| WESTINGHOUSE ELECTRIC COMPANY LLC Patent applications |
| Patent application number | Title | Published |
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| 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 |
| 20110002435 | 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 |
| 20110002434 | METHOD OF APPLYING A BURNABLE POISON ONTO THE EXTERIOR OF NUCLEAR FUEL ROD CLADDING - An article made by applying a burnable poison onto the cladding of a nuclear fuel rod, which involves providing a nuclear fuel rod and at least one application device, rotating the nuclear fuel rod, optionally removing one or more oxides and/or surface deposits on the outer surface of the nuclear fuel rod by spraying an abrasive material onto the nuclear fuel rod via the application device while adjusting the position of the application device in relation to the nuclear fuel rod, and applying burnable poison particles onto the outer surface of the nuclear fuel rod by spraying the burnable poison onto the nuclear fuel rod via the application device while adjusting the position of the application device in relation to the nuclear fuel rod, where the burnable poison particles are applied at a velocity sufficient to cause adhesion to the outer surface of the cladding. | 01-06-2011 |
| 20110002432 | INCORE INSTRUMENT CORE PERFORMANCE VERIFICATION METHOD - A subcritical physics testing program which utilizes vanadium self-powered incore instrumentation thimble assemblies to provide an actual measured powered distribution that is used to confirm that the core will operate as designed. The signals received from the incore detector elements are integrated until a fractional uncertainty is less than a specified level. The measured power distribution is then compared against a predicted power distribution for a given rod position or temperature difference. If the measured power distribution is within a specified tolerance to the predicted power distribution, then the core is expected to behave as predicted. | 01-06-2011 |
| 20100329890 | REACTOR COOLANT PUMP MOTOR LOAD-BEARING ASSEMBLY CONFIGURATION - A method of upgrading a 1,500 rpm reactor coolant pump motor having a vertically oriented rotor shaft supported by a lower guide bearing disposed in an oil reservoir. The method includes removing a lower guide bearing support to enhance circulation of the oil to facilitate cooling of the oil and operating the reactor coolant pump motor at 1,500 rpm without the lower guide bearing support or the oil baffle or flinger ring attached to the lower guide bearing support. The method further includes thickening a keeper to support the guide bearing. | 12-30-2010 |
| 20100329411 | 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. | 12-30-2010 |
| 20100322371 | OPTIMIZED FLOWER TUBES AND OPTIMIZED ADVANCED GRID CONFIGURATIONS - 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 tubular members and/or helical frame members. The tubular members/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 fuel rod contact portion may have a variable pitch. | 12-23-2010 |
| 20100310034 | NUCLEAR FUEL ASSEMBLY PROTECTIVE BOTTOM GRID - A protector bottom grid for a nuclear fuel assembly that includes three laterally staggered and horizontally oriented protrusions that extend into the fuel rod cell of a support grid below a vertically oriented spring. The three staggered protrusions extend into the cell a distance that maintains a space between the protrusions and the fuel rod. The vertically oriented spring biases the fuel rod against a dimple extending from the opposite cell wall that is at an elevation just above the spring. The protrusions below the spring trap incoming debris in the area of the fuel rod end cap and protect the fuel rod cladding from fretting. | 12-09-2010 |
| 20100296618 | EXPANSION GAP RADIATION SHIELD - An expansion gap radiation shield is formed from a flexible container housing a radiation shielding fluid, that is located within a variable gap in permanent shielding. The invention reduces radiation dose rates outside the gap when the radiation sources are located on the opposite side of the gap. The device accommodates varying gap sizes with no loss of shielding capability. | 11-25-2010 |
| 20100278704 | TWO STEP DRY UO2 PRODUCTION PROCESS UTILIZING A POSITIVE SEALING VALVE MEANS BETWEEN STEPS - The present invention provides a two-step process for producing nuclear grade, active uranium dioxide (UO | 11-04-2010 |
| 20100276128 | 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. | 11-04-2010 |
