General Electric Company

Global Nuclear Fuel - Americas, LLC

General Electric Co., Wilmington Facility

Gerald A Potts

Ronald B Adamson

Daniel Reese Lutz

David W White

Yang-Pi Lin

John Frederick Schardt

Angelo Chopelas

Mark Andrew Dubecky

Robert B Elkins

Anthony P Reese

Russell Edward Stachowski

Atul Arun Karve

Harold Hartney Yeager

Robert A Rand

John R Burr

Charles McNeely

Sheikh Tahir Mahmood

Hiroaki Kagami

Robert Schneider

Hideyuki Mukai

James Brian Smith

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 system and method for assessing failure of fuel rods are disclosed. The method may include monitoring fuel rod operational conditions, comparing the fuel rod parameters to parameters limits, calculating the fuel rod performance parameters to determine the likelihood of failure of individual fuel rods, and updating plant operating parameters based on the calculated fuel rod parameters. The system may input the calculated fuel rod parameters into a fuel failure model to assess the probability of failure, and predict the probability of failure of individual fuel rods based on fuel rod parameters in the fuel failure model.

Method and system for assessing failures of fuel rods

An apparatus generating a graphical image of a core of a boiling water reactor (BWR) using at least one data set of channel deformation data including: a computer system including a display device for presenting the graphical image and a processor generating the graphical image using the at least one data set; the graphical image of the core includes symbolic representations of control blades arranged in the core, indicia identify each control blade, and indicia regarding deformation of channels adjacent each control blade, and a viewer software tool executed by the processor which accesses the at least one data set and determines a location in the core of the control blades and channels, and correlates the deformation data with the channels for display on the graphical image.

Display, visualization, and processing tool for channel distortion and cell friction mitigation

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 zirconium-alloy fuel element cladding, a method for generating regions of coarse and fine intermetallic precipitates across the cladding wall is provided. The method includes steps of specific heat treatments and anneals that coarsen precipitates in the bulk of the cladding. The method also includes at least one step in which an outer region (exterior) of the cladding is heated to the beta or alpha plus beta phase, while an inner region (interior) is maintained at a temperature at which little or no metallurgical change occurs. This method produces a composite cladding in which the outer region comprises fine precipitates and the inner region comprises coarse precipitates.

Method of fabricating Zircaloy tubing having high resistance to crack

Method of fabricating zircaloy tubing having high resistance to crack

A Zircaloy cladding having an outer region comprising fine precipitates and inner region comprising coarse precipitates is provided. The outer region comprises about 10% and the inner region comprises about 90% of the cladding wall thickness. Such Zircaloy tubing is resistant to propagation of cracks and at the same time resistant to corrosion in boiling water reactors (BWR). Resistance to damage caused by the pellet-cladding-interaction can be achieved by standard application of a zirconium or zirconium-alloy liner on the tubing inside surface.

Zircaloy tubing having high resistance to crack propagation

Method of fabricating zircalloy tubing having high resistance to crack

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San Jose, CA, United States of America

Gerald A Potts