLUDGE LANCE - A moveable sludge lance ( | 10-25-2012 |
20120269310 | DYNAMIC PORT FOR MEASURING REACTOR COOLANT PUMP BEARING OIL LEVEL - A dynamic port that extends from the bottom wall of an oil reservoir that surrounds the lower guide bearing of a reactor coolant pump and is in fluid communication within an oil level gauge. The dynamic port is rotatable into and out of the oil flow path to adjust the dynamic oil level shown by the oil level gauge when the pump is at operating speed to be substantially equal to the static oil level when the motor is at rest. | 10-25-2012 |
20120263271 | NUCLEAR FUEL - A nuclear fuel pellet design that is a cylindrical axial profile with either a larger radius or conical shaped ends such that the as built diameter at the ends of the pellet are slightly smaller than at the middle and at normal operating conditions, the diameter at the ends is nearly the same as at the middle. Preferably, there are short chamfers at the axial ends of the pellet. | 10-18-2012 |
20120257706 | REACTOR VESSEL INTERNALS RADIATION ANALYSES - The present invention relates to a computational system and method to analyze a radiation field in a light water nuclear reactor. The system and method include a radiation transport module for performing a neutron and gamma transport calculation of a reactor geometry model of the light water nuclear reactor for at least one fuel cycle; a material analysis module for analyzing flux data calculated by the radiation transport module for materials and/or components within the light water nuclear reactor; and a mobile component tracking module for calculating radiation exposure for components within the light water nuclear reactor which have more than one different location during or between the at least one fuel cycle. | 10-11-2012 |
20120257705 | METHOD OF DETECTING AN EXISTENCE OF A LOOSE PART IN A STEAM GENERATOR OF A NUCLEAR POWER PLANT - A plurality of signal anomalies are identified in a number of tubes in a steam generator. Since the geometry of the steam generator is known, the location of each signal anomaly along each tube is converted into a location within the interior of the steam generator. If a plurality of signal anomalies are at locations within the steam generator that are within a predetermined proximity of one another, such a spatial confluence of signal anomalies is determined to correspond with a loose part situated within the steam generator. Additional methodologies can be employed to confirm the existence of the loose part. Historic tube sheet transition signal data can be retrieved and subtracted from present signals in order to enable the system to ignore the relatively strong eddy current sensor signal of a tube sheet which would mask the relatively weak signal from a loose part at the tube sheet transition. | 10-11-2012 |
20120250814 | NUCLEAR FUEL ASSEMBLY SUPPORT GRID - A spacer grid for a nuclear fuel assembly that exhibits increased crush strength. Each grid strap at the ligaments that support fuel rods has a spring or dimple to support the fuel rods under anticipated external loads during shipping and handling or in a seismic event. One or more elongated embossed ribs are provided on each of the fuel rod grid strap support ligaments to increase its moment of inertia by forming various shapes on the ligaments of the grid strap. Preferably, the ribs have a streamlined shape to prevent any excessive pressure drop. In this manner, the crush strength of a conventional short grid strap is increased without meaningful additional manufacturing costs or adverse effects to the neutron economy of the grid. | 10-04-2012 |
20120250813 | SELF-CONTAINED EMERGENCY SPENT NUCLEAR FUEL POOL COOLING SYSTEM - An auxiliary system for cooling a spent nuclear fuel pool through a submersible heat exchanger to be located within the pool. In each train or installation, a single loop or series of loops of cooling fluid (e.g., sea water or service water) is circulated. The system is modular, readily and easily installed during an emergency and can be self operating with its own power source. Multiple trains may be used in parallel in order to accomplish the required degree of spent fuel pool cooling required. | 10-04-2012 |
20120247727 | STEAM GENERATOR TUBE LANE FLOW BUFFER - A tube and shell steam generator in which a series of rods having a diameter substantially equal to that of the heat exchange tubing in the tube bundle are placed on either side of the tube lane to buffer the flow in the tube lane from the heat exchange tubes to attenuate turbulent forces on the first several rows of heat exchange tubes adjacent to the tube lane. | 10-04-2012 |
20120236978 | TOOL FOR DELIVERY OF TESTING ELEMENT TO A LIMITED ACCESS LOCATION WITHIN A NUCLEAR CONTAINMENT - An improved tool for delivery of a testing element to a limited access location within a nuclear containment includes a rotation apparatus having a connection element that is configured to have an aperture that is formed generally centrally therein. The aperture is structured to receive therein a cable that extends from a testing element, which that avoids the need to use slip rings or similar devices to permit the testing element to carried on a pivotable and rotatable structure. Moreover, an improved submersible machine is structured to carry the tool to a limited access location within a nuclear containment. | 09-20-2012 |