| 20100275694 | PIPE SCANNER - A pipe scanner for non-destructively scanning an extended length of the circumference of a pipe along an axial dimension. The pipe scanner includes a collar sized to fit around the outer circumference of the pipe. Wheels supported on the collar ride on the surface of the pipe while maintaining a space between the inner surface of the collar and the outer surface of the pipe. A track extends circumferentially around the collar for guiding a circumferential drive unit that rides on the track and carries a non-destructive sensor for monitoring the surface of the pipe as the circumferential drive unit moves around the track. An axial drive unit is connected to the collar, having a plurality of circumferentially spaced drive wheels in contact with the pipe for moving the collar along the extended length. | 11-04-2010 |
| 20100275691 | NON-DESTRUCTIVE PIPE SCANNER - A non-destructive examination pipe scanner that employs a main carriage on which a sensor is mounted, a tensioner carriage and an idler carriage. The main carriage and the tensioner carriage are positioned around the pipe at spaced locations and connected on either side by a spring band. The idler carriage is slidably supported by the spring band in between the main carriage and the tensioner carriage. The sensor on the main carriage collects data about a circumferential weld on the pipe as the main carriage drives the tensioner carriage and the idler carriage around the circumference. The tensioner carriage has a variable length connection that adjusts the tension on the spring band to urge the main carriage, idler carriage and tensioner carriage into contact with the surface of the pipe. | 11-04-2010 |
| 20100246748 | NUCLEAR FUEL ASSEMBLY WITH PIVOT DIMPLED GRIDS - A soft pivot dimple nuclear fuel assembly grid that utilizes a “dog bone” shaped window cutout and radius coining of edges perpendicular to coolant flow, to reduce the susceptibility of fuel rod leaking during the reactor operation. Radius coining allows the fuel rod to smoothly transition over the radiused edge to the flat rod contact section of the dimple. The symmetric “dog bone” shape enables the dimple to pivot during rod loading resulting in improved alignment between the dimple and the fuel rod, thereby minimizing scratching. The “dog bone” shape also allows for a large contact area dimple to be softer than a typical dimple which reduces contact stresses and fretting wear during reactor operations. | 09-30-2010 |
| 20100246746 | 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. | 09-30-2010 |
| 20100226472 | NUCLEAR FUEL ELEMENT AND ASSEMBLY - A nuclear fuel element having a thick walled lower section that transitions to a thinner walled upper section with the transition forming an annular interior ledge that supports the fuel pellets spaced above a bottom end plug. The space between the fuel pellets and the bottom end plug forms a gas collection plenum that assures the necessary void volume exists to maintain margin to rod internal pressure limits. | 09-09-2010 |
| 20100150715 | THERMALLY ACTIVATED SHUTDOWN SEALS FOR ROTATABLE SHAFTS - A thermally actuated shutdown seal for a rotating shaft having a narrow annular fluid flow path surrounding the shaft. The seal surrounds the shaft with the annulus therebetween during normal operation and constricts against the shaft when the shaft slows or stops rotating. The annulus is maintained open during normal operation by a spacer interposed between opposing ends of a split ring. When the shaft stops rotating, the temperature of the annulus rises, which actuates removal of the spacer from the split ring constricting the split ring against the shaft blocking the annulus. The blocked annulus causes a pressure differential across the seal which urges a polymer seal ring, downstream of the split ring against the shaft which seals the annulus. In one embodiment, the spacer is formed of a meltable material. In a second embodiment, the spacer is removed from the split ring by a thermally responsive actuator. | 06-17-2010 |
| 20100150298 | 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. | 06-17-2010 |
| 20100150294 | 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. | 06-17-2010 |
