Westinghouse Electric Corporation

Westinghouse Electric Company LLC

John P Foster

Samuel A Worcester

Robert J Comstock

George P Sabol

Anand M Garde

Mats Dahlback

Jonna Partezana Mundorff

Guirong Pan

Andrew J Mueller

Andrew Atwood

James P Dougherty

Raymond F Boyle

Charles S Cook

Richard J Jacko

Michael A McClarren

David Colburn

Pamela M Stevenson

Samuel G McDonald, Iii

Penney File

James A Boshers

Hsiang Ken Yueh

Dianna Lynn Svec Nuhfer

Ronald Kesterson

Terrence Cook

Anand Garde

Michael Mcclarren

Ken Yueh Hsiang

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 pressurized water reactor ('PWR') radial reflector fabricated of 20% to 30% cold worked AISI Type 316 stainless steel blocks will experience about 0.14% volumetric swelling at a neutron dose level of about 100 dpa at retirement. The reflector will not be susceptible to swelling-embrittlement at retirement.

PWR radial reflector

A high temperature-high pressure electrode for electrochemical potential measurement has a high temperature probe comprising an oxidized zirconium alloy tube member having an electroconductive core which includes a liquid electrolyte and a porous plug held in place by an oxidized zirconium alloy tube and end plug with an axial bore extending therethrough. The electroconductive core can comprise a liquid electrolyte of soaked zirconia sand and a second porous plug. In another embodiment, the zirconia sand is replaced by a surface oxidized zirconium alloy rod with grooves which extend between ends for containment of the liquid electrolyte. This embodiment also contains a second porous plug. In a modification of the latter embodiment, there is a bore in the proximal end of the surface oxidized rod which communicates with the grooves to provide electrical continuity. A Teflon sleeve forms an annular seal between the proximal end of the oxidized rod and the oxidized tube.

Rugged electrode for electrochemical measurements at high temperatures

A Zirlo alloy formed by beta quenching, hot deforming, recrystallize annealing and then cold deforming said alloy a plurality of times with recrystallize anneal steps performed between the cold deforming steps followed by stress relief annealing. The fabricating method can include a late stage beta quench step in place of one of the recrystallize anneal steps.

Zirlo alloy for reactor component used in high temperature aqueous

This is an improved method of fabricating Zircaloy-4 strip. The method is of the type wherein Zircaloy-4 material is vacuum melted, forged, hot reduced, beta-annealed, quenched, hot rolled, subjected to a post-hot-roll anneal and then reduced by at least two cold rolling steps, including a final cold rolling to final size, with intermediate annealing between the cold rolling steps and with a final anneal after the last cold rolling step. The improvement comprises: (a) utilizing a maximum processing temperature of 620.degree. C. between the quenching and the final cold rolling to final size; (b) utilizing a maximum intermediate annealing temperature of 520.degree. C.; and (c) utilizing hot rolling, post-hot-roll annealing, intermediate annealing and final annealing time-temperature combinations to give an A parameter of between 4.times.10.sup.-19 and 7.times.10.sup.-18 hour, where segment parameters are calculated for the hot rolling step and each annealing step, the segment parameters are calculated by taking the time, in hours, for which that step is performed, to the (-40,000/T) power, in which T is the temperature, in degrees K, at which the step is performed, and where the A parameter is the sum of the segment parameters. Preferably, the hot rolling and the post-hot-roll anneal are at 560.degree.-620.degree. C. and are for 1.5-3 hours and the intermediate annealing is at 400.degree.-520.degree. C. and is for 1.5-15 hours and the final anneal after the last cold rolling step is at 560.degree.-710.degree. C. for 1-5 hours, and the beta-anneal is at 1015.degree.-1130.degree. C. for 2-30 minutes.

