Combustion Engineering, Inc.

Westinghouse Electric Company LLC

Anand M Garde

John P Foster

Mats Dahlback

Robert J Comstock

Satya R Pati

Guirong Pan

Jonna Partezana Mundorff

Andrew Atwood

Andrew J Mueller

Michael A McClarren

Dianna Lynn Svec Nuhfer

Hsiang Ken Yueh

James A Boshers

Penney File

Gerald D Boyer

John F Kraus

Peter M Hess

Ronald Kesterson

Terrence Cook

David Colburn

The invention relates to zirconium-based alloys and articles produced therefrom, such as tubing or strips, which have at least one of excellent corrosion resistance to water or steam and creep resistance at elevated temperatures in a nuclear reactor. The alloys include from about 0.2 to 1.5 weight percent niobium, from about 0.01 to 0.6 weight percent iron, from about 0.0 to 0.8 weight percent tin, from about 0.0 to 0.5 weight percent chromium, from about 0.0 to 0.3 weight percent copper, from about 0.0 to 0.3 weight percent vanadium, and from about 0.0 to 0.1 weight percent nickel with the balance at least 97 weight percent zirconium, including impurities. Further, the articles are formed by processes that include final heat treatment of (i) SRA or PRXA (0-33% RXA), or (ii) RXA or PRXA (80-100% RXA).

Zirconium alloys with improved corrosion/creep resistance

Articles, such as tubing or strips, which have excellent corrosion resistance to water or steam at elevated temperatures, are produced from alloys having 0.2 to 1.5 weight percent niobium, 0.01 to 0.6 weight percent iron, and optionally additional alloy elements selected from the group consisting of tin, chromium, copper, vanadium, and nickel with the balance at least 97 weight percent zirconium, including impurities, where a necessary final heat treatment includes one of i) a SRA or PRXA (15-20% RXA) final heat treatment, or ii) a PRXA (80-95% RXA) or RXA final heat treatment.

Zirconium alloys with improved corrosion/creep resistance due to final heat treatments

A tensile test apparatus and method for testing a circumferential weld on a tubular member wherein the tube region of the weld is completely supported by a mandrel and the tubular specimen is gripped on both sides of the weld along the specimen's longitudinal length with a very small separation distance between the tensile test grips. If there are any non-bonded defects in the weld, they are captured by the weld fracture surface as mesa or void features with an orientation different from that of a well bonded region.

Mandrel supported tensile test to evaluate weld bonding

The alloy of the present invention features controlled amounts of tin, nitrogen, and niobium and includes tin (Sn) in a range of greater than 0 to 1.50 wt. %, wherein 0.6 wt. % is typical. The alloy also has iron (Fe) in a range of greater than 0 to 0.24 wt. %, and typically 0.12 wt. %; chromium (Cr) in a range of greater than 0 to 0.15 wt. % and typically 0.10 wt. %; nitrogen (N) in a range of greater than 0 to 2300 ppm; silicon, in a range of greater than 0 up to 100 ppm, and typically 100 ppm; oxygen (O) in a range of greater than 0 and up to 1600 ppm, and typically 1200 ppm; niobium (Nb) in a range of greater than 0 wt. % to 0.5 wt. % and typically 0.45 wt. %; and the balance zirconium.

Zirconium alloy with tin, nitrogen, and niobium additions

A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing on a weight percentage basis tin in a range of 0.4 to 1.0 percent and typically 0.5; iron in a range of 0.3 to 0.6 percent, and typically 0.46 percent; chromium in a range of 0.2 to 0.4 percent, and typically 0.23 percent; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 1200 to 2500 ppm, typically 1800 to 2200 ppm. The high oxygen level assists in reducing hydrogen uptake of the alloy compared to Zircaloy-4, for example.

