Abstract: A high strength cladding tube for nuclear fuel for a nuclear reactor having an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: A fuel assembly for a pressurized water reactor having a fuel rod with a high strength cladding tube including an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: A method and apparatus is disclosed for converting a reheat steam turbine plant to a non-reheat, combined cycle plant without requiring internal modification of the steam turbine unit. A reheat steam turbine power plant with a boiler and a steam turbine with a plurality of turbine sections, receiving steam under pressure and successively expanding the steam through the turbine sections, is converted to a non-reheat combined cycle plant by replacing the boiler with a dual-pressure heat recovery steam generator capable of producing main steam and secondary steam at differential temperature and pressure and by installing a trimming system to allow adjustment of the pressure drop between the exhaust of the first turbine section and the inlet of the second turbine section to prevent damage to the turbine blades.

Abstract: High strength zirconium alloys with improved strength and creep resistance having 1.5 to 6 weight percent bismuth and an element or mixtures of elements selected from the group of molybdenum, tin and niobium.

Abstract: A nuclear fuel rod for a pressurized water reactor having an elongated hollow metallic tubular cladding for containing nuclear fuel, the tubular cladding having an inner tubular layer of a zirconium alloy with alloying components of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium.

Abstract: A reactor pressure vessel top guide structure inspection and transport system comprising a remotely operable and positionable underwater vehicle for traversing the top guide structure of a reactor pressure vessel and for positioning an inspection system for detecting defects in the top guide structure.

Abstract: A method and apparatus is disclosed for converting a reheat steam turbine plant to a non-reheat, combined cycle plant without requiring internal modification of the steam turbine unit. A reheat steam turbine power plant with a boiler and a steam turbine with a plurality of turbine sections, receiving steam under pressure and successively expanding the steam through the turbine sections, is converted to a non-reheat combined cycle plant by replacing the boiler with a dual-pressure heat recovery steam generator capable of producing main steam and secondary steam at differential temperature and pressure and by installing a trimming system to allow adjustment of the pressure drop between the exhaust of the first turbine section and the inlet of the second turbine section to prevent damage to the turbine blades.

Abstract: An apparatus for measuring fuel assembly bow and twist comprising a fixturing apparatus, a reference device and an ultrasonic measuring device.

Abstract: A nuclear fuel assembly for a boiling water reactor, comprising an outer channel having a substantially square cross-sectional area, and fuel rods arranged with a predetermined pitch in an array of 10.times.12 where the centers of the fuel rods are located at the vertices of equilateral triangles.

Abstract: A method of repairing a nuclear fuel rod assembly having a damaged fuel rod and/or a damaged spacer cell.

Abstract: A moisture detection system for in-process nuclear fuel powder components includes a container within which the powder is located, a purge gas supply for providing a purge gas within the container, the purge gas contacting the powder and absorbing moisture therefrom, an off-gas line for transporting at least a portion of the purge gas after the purge gas contacts the powder, and a moisture detector for determining the moisture in the purge gas. The moisture content of the purge gas provides an indication of the moisture content of the powder. A process for detecting moisture of an in-process nuclear fuel powder component includes the steps of supplying a purge gas, contacting the powder with the purge gas, transporting at least a part of the purge gas through an off-gas line, and measuring the moisture of the purge gas in the off-gas line, the moisture measurement providing a measure of moisture in the powder.

Abstract: A method for determining the density of nuclear fuel pellet using gas displacement and isothermal gas expansion.

Abstract: A boiling water reactor fuel assembly with fuel rods having variable fuel rod pitches.

Abstract: Triangular lattice for LWR square fuel assemblies.

Abstract: A nuclear fuel assembly for a pressurized water reactor having lower and upper tie plates, guide tubes, spacer grids, an instrumentation tube, and extended fuel rods which extend to the lower tie plate and which are spaced radially and supported along the guide tubes by the spacer grids, at least one of the extended fuel rods having at a lower end a fuel rod lower end cap secured by a first spring within an aperture in the lower tie plate and which exerts a lateral force against the lower end cap. The upper tie plate further includes a fuel rod support housing which extends down over the upper end of the at least one of the extended fuel rods and has a second spring positioned in a bore in the fuel rod support housing which exerts a lateral force on the upper end of the extended fuel rod positioned within the bore in the fuel rod support housing.

Abstract: A method for retrofitting an irradiated nuclear fuel assembly including removing a portion of the inlet nozzle and installing a debris filter.

Abstract: The present invention provides a fuel rod for a light water nuclear reactors that comprises a part-length fuel rod and an extension tube having at least one wall member defining an enclosed flow path therethrough, the extension tube being coupled to a portion of the part-length fuel rod so as to be disposed axially above the part-length fuel rod, and including at least one inlet opening, for allowing fluid that surrounds the rod and initially comprises a two phase mixture of steam and liquid, to enter the enclosed fluid path and at least one outlet opening located above the inlet opening, the extension tube includes means for separating at least some of the steam located in the fluid from the liquid located therein. The improved fuel rod allows steam to bypass the upper active portions of the fuel assembly.

Abstract: A boiling water reactor fuel assembly having a lowered fuel rod support plate and increased fuel rod length while restraining the fuel rods against lateral vibration.

Abstract: A reactor coolant debris filter for insertion within a nuclear fuel assembly support.

Abstract: A method for retrofitting an irradiated nuclear fuel assembly including removing a portion of the inlet nozzle and installing a debris filter.