Global Nuclear Fuel - Americas, LLC

General Electric Company

Ge-hitachi Nuclear Energy Americas LLC

Yang-Pi Lin

David W White

Kevin L Ledford

Paul Everett Cantonwine

Gerald A Potts

John Frederick Schardt

Yuk-Chiu Lau

Dennis Michael Gray

Young Jin Kim

Todd Charles Curtis

Timothy W Clark

Charles Beaty Patterson

Robert B Elkins

Peter Louis Andresen

Daniel Reese Lutz

Nicholas R Gullette

Mark Andrew Dubecky

Hiroaki Kagami

Hideyuki Mukai

Sheikh Tahir Mahmood

David William White

Garrett Scott Nave

Patricia Mccumbee

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

A method and apparatus for a fret resistant fuel rod for a Boiling Water Reactor (BWR) nuclear fuel bundle. An applied material entrained with fret resistant particles is melted or otherwise fused to a melted, thin layer of the fuel rod cladding. The applied material is made of a material that is chemically compatible with the fuel rod cladding, allowing the fret resistant particles to be captured in the thin layer of re-solidified cladding material to produce an effective and resilient fret resistant layer on an outer layer of the cladding.

Method and apparatus for a fret resistant fuel rod for a light water reactor (LWR) nuclear fuel bundle

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

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Wilmington, NC, United States of America

Yang-Pi Lin