20120235675 | INSPECTION MODE SWITCHING CIRCUIT - An eddy current probe testing apparatus structured to operate concurrently in a driver pick-up mode and said impedance mode is provided. The eddy current probe has two coils. The eddy current probe testing apparatus also includes a signal producing device, an output device, and a switch assembly. The switch assembly is structured to switch how an input signal from the signal producing device is provided to the two coils. Thus, an inspection may be performed in two modes concurrently. | 09-20-2012 |
20120230459 | METHOD OF IMPROVING WEAR AND CORROSION RESISTANCE OF ROD CONTROL CLUSTER ASSEMBLIES - The present invention relates to tubular elements, such as fuel assembly tubes, which are designed to be used in high pressure and high temperature water in nuclear reactors, such as pressurized water nuclear reactors. In particular, the present invention relates to a method of improving wear resistance and corrosion resistance by depositing a protective coating having a depth of from about 5 to about 25 μm on the surface of the tubular elements. The coating is provided by nitriding the tubular element at a temperature of from about 400° C. to about 440° C. The nitridation of the tubular element can be carried out for a duration of from about 12 hours to about 40 hours. | 09-13-2012 |
20120224663 | NUCLEAR STEAM GENERATOR SUPPORT AND ALIGNMENT STRUCTURE - A nuclear steam generator support and alignment system that supports the entire weight of a nuclear steam generator on the walls of the shielding compartment in which the steam generator is designed to operate within. The support includes hydraulic positioners that can raise, lower, rotate and tilt the steam generator to align the steam generator with reactor coolant piping to which it is to be connected. | 09-06-2012 |
20120195402 | NEUTRON SOURCE ASSEMBLY - A neutron source rodlet assembly having a separate source capsule assembly that is not encapsulated within the neutron source rodlet assembly. The neutron source rodlet assembly is made up, at least in part, of a neutron source positioning rodlet assembly and the source capsule assembly configured such that assembly together is feasible at a remote site and they can be shipped separately. The source capsule assembly has outer and inner capsules with the outer capsule having a threaded stud at one end that mates with a complimentary threaded recess on the neutron source positioning rodlet assembly. The inner capsule contains a neutron source. The neutron source positioning rodlet assembly and the source capsule assembly are locked together at their interface when the threaded joint is completely tightened. A secondary neutron source material may also be encapsulated within a hollow portion of the neutron source positioning rodlet assembly. | 08-02-2012 |
20120185222 | FULL SPECTRUM LOCA EVALUATION MODEL AND ANALYSIS METHODOLOGY - This invention relates to a computational system and method for performing a safety analysis of a postulated Loss of Coolant Accident in a nuclear reactor for a full spectrum of break sizes including various small, intermediate and large breaks. Further, modeling and analyzing the postulated small break, intermediate break and large break LOCAs are performed with a single computer code and a single input model properly validated against relevant experimental data. Input and physical model uncertainties are combined following a random sampling process, e.g., a direct Monte Carlo approach (ASTRUM-FS) and advanced statistical procedures are utilized to show compliance with Nuclear Regulatory Commission 10 CFR 50.46 criteria. | 07-19-2012 |
20120177170 | NUCLEAR FUEL ROD PLENUM SPRING ASSEMBLY - A nuclear fuel rod plenum spring assembly that has a spacer affixed to the lower end of the ground torsion spring. The spacer has a substantially flat surface on its underside that presses against the upper surface of the upper fuel pellets to spread the load of the spring over the top surface of the upper most fuel pellet. | 07-12-2012 |
20120177167 | SELF-POWERED WIRELESS IN-CORE DETECTOR - A method and apparatus for monitoring a parameter in an irradiated environment and communicating a signal representative of the monitored parameter to a less caustic environment that employs a wireless transmitter that is powered by the irradiated environment. The power for the wireless transmitter is derived from a self-powered radiation detector disposed within the radioactive environment. | 07-12-2012 |
20120177166 | WIRELESS IN-CORE NEUTRON MONITOR - An in-core neutron monitor that employs vacuum microelectronic devices to configure an in-core instrument thimble assembly that monitors and wirelessly transmits a number of reactor parameters directly from the core of a nuclear reactor without the use of external cabling. The in-core instrument thimble assembly is substantially wholly contained within an instrument guide tube within a nuclear fuel assembly. | 07-12-2012 |
20120170705 | OPTIMUM CONFIGURATION FOR FAST REACTORS - A nuclear reactor having a liquid metal or molten salt coolant in a riser space | 07-05-2012 |
20120167839 | ANTI-VIBRATION TUBE SUPPORT PLATE ARRANGEMENT FOR STEAM GENERATORS - A means of offsetting semi-circular tube support plates typically present in heat exchangers with cross flow baffles, such as axial flow economizers, utilizing the motive force of steam generator pressurization. The offset slightly flexes the tubes, thereby providing a preload which minimizes the potential for tube vibration and wear. | 07-05-2012 |