| 20100128834 | ZIRCONIUM ALLOYS WITH IMPROVED CORROSION RESISTANCE AND METHOD FOR FABRICATING ZIRCONIUM ALLOYS WITH IMPROVED CORROSION RESISTANCE - 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 0.45 weight percent iron, at least one additional alloy element selected from 0.02 to 0.8 weight percent tin, 0.05 to 0.5 weight percent chromium, 0.02 to 0.3 weight percent copper, 0.1 to 0.3 weight percent vanadium, 0.01 to 0.1 weight percent nickel, the balance at least 97 weight percent zirconium, including impurities, wherein the alloy may be fabricated from a process of forging the zirconium alloy into a material, beta quenching the material, forming the material by extruding or hot rolling the material, cold working the material with one or a multiplicity of cold working steps, wherein the cold working step includes cold reducing the material and annealing the material at an intermediate anneal temperature of 960°-1105° F., and final working and annealing of the material. The articles formed also show improved weld corrosion resistance with the addition of chromium. | 05-27-2010 |
| 20100119030 | NUCLEAR REACTOR INTERNALS ALIGNMENT CONFIGURATION - An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate. | 05-13-2010 |
| 20100111800 | PRODUCTION OF NUCLEAR GRADE ENRICHED GADOLINIUM AND ERBIUM USING VOLATILE Gd OR Er SPECIES USING AN AERODYNAMIC PROCESS - A method of making Gd or Er isotopes from gaseous compounds containing —BH | 05-06-2010 |
| 20100071470 | STEAM GENERATOR NONDESTRUCTIVE EXAMINATION METHOD - A method of examining a steam generator heat exchange tube from the outside surface employing ultrasonic nondestructive inspection techniques. An ultrasonic transducer contacts the outside surface of the tube and transmits a pseudo helical Lamb wave into the wall of the tube chosen to have a mode that does not significantly interact with water in the tube. The reflected waves are then analyzed for changes in modes to identify defects in the wall of the tube. | 03-25-2010 |
| 20090285345 | METHOD AND TOOLING FOR DISMANTLING, CASKING AND REMOVAL OF NUCLEAR REACTOR CORE STRUCTURES - A method is provided for removing radioactive internals structural members in the core of a reactor pressure vessel in a containment vessel. The method includes placing a first cask in a first internals assembly, detaching radioactive first internals structural members from second internals structural members of the first internals assembly, placing the detached first internals structural members in the first cask, placing the first internals assembly in a second cask, and removing the second cask containing the first internals assembly and containing the casked detached radioactive first internals members from the containment vessel. The first internal members may be radioactive baffle plates, and the second internals members may be former plates bolted to the radioactive baffle plates. Novel tooling, framework and fixtures facilitate disassembling, moving and storing the radioactive baffle plates. | 11-19-2009 |
| 20090252281 | TUBE-IN-TUBE THREADED DASHPOT END PLUG - A nuclear fuel assembly having a tube-in-tube control rod guide tube design that incorporates an end plug that extends axially upward to an elevation above the lower most grid where it is sealed at its upper end to the lower end of the control rod guide tube. The guide tube lower end plug has a threaded recess in its upper surface that mates with a corresponding dashpot end plug threaded extension that is formed as an insert in the lower end of the guide tube. A hole formed through the outer wall of the guide tube end plug at the elevation of the lower portion of the recess provides a positive inspection port for assuring the proper seating of the dashpot. A method of manufacture of such a fuel assembly is also disclosed. | 10-08-2009 |
| 20090245451 | MOBILE RIGGING STRUCTURE - A U-framed structure that spans a reactor vessel of a pressurized water reactor power plant and rides on wheels that fit on the rails of the nuclear plant's refueling machine. A curved monorail is supported on the underside of the U-frame structure and guides a trolley system which travels on the monorail. The trolley system supports a hoist which is used for lifting, positioning and lowering reactor service equipment on the floor of the power plant's refueling canal. | 10-01-2009 |
| 20090141850 | PRESSURIZED WATER REACTOR PRESSURIZER HEATER SHEATH - A pressurizer whose heater sheaths are conditioned to reduce the residual stresses resulting from cold working during manufacture. After material conditioning, the heater sheath undergoes a surface conditioning treatment to add outer surface compressive stresses. | 06-04-2009 |
