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

US8043448B2 · kind B2 · utility

Assignee

• Global Nuclear Fuel-Americas, LLC · US

Inventors

• David W. White · Wilmington, US
• Daniel R. Lutz · San Jose, US
• Yang-Pi Lin · Wilmington, US
• John Schardt · Wilmington, US
• Gerald Potts · Wilmington, US
• Robert H. Elkins · Hinsdale, US
• Hiroaki Kagami · Yokohama, JP
• Hideyuki Mukai · Mito, JP

Key dates

• Filing date: Sep 8, 2004
• Grant date: Oct 25, 2011
• Expiry date: Apr 26, 2029

Classification

• Technology area (CPC Y): Emerging Cross-Sectional Technologies
• CPC primary: Y02E30/30
• WIPO field: Materials, metallurgy
• WIPO sector: Chemistry

Abstract

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 α+β or β-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 2×10−19 hour with the workpiece processing generally being limited to temperatures below 680° C. for extrusion and below 625° C. for all other operations, thereby simplifying the fabrication of the nuclear reactor components while providing corrosion resistance comparable with conventional alloys.