20120167533 | DEMISTER VANE IN SITU CLEANING FIXTURE - This invention relates to an in-situ cleaning fixture for demister separator vane assemblies. The cleaning fixture includes an injection chamber with an inlet and a perforated plate which extends horizontally, substantially over the length of the demister separator vane assembly. A cleaning fluid source is connected to the injection chamber during a cleaning process. The injection chamber is designed to receive the cleaning fluid, pass the cleaning fluid through the holes formed in the plate, downwardly along a plurality of vanes in the demister separator vane assembly and into the drain trough. The in-situ cleaning fixture facilitates removal of scale build-up from the demister separator vane assembly and thereby restores the efficacy of the vanes. | 07-05-2012 |
20120155597 | NUCLEAR REACTOR AUTOMATIC DEPRESSURIZATION SYSTEM - A blocking device for preventing the actuation of an automatic depressurization system in a pressurized nuclear reactor system due to spurious signals resulting from a software failure. The blocking signal is removed when the coolant level within the core makeup tanks drop below a predetermined level. | 06-21-2012 |
20120155596 | NUCLEAR CONTROL ROD POSITION INDICATION SYSTEM - A high temperature reed switch position indicator for a pressurized water reactor in which the drive rod housing is completely immersed in the reactor coolant. The reed switch sensor modules positioned along the control rod drive rod travel housing are constructed solely of metallic, ceramic and glass materials and are sealed within an outer housing to isolate the sensor assembly from the coolant. | 06-21-2012 |
20120148011 | NUCLEAR REACTOR CAVITY ARRANGEMENTS FOR ICE CONDENSER PLANTS - A pressurized water reactor nuclear containment radiation shield which surrounds the upper portion of a pressure vessel in an ice condenser containment. The vertical walls of the neutron shield are formed in vertical sections with the lower and upper sections operable during outages, to open to promote air flow cooling along the walls in the vicinity of the vessel head. | 06-14-2012 |
20120148010 | TOP GUIDE INSPECTION FIXTURE - A fixture for performing visual inspections of the underside of the top guide of a boiling water reactor. The inspections are performed in the reactor vessel, under water, and includes a framed structure that rests on top of the top guide and supports a wheel track within a fuel assembly opening in the top guide, that follows the contour of the opening. A camera support, suspended from the frame, is then remotely, manually rotated to follow the contour of the wheel track as the fixture maintains the camera at a fixed angle and known constant distance from the underside of the top guide. | 06-14-2012 |
20120148008 | NUCLEAR REACTOR INTERNAL HYDRAULIC CONTROL ROD DRIVE MECHANISM ASSEMBLY - A control rod drive system for a nuclear reactor that employs hydraulic cylinders to operate a conventional plunger/gripper drive system to incrementally move control rods into and out of the core of a reactor. The pressure differential for driving hydraulic pistons within the cylinders is obtained from the difference in pressure between the outside and inside of the core barrel of the reactor and control of the pistons is obtained from external solenoids attached to the reactor control system. The external solenoids regulate a charging pump feed to Poppet valves that control the hydraulic feed to the cylinders. A hydraulic piston/cylinder drive is also provided for the shutdown rods which operate in either an all in or out of the core condition. | 06-14-2012 |
20120148007 | NUCLEAR REACTOR INTERNAL ELECTRIC CONTROL ROD DRIVE MECHANISM ASSEMBLY - A magnetic jack control rod drive rod drive system having the magnetic coils that operate the moving parts of the drive system wound from anodized aluminum magnet wire or ceramic coated nickel clad copper and enclosed within a hermetically sealed housing that is pressurized with helium. | 06-14-2012 |
20120148006 | NUCLEAR REACTOR CONTROL ROD DRIVE MECHANISM - A magnetic jack control rod drive mechanism for a nuclear reactor in which the stationary gripper coil, moveable gripper coil and lift coil are constructed with ceramic or quartz insulation. | 06-14-2012 |
20120103578 | MODULAR PLATE AND SHELL HEAT EXCHANGER - A modular plate and shell heat exchanger in which welded pairs of heat transfer plates are tandemly spaced and coupled in parallel between an inlet and outlet conduit to form a heat transfer assembly. The heat transfer assembly is placed in the shell in order to transfer heat from a secondary to a primary fluid. Modules of one or more of the heat transfer plates are removably connected using gaskets at the inlet and outlet conduits which are connected to a primary fluid inlet and a primary fluid outlet nozzle. The heat transfer assembly is supported by a structure which rests on an internal track which is attached to the shell and facilitates removal of the heat transfer plates. The modular plate and shell heat exchanger has a removable head integral to the shell for removal of the heat transfer assembly for inspection, maintenance and replacement. | 05-03-2012 |
20120099693 | UNIRRADIATED NUCLEAR FUEL COMPONENT TRANSPORT SYSTEM - An unirradiated nuclear fuel assembly and fuel component shipping cask that employs a liner with a universal, removable, reusable axial restraint device that can accommodate various fuel assembly designs. The restraint device has a top shear plate with a groove that encircles its peripheral edge and mates with corresponding rails on each of the walls of the liner. The top shear plate includes an anchoring mechanism for supporting a side of the top shear plate against an abutting side of a stationary wall of the liner. | 04-26-2012 |