Zircaloy-4 alloy having uniform and nodular corrosion resistance

This is an alloy comprising, by weight percent, 0.5-2.0 niobium, 0.7-1.5 tin, 0.07-0.14 iron, and 0.03-0.14 of at least one of nickel and chromium, and at least 0.12 total of iron, nickel and chromium, and up to 220 ppm C, and the balance essentially zirconium. Preferably, the alloy contains 0.03-0.08 chromium, and 0.03-0.08 nickel. The alloy is also preferably subjected intermediate recrystallization anneals at a temperature of about 1200.degree.-1300.degree. F., and to a beta quench two steps prior to final size.

Method of producing zirlo material for light water reactor applications

Zircaloy-4 processing for uniform and nodular corrosion resistance

This invention is for the processing of a somewhat broader range of compositions, including ZIRLO material. It controls creep rate in an alloy having, by weight percent, 0.5-2.0 niobium, 0.7-1.5 tin, 0.07-0.28 of at least one of iron, nickel and chromium and up to 220 ppm carbon, and the balance essentially zirconium. The method is of a type which utilizes subjecting the material to a post extrusion anneal, a series of intermediate area reductions and intermediate recrystallization anneals, with one of the intermediate recrystallization anneals possibly being a late stage beta-quench, a final pass area reduction, and a final stress relief anneal. The creep rate is controlled to about the desired amount per hour by the use of an average intermediate recrystallization annealing temperature and a final pass area reduction combination selected from the designated area of a designated figure, with the figure being selected based on whether the post extrusion anneal was an alpha or a beta anneal; on whether the final anneal was a stress relief anneal or a recrystallization anneal, and whether or not a late stage beta-quench was utilized, and the desired creep rate range. The method may also comprise subjecting the material to an alpha post extrusion anneal and a final stress relief anneal, and controlling the creep rate by the use of certain intermediate recrystallization annealing temperatures and final pass true area reductions.

Processing zirconium alloy used in light water reactors for specified

Zirlo material for light water reactor applications

A method of producing enhanced radial texture in zirconium alloy tubing suitable for use in forming cladding for nuclear fuel rods is provided. The tubing production method described herein employs a combination of mechanical expansion and heat treatment steps in the final stage of tubing formation to produce a single peak radial texture in the tubing, thereby imparting enhanced resistance to pellet-cladding-interaction to the finished tubing. The tubing is preferably processed to a diameter within less than about 10 to 20% of the desired final diameter, annealed, and expanded less than about 10 to 20% to the desired final diameter, thereby producing a unique radial texture in the finished tubing. In an alternative method, the finally expanded tubing is subjected to a final recrystallization anneal to produce a significantly enhanced split radial texture.

Single peak radial texture zircaloy tubing

This invention describes a tubular water reactor fuel cladding having an outer cylindrical layer composed of a conventional zirconium base alloy. Bonded to the outer cylindrical layer is a second layer composed of an alloy selected from the group of zirconium base alloys consisting of: about 0.19 to 0.6 wt. % tin, about 0.19 to 0.5 wt. % iron, and about 100 to 700 ppm oxygen; or about 0.4 to 0.6 wt. % tin, about 0.1 to 0.3 wt. % iron, about 0.1 to 0.3 wt. % nickel, and about 100 to 700 ppm oxygen.

Water reactor fuel cladding

This is a process for producing material for lining reactor fuel element claddings. Rather than using unalloyed zirconium, this invention provides for an alloy liner for the cladding. The process uses electron beam melting of zirconium, to give very low metallic impurities to reduce solid solution strengthening and second phase formation and property variability from lot to lot, while using alloying to reduce the susceptibility to steam corrosion. Preferably, oxygen is controlled to a low level as well, to provide a low, but fabricable, hardness in the alloyed liner material.

Process for the control of liner impurities and light water reactor

A nuclear fuel cladding tube composed of a single zirconium base alloy is provided preferably with a cold worked and stress relieved microstructure throughout. The cladding is characterized by excellent aqueous corrosion resistance as well as excellent resistance to PCI crack propagation. The alloy is selected from a group of zirconium base alloys containing tin, iron, chromium, and/or nickel and low levels of impurities.

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Monroeville, PA, United States of America

John P Foster