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: Zirconium alloys for use in an aqueous environment subject to high fluence of a water reactor and characterized by improved corrosion resistance, consisting essentially of 0.3 to 1.8 weight percent tin, 0.1 to 0.65 weight percent iron, the balance of alloys being essentially nuclear grade zirconium with incidental impurities and having a microstructure of Zr.sub.3 Fe second phase precipitates distributed uniformly intragranularly and intergranularly to form radiation resistant second phase precipitates in the alloy matrix.

Abstract: Zirconium alloys for use in an aqueous environment subject to high fluence of a water reactor and characterized by improved corrosion resistance, consisting essentially of from 0.5 to 3.25 weight percent niobium, from 0.3 to 1.8 weight percent tin, the balance of alloys being essentially nuclear grade zirconium with incidental impurities and having a microstructure of beta niobium second phase precipitates distributed uniformly intragranularly and intergranularly to form radiation resistant second phase precipitates in the alloy matrix.

Abstract: A process for fabricating nuclear fuel rod cladding tube comprising beta quenching a zirconium alloy billet consisting essentially of from 0.5 to 3.25 weight percent niobium, from 0.3 to 1.8 weight percent tin, the balance of the alloy being essentially nuclear grade zirconium with incidental impurities by heating to a temperature in the beta range above 950.degree. C. and rapidly quenching the billet to a temperature below the .alpha. plus .beta. to .alpha. transformation temperature to form a martensitic structure; extruding the beta-quenched billet at a temperature below 600.degree. C. to form a hollow; annealing the hollow by heating at a temperature up to 590.degree. C.; pilgering the annealed hollow; and final annealing the pilgered annealed hollow to a temperature up to 590.degree. C.

Abstract: A process for fabricating nuclear fuel rod cladding tube comprising beta quenching a zirconium alloy billet consisting essentially of 0.3 to 1.8 weight percent tin, 0.1 to 0.65 weight percent iron, the balance of the alloy being essentially nuclear grade zirconium with incidental impurities by heating to a temperature in the beta range greater than about 1000.degree. C. and rapidly quenching the billet to a temperature below the .alpha. plus .beta. to a transformation temperature to form a martensitic structure; extruding the beta-quenched billet at a temperature between 600.degree. and 750.degree. C. to form a hollow; annealing the hollow by heating at a temperature up to about 700.degree. C.; pilgering the annealed hollow; and final annealing the pilgered annealed hollow to a temperature up to about 700.degree. C. to form the nuclear fuel rod cladding tube comprising the alloy having a microstructure of Zr.sub.

Abstract: An electric load center having a neutral bar having an elongated body portion with a trapezoidal shaped cross-sectional area with outwardly sloping downwardly extending side walls having apertures for insertion of electrical wires and a base portion having mounting rails being adapted to form a generally T-shaped slot, and a basepan having a surface with an upwardly extending flexible snap hook for engaging the mounting rail of the T-shaped slot.

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: A neutral tie bar comprised of a unitary member having a center portion and vertically offset laterally extending ends. Disposed between the center portion and each end is a cable lug for a branch neutral cable. Extending between one neutral cable lug and an end is a neutral cable lug for an incoming neutral cable. The neutral tie bar being adapted to be electrically connected to the enclosure of the load center.

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 system and method of the present invention measures the relative stress in the cladding of a nuclear fuel rod caused by the internal gas pressure of the rod. This is done by determining the attenuation ratio of two specific frequencies from a broadband low frequency acoustic energy spectrum. Relative pressure differentials are measurable and therefore any loss of fuel rod gas pressure is a direct indication of fuel rod cladding failure.

Abstract: An electric load center having a basepan having a surface with two spaced apart integrally formed wall members extending upward from the surface forming a barrier between the bus bars of the load center.

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: 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: An electric load center having a basepan having a surface neutral bar wire guide comprising a wall upwardly extending from the surface member and having semicircular slots extending therethrough at an upper surface for receiving and guiding a return wire into an aperture in the neutral bar.

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: An electric load center having a basepan having a surface, the surface having a tab with a tapered lower edge and a prong extending from a side edge of the tab for capture by lances found in the enclosure of the load center.

Abstract: A basepan for an electric load center having a surface with integrally formed posts with an undercut extending around its periphery for securing a component of the load center to the surface by heat staking.

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.