20120099692 | SUBMERSIBLE MACHINE STRUCTURED TO CARRY A TOOL TO A LIMITED ACCESS LOCATION WITHIN A NUCLEAR CONTAINMENT - A tool for delivery of a testing element to a limited access location within a nuclear containment includes a rotation apparatus having a connection element that is configured to have an aperture that is formed generally centrally therein. A submersible machine is structured to carry the tool to a limited access location within a nuclear containment. The submersible machine includes a mount apparatus configured to be movably clamped onto a steam dam of a nuclear reactor apparatus and to drive the submersible machine circumferentially along the steam dam to inspect a plurality of jet pumps or other limited access locations of the nuclear reactor apparatus. The improved submersible machine advantageously further includes an adjustment table between the mount apparatus and a frame that carries the tool to enable rapid accurate positioning of the frame, and thus the tool, once the submersible machine has been mounted to the steam dam. | 04-26-2012 |
20120089359 | CALIBRATION DETECTION SYSTEM AND METHOD - An improved calibration detection system for use in calibrating an electronic apparatus includes a processor apparatus, an evaluation apparatus, and a connection apparatus. The connection apparatus includes a plurality of leads and is operated by the processor apparatus to internally switch and connect the various leads with various elements of the evaluation apparatus. By enabling all of the leads to be connected at the outset with the electronic apparatus and by internally switching the connections between the leads and the various elements of the evaluation apparatus, the calibration detection system saves time and avoids error in performing a testing protocol. | 04-12-2012 |
20120087458 | NUCLEAR FUEL ASSEMBLY HOLD DOWN SPRING - A nuclear fuel assembly having a plurality of multi-leaf hold down spring sets extending from a top nozzle. Each spring set consists of a multiple number of springs leafs in order to provide a large working range of spring deflection. Each spring leaf has a straight, flat base section followed by a straight, flat tapered beam with a secondary spring set having a curvature at its peripheral end. | 04-12-2012 |
20120087454 | PRIMARY NEUTRON SOURCE MULTIPLIER ASSEMBLY - A neutron emitting assembly, which is useful in nuclear reactors and other industrial applications, is made of a major amount of beryllium encapsulating a minor amount of | 04-12-2012 |
20120081108 | NONDESTRUCTIVE INSPECTION METHOD FOR A HEAT EXCHANGER EMPLOYING ADAPTIVE NOISE THRESHOLDING - A method of eddy current testing for flaws in a tube is provided that includes passing an eddy current probe through the tube and obtaining eddy current data for a number of positions along the tube, analyzing the eddy current data to generate background noise data for a number of positions along the tube, analyzing the eddy current data to generate extracted data for a number of positions along the tube, and determining whether a flaw of a particular category is present in the tube based on a set of one or more of rules applied to at least a portion of the extracted data, wherein at least one of the rules uses a particular part of the extracted data and employs a threshold that is a function a particular part of the background noise data that is associated with the particular part of the extracted data. | 04-05-2012 |
20120076255 | ALTERNATE FEEDWATER INJECTION SYSTEM TO MITIGATE THE EFFECTS OF AIRCRAFT IMPACT ON A NUCLEAR POWER PLANT - The present invention relates to an alternate feedwater injection system to at least partially mitigate the effects of an aircraft impact on a light water nuclear reactor positioned in a reactor building. The light water nuclear reactor has a primary system and a reactor core. The alternate feedwater injection system includes a water storage tank, an injection point into the primary system, a pump capable to transfer water from the water storage tank to the injection point and ultimately to the reactor core. The water storage tank and pump are located external to a reactor building and outside of an identified aircraft impact area or inside the identified aircraft impact area and provided with a means of protection from the aircraft impact. | 03-29-2012 |
20120069947 | SYSTEM FOR EXCHANGING A COMPONENT OF A NUCLEAR REACTOR - A system for installing or removing a component of a nuclear reactor, such as a CRDM, includes a riser apparatus having a lift assembly structured to hold and support the component and a first drive assembly coupled to the lift assembly and structured to selectively move the lift assembly and the component along a length of the riser apparatus, and a transition cart movable along an under vessel area of the nuclear reactor and having a pivot mechanism, wherein the riser apparatus is selectively engageable with the pivot mechanism and the pivot mechanism is structured to selectively rotate the riser apparatus from a horizontal position to a vertical position. The riser apparatus may also include a second drive assembly structured to selectively move the riser apparatus relative to the transition cart in a direction parallel to a longitudinal axis of the riser apparatus. | 03-22-2012 |
20120065927 | METHOD FOR AUTOMATED POSITION VERIFICATION - An improved method for verifying a position of a sensor with respect to an object under test includes detecting a signal from the sensor that is positioned at a given location on an object under test and comparing the signal from the sensor with a historical signal that is associated with a Uniquely Identified Location (UIL) on the object under test. If the two signals are consistent, and if the position of the sensor at the given location on the object under test is the same as the UIL, it is concluded that the position of the sensor is correct. | 03-15-2012 |
