Abstract: An improved method of selecting and planning the performance of various maintenance activities on a facility such as a nuclear power plant includes determining the net present value of a number of future net savings that are expected to result from performance of the maintenance activity at a given time, and selecting and planning the maintenance activities in a fashion that maximizes net present value. The method includes, for each of a number of components and a number of time periods, determining a change in the probability that a component will fail within a time period, with the change resulting from an assumption that a maintenance activity is performed. The change in probability is multiplied with the losses associated with a failure in order to determine a gross savings from which costs are subtracted to determine net savings.

Abstract: The present invention provides a gray rod control assembly for a nuclear reactor. The gray rod control assembly includes a spider assembly having a plurality of radial extending flukes and a plurality of gray rod assemblies coupled to the flukes of the spider assembly. Each of the gray rod assemblies includes an elongated tubular member, a first end plug, a second end plug, and a neutron absorber. The neutron absorber includes a matrix of refractory metal fabricated to be porous into which a metal or metal alloy is infused. The neutron absorber is distributed among a plurality of the gray rod assemblies.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

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 tertiary shutdown system for a liquid metal reactor that eliminates the need for considering an ATWS in setting the thermal power limits of the reactor. The shutdown system includes a reservoir of neutron absorber material that is sealed by a valve that may actively dispense the absorber upon operator command, into a stagnant pool of sodium in the core that is confined to prevent the absorber material from entering the coolant flowing through the core. Additionally, the valve may be passively open to release the absorber material into the stagnant pool of sodium when the temperature at the valve exceeds a predetermined limit.

Abstract: 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.

Abstract: 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.

Abstract: The Advanced Logic System (ALS) is a complete control system architecture, based on a hardware platform rather than a software-based microprocessor system. It is significantly different from other PLC-type control system architectures, by implementing a FPGA in the central control unit. Standard FPGA logic circuits are used rather than a software-based microprocessor which eliminate problems with software based microprocessor systems, such as software common-mode failures. It provides a highly reliable system suitable for safety critical control systems, including nuclear plant protection systems. The system samples process inputs, provides for digital bus communications, applies a control logic function, and provides for controlled outputs. The architecture incorporates advanced features such as diagnostics, testability, and redundancy on multiple levels. It additionally provides significant improvements in failure detection, isolation, and mitigation for the highest level of integrity and reliability.

Abstract: 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.

Abstract: 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.

Abstract: An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

Abstract: 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.

Abstract: A control rod having a lower tip absorber material which exhibits substantially lower irradiation induced swelling than a second absorber material which extends above the lower tip absorber material. The lower tip absorber material having a substantially lower reactivity worth than the second absorber material, extends from a lower end plug of the control rod to an elevation just above a dashpot in a thimble guide tube in a nuclear fuel assembly when the control rod is fully inserted within the thimble guide tube.

Abstract: 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.