Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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).

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: A mobile boration system (60) has a number of components that are mobile and include a water source (10), H2BO3 powder supply (14), a mixer to mix the solution (20) capable of providing a boric acid solution (30) with minimal air entrainment and optional heat exchanger(s) (12), and wherein the system (60) is capable of transport to a nuclear power plant facility by land, sea or air, rather than being in place in a large vulnerable footprint.

Abstract: 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.