20120055330 | SYSTEM AND METHOD FOR REMOVAL OF DISSOLVED GASES IN MAKEUP WATER OF A WATER-COOLED NUCLEAR REACTOR - The present invention relates to a system and method for removing dissolved gas from makeup water in a water-cooled nuclear reactor. The present invention includes a storage tank for containing the makeup water that includes the dissolved gas, a membrane system positioned downstream of the storage tank to at least partially remove the dissolved gas front the makeup water; and a transport mechanism to transfer the makeup water from an outlet of the membrane system for use in the water-cooled nuclear reactor. The dissolved gas can include at least one of dissolved oxygen, dissolved nitrogen, dissolved argon and mixtures thereof. | 03-08-2012 |
20120014493 | REACTOR HEAD SEISMIC SUPPORT TIE ROD SYSTEM - A quick disconnect for a control rod drive mechanism seismic support tie rod system that is remotely operable from a nuclear power plant's operating deck. A wall mounted anchor in the reactor cavity contains one half of a disconnect coupling that interfaces with the other half of the disconnect coupling on the ends of the tie rods employing a remote winching system that is actuated from the top of the reactor head assembly. A latching mechanism is then actuated from the refueling cavity operating deck to lock the tie rod in place and prevent displacement during a seismic or pipe break event. The tie rod may similarly be unlocked from the wall anchor and raised above the reactor head assembly as part of a reactor head disassembly operation to gain access to the core of the reactor vessel for refueling. | 01-19-2012 |
20120002777 | TRIURANIUM DISILICIDE NUCLEAR FUEL COMPOSITION FOR USE IN LIGHT WATER REACTORS - The present invention relates to nuclear fuel compositions including triuranium disilicide. The triuranium disilicide includes a uranium component which includes uranium-235. The uranium-235 is present in an amount such that it constitutes from about 0.7% to about 20% by weight based on the total weight of the uranium component of the triuranium disilicide. The nuclear fuel compositions of the present invention are particularly useful in light water reactors. | 01-05-2012 |
20120000290 | INSPECTION VEHICLE FOR A TURBINE DISK - An inspection vehicle structured to inspect a portion of the turbine disk, preferably the blade attachment hubs, while the turbine disk is disposed within a turbine housing assembly is provided. A turbine disk is generally planar but includes a inner hub and an outer blade attachment hub. The inner hub is coupled to a shaft and the blade attachment hub provides a surface to which removable blades are attached. The area between the inner hub and outer blade attachment hub is substantially planar. The inner and blade attachment hubs are the “inspection areas” that the inspection vehicle is structured to inspect. The inspection vehicle travels over, and is magnetically coupled to, the planar surface between the two hubs. | 01-05-2012 |
20110314978 | PIPE LATHE AND SUBASSEMBLY THEREFOR - A subassembly is provided for a pipe lathe. The pipe lathe includes a segmented base ring and a drive gear assembly. The segmented base ring is structured to be removably coupled to a work piece having a perimeter. The subassembly includes a gear ring, a separate ring member, at least one tool mounting portion disposed on the separate ring member, and a plurality of fasteners. The fasteners extend through the apertures of the separate ring member and fasten the separate ring member to the gear ring, in order that the separate ring member rotates with the gear ring but not independently with respect thereto. A number of tool assemblies mount to the tool mounting portion to machine the work piece. The gear ring includes a plurality of teeth, which cooperate with the drive gear assembly of the pipe lathe to rotate the subassembly about the perimeter of the work piece. | 12-29-2011 |
20110299648 | CONTROL ROD DRIVE SHAFT UNLATCHING TOOL - A CRDS unlatching tool includes a support assembly and a latching assembly, wherein the support assembly is received within the latching assembly in a manner wherein the latching assembly is moveable relative to the support assembly. The support assembly has a plurality of latch fingers and at least one pin, each of the latch fingers being movable between a latched position wherein the latch finger is structured to engage and hold the CRDS an unlatched position wherein the latch finger is structured to not engage the CRDS. The latching assembly includes a first sleeve member and a second sleeve member, the second sleeve member having at least one slot, wherein the at least one pin is moveably received within the at least one slot. The latching assembly is movable from a latched state to an unlatched state wherein the latch fingers are actuated by the first sleeve member. | 12-08-2011 |
20110293466 | ZIRCONIUM ALLOYS WITH IMPROVED CORROSION/CREEP RESISTANCE DUE TO FINAL HEAT TREATMENTS - Articles, such as tubing or strips, which have excellent corrosion resistance to water or steam at elevated temperatures, are produced from alloys having 0.2 to 1.5 weight percent niobium, 0.01 to 0.6 weight percent iron, and optionally additional alloy elements selected from the group consisting of tin, chromium, copper, vanadium, and nickel with the balance at least 97 weight percent zirconium, including impurities, where a necessary final heat treatment includes one of i) a SRA or PRXA (15-20% RXA) final heat treatment, or ii) a PRXA (80-95% RXA) or RXA final heat treatment. | 12-01-2011 |
