General Electric Company

Global Nuclear Fuel - Americas, LLC

Ge-hitachi Nuclear Energy Americas LLC

David W White

Yang-Pi Lin

Daniel Reese Lutz

Gerald A Potts

John Frederick Schardt

Yuk-Chiu Lau

Dennis Michael Gray

Young Jin Kim

Robert B Elkins

John L Harmon

Kevin L Ledford

Paul Everett Cantonwine

Todd Charles Curtis

Timothy W Clark

Nicholas R Gullette

Charles Beaty Patterson

Harold B King

Frederick C Schoenig, Jr

Russell Patrick Higgins

Gerald M Latter

Anthony P Reese

Edward A Croteau

Christian D Frederickson

Donald G Muncy

Fletcher C Eddens

Mark Andrew Dubecky

Richard G Patterson

Sheikh Tahir Mahmood

Hiroaki Kagami

Hideyuki Mukai

Daniel R Lutz

Nuclear reactor components are treated with thermal methods to increase wear resistance. Example treatments include thermal treatments using particulate or powderized materials to form a coating. Methods can use cold spray, with low heat and high velocities to blast particles on the surface. The particles impact and mechanically deform, forming an interlocking coating with the surface and each other without melting or chemically reacting. Materials in the particles and resultant coatings include metallic alloys, ceramics, and/or metal oxides. Nuclear reactor components useable with methods of increased wear resistance include nuclear fuel rods and assemblies containing the same. Coatings may be formed on any desired surface, including fuel rod positions where spacer contact and fretting is most likely.

Fuel rods with wear-inhibiting coatings

Nuclear reactor components are treated with thermal methods to increase wear resistance. Example treatments include thermal treatments using particulate or powderized materials to form a coating. Methods can use cold spray, with low heat and high velocities to blast particles on the surface. The particles impact and mechanically deform, forming an interlocking coating with the surface and each other without melting or chemically reacting. Materials in the particles and resultant coatings include metallic alloys, ceramics, and/or metal oxides. Nuclear reactor components usable with methods of increased wear resistance include nuclear fuel rods and assemblies containing the same. Coatings may be formed on any desired surface, including fuel rod positions where spacer contact and fretting is most likely.

Fuel rods with wear-inhibiting coatings and methods of making the same

There is provided a fuel rod assembly comprising a first component of a zirconium-based material. The first component is in contact with or is located adjacent to a second component of a material different from the zirconium-based material, e.g. a nickel-based or iron-based alloy. A coating is disposed on an outer surface of the first component, which is effective to reduce an electrochemical corrosion potential difference between the first component and the second component relative to an electrochemical corrosion potential difference between the first component and the second component without the coating.

Fuel rod assembly and method for mitigating the radiation-enhanced corrosion of a zirconium-based component

An alloy according to example embodiments of the present invention may include zirconium, tin, iron, chromium, and nickel, with a majority of the alloy being zirconium. The composition of the alloy may be about 0.85-2.00% tin by weight, about 0.15-0.30% iron by weight, about 0.40-0.75% chromium by weight, and less than 0.01% nickel by weight. The alloy may further include 0.004-0.020% silicon by weight, 0.004-0.020% carbon by weight, and/or 0.05-0.20% oxygen by weight. Accordingly, the alloy exhibits reduced hydrogen absorption and improved corrosion resistance and may be used to form a fuel assembly component.

Zirconium alloys exhibiting reduced hydrogen absorption

Disclosed herein are zirconium-based alloys and methods of fabricating nuclear reactor components, particularly fuel cladding tubes, from such alloys that exhibit improved corrosion resistance in aggressive coolant compositions. The fabrication steps include a late-stage &#x3b2;-treatment on the outer region of the tubes. The zirconium-based alloys will include between about 1.30 and 1.60 wt % tin; between about 0.06 and 0.15 wt % chromium; between about 0.16 and 0.24 wt % iron, and between 0.05 and 0.08 wt % nickel, with the total content of the iron, chromium and nickel comprising above about 0.31 wt % of the alloy and will be characterized by second phase precipitates having an average size typically less than about 40 nm. The final finished cladding will have a surface roughness of less than about 0.50 &#x3bc;m Ra and preferably less then about 0.10 &#x3bc;m Ra.