Zirconium alloy with superior corrosion resistance

An improved short-term autoclave (10) test for ex-reactor evaluation of in-reactor corrosion resistance of zirconium alloy members (16) for use in pressurized water reactors and pressurized heavy water reactors by:

Autoclave corrosion test for zirconium alloys

A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing tin in a range of 0.45 to 0.75 wt. %, and typically 0.6 wt. %; iron in a range of 0.4 to 0.53 wt. %, and typically 0.45 percent; chromium in a range of 0.2 to 0.3 wt. %, and typically 0.25 percent; niobium in a range of 0.3 to 0.5 wt. %, and typically 0.45 wt. %; nickel in a range of 0.012 to 0.03 wt. %, and typically 0.02 wt. %; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 1,000 to 2,000 ppm, and typically 1,600 ppm, with the balance zirconium. The addition of iron and niobium improves mechanical properties of the alloy with its lower level of tin, while corrosion resistance is addressed by having an iron level of 0.45 wt. % and an iron/chromium ratio on the order of 1.5. The addition of niobium also counters the effect of higher iron on the hydrogen absorption characteristics of the alloy. The addition of nickel, silicon, carbon, and oxygen as alloying elements provide desired corrosion resistance and strength.

Zirconium alloy with improved post-irradiation properties

A zirconium alloy which imparts good creep strength, while also providing favorable neutron cross section, improved corrosion resistance, low hydrogen uptake and good fabricability is described which contains vanadium in a range of from an amount effective to indicate its greater-than-trace presence up to 1.0 wt %, wherein either limit is typical; niobium in a range of from an amount effective to indicate its greater-than-trace presence up to 1.0 wt %, wherein either limit is typical; antimony in a range of from an amount effective to indicate its greater-than-trace presence up to 0.2 wt %, wherein either limit is typical; tellurium in a range of from an amount effective to indicate its greater-than-trace presence up to 0.2 wt %, wherein either limit is typical; tin in a range of from an amount effective to indicate its greater-than-trace presence up to 0.5 wt %, wherein either limit is typical; iron in a range of 0.2 to 0.5 wt %, typically 0.35 wt %; chromium in a range of from 0.1 to 0.4 wt %, typically 0.25 wt %; silicon in a range of 50 to 200 ppm, wherein either limit is typical; and oxygen in a range of from an amount effective to indicate its greater-than-trace presence up to 2200 ppm, wherein either limit is typical and the balance zirconium.

Creep resistant zirconium alloy

A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance under neutron irradiation environment in a zirconium alloy containing tin in a range of 0.8 to 1.2 percent; iron in a range of 0.2 to 0.5 percent, and typically 0.35 percent; chromium in a range of 0.1 to 0.4 percent, and typically 0.25 percent; niobium in a range of from a measurable amount up to 0.6 percent, and typically 0.30 percent; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 900 to 1800 ppm, typically 1600 ppm. The silicon is added as an alloying element to reduce hydrogen absorption by the alloy and to reduce variations in the corrosion resistance with variations in the processing history of the alloy.

Zirconium alloy with superior ductility

An improved corrosion resistant ductile modified zirconium alloy for extended burnups in water-moderated nuclear reactor core structural components, fuel cladding and analogous corrosive environment uses is provided. It comprises:

Corrosion resistant zirconium alloy

A stabilized alpha metal matrix provides an improved ductility after irradation without loss of corrosion resistance in a 'Zircaloy' alloy modified with measurable amounts of up to 0.6 percent by weight of niobium or 0.1 percent by weight of molybdenum. Tin is present in the Zircaloy in the range of 1.2 to 1.70 percent by weight and the oxygen level is in the range of from 1000 to 1600 ppm. Iron and chromium alloying element levels are those of typical Zircaloys. The average intermetallic precipitates' particle sizes are in the range of from 1200 to 1800 angstroms, thereby providing optimum corrosion resistance of the improved alloy in both boiling water and pressurized water reactors.

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West Simsbury, CT, United States of America

Anand M Garde