20110280360 | WEDGE POSITIONING APPARATUS FOR JET PUMP ASSEMBLIES IN NUCLEAR REACTORS - An auxiliary wedge positioning apparatus/assembly | 11-17-2011 |
20110274231 | DUAL DRIVE WINCH AND NUCLEAR REACTOR VESSEL MAINTENANCE APPARATUS EMPLOYING SAME - A dual drive winch having a drive assembly having a first shaft that is selectively movable between a first engaged position and a first disengaged position, and a second shaft that is selectively movable between a second engaged position and a second disengaged position. When the first shaft is in the first engaged position and the second shaft is in the second engaged position simultaneously, rotation of either the first shaft or the second shaft will, through a coupling mechanism, cause rotation of the other of the first shaft and the second shaft. | 11-10-2011 |
20110264426 | METHODOLOGY FOR MODELING THE FUEL ROD POWER DISTRIBUTION WITHIN A NUCLEAR REACTOR CORE - A method for modeling a nuclear reactor core that follows the history of each fuel pin and employs fuel pin flux form factors to explicitly track each fuel pin's fluence and burnup along its axial length and uses this information to obtain fundamental data for each fuel rod, i.e. fuel rod cross-sections, for each fuel pin segment. The data obtained for the fuel pins segments are employed to adjust the fuel pin flux form factors to match the real fuel pins' history so that the fuel rod power distribution can be precisely calculated based on the fuel rod cross-sections and the flux form factors. | 10-27-2011 |
20110235769 | CONTROL ROD TRANSFER DEVICE - A telescoping rod control cluster assembly change tool for moving control rod assemblies among fuel assemblies in a nuclear facility. The operation of the tool is completely mechanical and the telescoping feature enables the tool to have a relatively low profile when it is being moved and stored without housing a control rod assembly. Rigidly supported alignment cards guide a gripper that attaches to the control rod assembly as the control rod assembly is withdrawn into the tool with the alignment cards preventing any lateral or rotational movement of the gripper. | 09-29-2011 |
20110219609 | UNDER VESSEL LOCAL POWER RANGE MONITOR EXCHANGE TOOL - A tool for use in servicing an LPRM assembly of a nuclear reactor vessel includes a structural member having a first bore that is structured to receive an LPRM device associated with the LPRM assembly, a headpiece provided at a first end of the structural member, and a nut engaging assembly slideably mounted on the structural member. The headpiece has a plurality of projections structured to mate with a plurality of bores provided in a seal of the LPRM assembly to enable the seal to be removed, and the nut engaging assembly has a housing that defines a second bore and that has first and second nut engaging portions. The nut engaging assembly is free to slide along the structural member and over the headpiece to a position wherein the engaging portions extend beyond the headpiece so that they may be used to remove the assembly nut. | 09-15-2011 |
20110216873 | PROTECTIVE GRID ATTACHMENT - A fuel assembly for a pressurized water reactor that has a protective grid attached to the bottom nozzle through a spacer insert captured between a control rod guide thimble end plug and the bottom nozzle. A thimble screw attaches the bottom nozzle to the control rod guide thimble end plug through a central opening in the spacer insert. The control rod guide thimble end plug is provided with a raised annular boss that encircles the thimble screw shank and rests against the upper surface of the bottom nozzle through the opening in the spacer insert. The opening in the spacer insert is large enough to provide both an axial and radial clearance between the spacer insert and the end plug to accommodate differences in thermal expansion. | 09-08-2011 |
20110185989 | MINATURE SLUDGE LANCE APPARATUS - A miniature sludge lance for a steam generator in a pressurized water nuclear reactor is provided. The sludge lance is structured to enter the steam generator via an inspection opening and has a body sufficiently thin to fit between adjacent tubes. The sludge lance rail has at least two types of nozzle assemblies that may be attached thereto. One nozzle assembly rotates and another nozzle assembly translates in a vertical direction. A drive assembly, a mounting assembly, an oscillation assembly, and flow straighteners are also provided. | 08-04-2011 |
20110185988 | MINATURE SLUDGE LANCE APPARATUS - A miniature sludge lance for a steam generator in a pressurized water nuclear reactor is provided. The sludge lance is structured to enter the steam generator via an inspection opening and has a body sufficiently thin to fit between adjacent tubes. The sludge lance rail has at least two types of nozzle assemblies that may be attached thereto. One nozzle assembly rotates and another nozzle assembly translates in a vertical direction. A drive assembly, a mounting assembly, an oscillation assembly, and flow straighteners are also provided. | 08-04-2011 |
20110180022 | MINATURE SLUDGE LANCE APPARATUS - A miniature sludge lance for a steam generator in a pressurized water nuclear reactor is provided. The sludge lance is structured to enter the steam generator via an inspection opening and has a body sufficiently thin to fit between adjacent tubes. The sludge lance rail has at least two types of nozzle assemblies that may be attached thereto. One nozzle assembly rotates and another nozzle assembly translates in a vertical direction. A drive assembly, a mounting assembly, an oscillation assembly, and flow straighteners are also provided. | 07-28-2011 |