Zirconium alloy fuel cladding for operation in aggressive water chemistry

A zirconium alloy suitable for forming reactor components that exhibit reduced irradiation growth and improved corrosion resistance during operation of a light water reactor (LWR), for example, a boiling water reactor (BWR). During operation of the reactor, the reactor components will be exposed to a strong, and frequently asymmetrical, radiation fields sufficient to induce or accelerate corrosion of the irradiated alloy surfaces within the reactor core. Reactor components fabricated from the disclosed zirconium alloy will also tend to exhibit an improved tolerance for cold-working during fabrication of the component, thereby simplifying the fabrication of such components by reducing or eliminating subsequent thermal processing, for example, anneals, without unduly degrading the performance of the finished component.

LWR flow channel with reduced susceptibility to deformation and control blade interference under exposure to neutron radiation and corrosion fields

Disclosed herein are zirconium-based alloys that may be fabricated to form nuclear reactor components, particularly fuel cladding tubes, that exhibit sufficient corrosion resistance and hydrogen absorption characteristics, without requiring a late stage &#x3b1;+&#x3b2; or &#x3b2;-quenching processes. The zirconium-base alloys will include between about 1.30-1.60 wt % tin; 0.0975-0.15 wt % chromium; 0.16-0.24 wt % iron; and up to about 0.08 wt % nickel, with the total content of the iron, chromium and nickel comprising at least about 0.3175 wt % of the alloy. The resulting components will exhibit a surface region having a mean precipitate sizing of between about 50 and 100 nm and a Sigma A of less than about 2&#xd7;10hour with the workpiece processing generally being limited to temperatures below 680&#xb0; C. for extrusion and below 625&#xb0; C. for all other operations, thereby simplifying the fabrication of the nuclear reactor components while providing corrosion resistance comparable with conventional alloys.

Non-heat treated zirconium alloy fuel cladding and a method of manufacturing the same

In a nuclear reactor fuel bundle assembly having a plurality of fuel rods and a non-round water rod extending between upper and lower tie plates, and having at least one fuel rod spacer located along the non-round water rod, an improvement which includes cooperating components on the fuel rod spacer and on the water rod for permitting movement of the spacer along the water rod to a desired axial location and for thereafter preventing further axial movement of the spacer in at least one of two opposite axial directions, the cooperating components including at least one spring on one of the spacer and the water rod.

Spacer capture mechansim for non-round water rods

In the automated production line manufacture of nuclear fuel bundle spacers, rod-supporting ferrules are cut from tubular stock, checked for length, laser cut to create predetermined features in the ferrule sidewall, formed with rod-centering stops in the ferrule sidewall, and checked for stop height dimensional accuracy. Pairs of ferrules are subassembled with a captive spring and assembled into a spacer assembly fixture together with other spacer parts. The fixtured spacer parts are conjoined by laser welding to create a structurally rigid spacer. The various production line operating stations are served by separate controllers linked to a central computer for production status reporting, parts traceability, and overall control purposes.

Method of manufacturing nuclear fuel bundle spacers

There is provided a nuclear fuel element having a zirconium alloy cladding tube with improved corrosion resistance. The cladding tube comprises a metallurgical gradient across the width of the tube wall wherein the tube has a more corrosion-resistant metallurgical condition at the outer circumference and a less corrosion-resistant metallurgical condition at the inner circumference. The metallurgical gradient can be generated by heating an outer circumferential portion of the tube to the high alpha or mixed alpha plus beta range while maintaining the inner surface at a lower temperature, followed by cooling of the tube.

Heat treated tube

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Wilmington, NC, United States of America

David W White