Abstract: Embodiments of the invention relate to an axial locking system for a row of blades in a turbine engine. While achieving the desired axial restraint, the system is configured to allow remote manipulation of the locking system. For example, a locking system provided on the upstream side of a row of blades can be accessed from the downstream side of the blades. Such a system can provide significant cost savings during the installation and removal of the blades. In one embodiment, the axial locking system can include a plate with a rotatable cam. Cooperating with the disc, the cam can be rotated to bring the radially outer end of the plate into locking engagement with the turbine blades. The cam can be manipulated from the other side of the disc by a tool extending through a passage formed between the disc and the blade root.

Abstract: High strength zirconium alloys with improved strength and creep resistance having 1.5 to 6 weight percent bismuth, and niobium.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A method for removing low levels of heavy metal from chemical/industrial waste streams containing interfering anions which typically form strong complexes with heavy metals such as uranium utilizes an ion exchange resin having amino-phosphonic functional groups which selectively capture the heavy metal from the chemical waste stream in an efficient manner such that the amount of heavy metal in the waste stream is reduced to sub-ppm levels. After capture, the resin is treated with a sequestering agent, such as a solution of ammonium fluoride or ammonium sulfate, which releases the heavy metal from the resin. The resin is then regenerated with a sodium solution and returned to service. Utilizing ion exchange and a resin containing amino-phosphonic functional groups allows removal of heavy metal to a level of less than or equal to 0.1 ppm which is a substantial reduction in view of the presence of significant quantities of complexing anions in the waste stream.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: An apparatus useful in clamping stator coil windings of a turbine generator includes a retaining platform for extending across a pair of the coil windings positioned in a spaced relation. The retaining platform includes a bore extending therethrough. A connecting insert member is slidably carried within the bore and has a through-hole with separable mating segments for admitting a rod into the through-hole. An elongated rod passes through the through-hole of the connecting insert member, which rod includes a bulbous shaped end dimensioned for passing through the bore of the retaining platform but cannot pass through the through-hole of the connecting insert member. Thus providing tension in the elongated rod biases the platform against the pair of coil windings.

Abstract: An ultrasonic inspection system for ultrasonic inspection of the guide structure of a reactor pressure vessel having two shear wave transducers, one positioned to face in a first direction about 30.degree. from normal to a top edge of the grid members of the top guide structure and a second one positioned to face in a second direction about 30.degree. from normal to a top edge of the grid members of the top guide.

Abstract: A process for recovering uranium from a waste sludge involves treating the sludge with a non-reactive, non-absorptive cellulose filtration media, adding an oxidant and heating the mixture to dissolve any uranium present. The mixture is passed through a filtration device precoated with the same filtration media, where the solids are captured while the dissolved uranium passes through as part of the filtrate, for recovery.

Abstract: A corrosion and hydride resistant nuclear fuel rod having a highly corrosion resistant outer portion in which hydride precipitation is inhibited and an inner portion in which hydride precipitation is promoted.

Abstract: A nuclear fuel pellet having concentric layers of nuclear fuel in which an outer layer has less fissionable nuclear fuel per unit volume and substantially the same amount of fertile material per unit volume as an inner core.

Abstract: A nuclear fuel pellet having concentric layers of nuclear fuel in which an outer layer has less fissionable nuclear fuel per unit volume and substantially the same amount of fertile material per unit volume as an inner core.

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 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: 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 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.