20110180021 | MINIATURE SLUDGE LANCE APPARATUS - A miniature sludge lance for a steam generator in a pressurized water nuclear reactor is provided. The sludge lance is structured to enter the steam generator via an inspection opening and has a body sufficiently thin to fit between adjacent tubes. The sludge lance rail has at least two types of nozzle assemblies that may be attached thereto. One nozzle assembly rotates and another nozzle assembly translates in a vertical direction. A drive assembly, a mounting assembly, an oscillation assembly, and flow straighteners are also provided. | 07-28-2011 |
20110174159 | PUMP SUCTION GAS SEPARATOR - A gas separator on pipe structured to transport a liquid is provided. The gas separator includes a section of piping, located upstream of a pump, having an increased diameter which is in fluid communication with an overhead pocket. Fluid flow in the portion of the pipe having an increased diameter is at a slower rate than other portions of the pipe. The slower fluid speed allows entrained gasses to stratify and float to the top of the pipe whereupon the gas will flow into the pocket. Thus, the fluid downstream from the gas separator has a reduced amount of gas in the liquid flow. | 07-21-2011 |
20110172980 | Method of Modeling Steam Generator and Processing Steam Generator Tube Data of Nuclear Power Plant - An improved method of inspecting the tubes of a steam generator of a nuclear reactor involves modeling the steam generator and comparing signals of a tube from an eddy current sensor with aspects of the model to determine whether further analysis is required. The model can advantageously include exception data with regard to particular regions of interest (ROIs) of particular tubes that is based upon historic data collected from the steam generator. | 07-14-2011 |
20110172964 | Method of Processing Steam Generator Tubes of Nuclear Power Plant - An improved method of inspecting the tubes of a steam generator of a nuclear reactor involves collecting historic data regarding the tube sheet transition regions of each tube for use during a subsequent analysis to create a new simpler signal from which historic artifacts have been removed. | 07-14-2011 |
20110170650 | PRESSURIZER WITH A MECHANICALLY ATTACHED SURGE NOZZLE THERMAL SLEEVE - A thermal sleeve is mechanically attached to the bore of a surge nozzle of a pressurizer for the primary circuit of a pressurized water reactor steam generating system. The thermal sleeve is attached with a series of keys and slots which maintain the thermal sleeve centered in the nozzle while permitting thermal growth and restricting flow between the sleeve and the interior wall of the nozzle. | 07-14-2011 |
20110164719 | NUCLEAR FUEL ASSEMBLY DEBRIS FILTER BOTTOM NOZZLE - A debris filter bottom nozzle for a pressurized water nuclear reactor fuel assembly that employs a corrugated screen in combination with flow through holes in an adapter plate to filter out potentially damaging debris. The area between the screen and the adapter plate defines a plenum that forms a collection point for the debris and coolant access is provided to the plenum through openings in the screen and sidewalls of the nozzle. | 07-07-2011 |
20110150163 | PROCESS FOR APPLICATION OF LUBRICANT TO FUEL ROD DURING FUEL ASSEMBLY LOADING PROCESS - The present invention relates generally to nuclear reactors, and more particularly, to nuclear reactors having fuel assemblies that employ support grids. A method of reducing friction and physical contact between a fuel rod and support grid in a nuclear fuel assembly is provided. The method includes applying a lubricant composition to the outer surface of the fuel rod during fuel assembly fabrication and removing the lubricant composition afterward. | 06-23-2011 |
20110148402 | INSPECTION MODE SWITCHING CIRCUIT - An eddy current probe testing apparatus structured to operate concurrently in a driver pick-up mode and said impedance mode is provided. The eddy current probe has two coils. The eddy current probe testing apparatus also includes a signal producing device, an output device, and a switch assembly. The switch assembly is structured to switch how an input signal from the signal producing device is provided to the two coils. | 06-23-2011 |
20110143578 | ELECTRICAL CONNECTOR ASSEMBLY, TEST LEAD ASSEMBLY THEREFOR, AND ASSOCIATED METHOD - A test lead assembly is provided for an electrical connector assembly, such as a terminal board. The terminal board includes a generally planar member and a number of fasteners, such as terminal screws, which are structured to fasten and electrically connect electrical conductors to the generally planar member. The test lead assembly includes an extension member having first and second opposing ends, and an intermediate portion extending therebetween. The first end is fastened to the enlarged head of a corresponding one of the terminal screws. A connection element is disposed at or about the second end of the extension member. In one embodiment the connection element is a thumb screw that electrically connects a test element to the extension member. The test lead assembly enables the terminal board to be tested, without loosening or otherwise disturbing the electrical connections of the terminal board. An associated method is also disclosed. | 06-16-2011 |
20110125462 | TETHERLESS TUBE INSPECTION SYSTEM - Apparatus and a method to inspect tubing by means of a free flying, autonomous inspection head that is not attached by wires to external control and data acquisition equipment. The inspection head travels through the tube with an attached module that integrates all the necessary support for the electronic and mechanical control of a nondestructive sensor within the inspection head. | 05-26-2011 |
20110096890 | MODULAR RADIAL NEUTRON REFLECTOR - A lower internals nuclear reactor structure having a tubular core barrel with an upper and lower open end, coaxially supported therein. A reflector having an outside curvature that substantially matches the curvature of the inside surface of the core barrel and substantially contacts the inside surface substantially over an axial length of the core, is fixedly connected to the inside surface of the core barrel at a plurality of axial and circumferential locations to be substantially supported by the inside surface of the core barrel. | 04-28-2011 |
20110089937 | EDDY CURRENT INSPECTION PROBE - An eddy current probe for inspecting steam generator tubing, that has radially outwardly biased rollers that function to center the probe and reduce friction as the probe moves along the interior of the steam generator heat exchanger tube walls. The rollers may include a braking system which controls the drag on the rollers and thus the speed of the probe along the tubing. The direction of travel of the rollers is remotely adjustable to control the inspection pattern and the force of the rollers against the interior surface of the tubing can be remotely controlled. | 04-21-2011 |
20110079186 | MINATURE SLUDGE LANCE APPARATUS - A miniature sludge lance for a steam generator in a pressurized water nuclear reactor is provided. The sludge lance is structured to enter the steam generator via an inspection opening and has a body sufficiently thin to fit between adjacent tubes. The sludge lance rail has at least two types of nozzle assemblies that may be attached thereto. One nozzle assembly rotates and another nozzle assembly translates in a vertical direction. A drive assembly, a mounting assembly, an oscillation assembly, and flow straighteners are also provided. | 04-07-2011 |
20110069802 | CONTROL ROD DRIVE OUTER FILTER REMOVAL TOOL - An outer filter removal tool for a boiling water reactor control rod drive that uses a mechanical advantage obtained through the use of lead screw threads to pull the outer filter off of the control rod drive. Fingers on the tool are closed around the upper flange of the outer filter by sliding a collar over the outwardly biased fingers. A shaft extending through the tool is rotated which in turn extends a push plate against the control rod drive index tube causing the fingers to pull against the upper flange on the outer filter until the filter is freed from the control rod drive. The tool will hold the filter in place until affirmatively released for proper disposal. | 03-24-2011 |
20110056950 | DOMED DIAPHRAGM / INSERT PLATE FOR A PRESSURE VESSEL ACCESS CLOSURE - A pressure vessel closure for an access opening that has a sealing surface surrounding the access opening and one of either a diaphragm or insert that spans the access opening with a peripheral flange that rests on the sealing surface. The insert or diaphragm has a continuously rounded portion that extends into the access opening and a cover extends over the insert or diaphragm. A gasket is interposed between the flange of the insert and the sealing surface or a fillet weld attaches the flange of the diaphragm to the sealing surface. A locking device secures the cover to a wall of the pressure vessel and urges the flange against the sealing surface. | 03-10-2011 |
20110033020 | HELICALLY FLUTED TUBULAR FUEL ROD SUPPORT - A support grid for a nuclear fuel assembly, the fuel rod assembly having a generally cylindrical fuel rod with a diameter, wherein the support grid includes a frame assembly having a plurality of outer straps and a plurality of helical frame members. The helical frame members have a contact portion structured to contact an adjacent helical frame member and at least one helical fuel rod contact portion with a lesser diameter. The lesser diameter is generally equivalent to the fuel rod diameter such that a fuel rod disposed in the helical frame member would engage the inner helical frame member at helical fuel rod contact portion. The helical frame members are coupled to each other at the contact portions thereby forming a grid. The plurality of outer straps are disposed about the perimeter of the helical frame members. | 02-10-2011 |
20110026660 | DIGITAL NUCLEAR CONTROL ROD CONTROL SYSTEM - A digital rod control system that employs separate power modules to energize the respective coils of a magnetic jack control rod drive rod drive system so that two, independently powered grippers can simultaneously support the control rod drive rod when it is not in motion to avoid dropped rods. The basic building block of the system is two or more selecting cabinets which receive multiplex power from at least one moving cabinet and are under the control of a single logic cabinet. Each of the cabinets include monitoring features to confirm the reliability of the system. | 02-03-2011 |
20110002436 | NUCLEAR FUEL ASSEMBLY SUPPORT GRID - A nuclear fuel assembly support grid formed from an array of a plurality of orthogonally arranged straps in an egg-crate configuration with angled trailing and/or leading edges that are designed to break the correlation of vortices shed from the edges of the grid straps by varying the phase of the vortices to avoid resonant vibration of the straps. | 01-06-2011 |