US20160304991A1 - Zirconium alloy having excellent corrosion resistance and creep resistance and method of manufacturing the same - Google Patents

Zirconium alloy having excellent corrosion resistance and creep resistance and method of manufacturing the same Download PDF

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Publication number
US20160304991A1
US20160304991A1 US15/097,354 US201615097354A US2016304991A1 US 20160304991 A1 US20160304991 A1 US 20160304991A1 US 201615097354 A US201615097354 A US 201615097354A US 2016304991 A1 US2016304991 A1 US 2016304991A1
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rolled
zirconium alloy
cold
vacuum annealing
subjecting
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Inventor
Min Young CHOI
Yong Kyoon Mok
Yoon Ho Kim
Yeon Soo Na
Chung Yong Lee
Tae Sik JUNG
Dae Gyun GO
Seung Jae Lee
Jae Ik Kim
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Kepco Nuclear Fuel Co Ltd
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Kepco Nuclear Fuel Co Ltd
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Assigned to KEPCO NUCLEAR FUEL CO., LTD. reassignment KEPCO NUCLEAR FUEL CO., LTD. ASSIGNMENT OF ASSIGNORS INTEREST (SEE DOCUMENT FOR DETAILS). Assignors: CHOI, MIN YOUNG, GO, DAE GYUN, JUNG, TAE SIK, KIM, JAE IK, KIM, YOON HO, LEE, CHUNG YONG, LEE, SEUNG JAE, MOK, YONG KYOON, NA, YEON SOO
Publication of US20160304991A1 publication Critical patent/US20160304991A1/en
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    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C16/00—Alloys based on zirconium
    • B—PERFORMING OPERATIONS; TRANSPORTING
    • B22—CASTING; POWDER METALLURGY
    • B22F—WORKING METALLIC POWDER; MANUFACTURE OF ARTICLES FROM METALLIC POWDER; MAKING METALLIC POWDER; APPARATUS OR DEVICES SPECIALLY ADAPTED FOR METALLIC POWDER
    • B22F1/00—Metallic powder; Treatment of metallic powder, e.g. to facilitate working or to improve properties
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22C—ALLOYS
    • C22C1/00—Making non-ferrous alloys
    • C22C1/02—Making non-ferrous alloys by melting
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
    • C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
    • C22F1/16—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of other metals or alloys based thereon
    • C22F1/18—High-melting or refractory metals or alloys based thereon
    • C22F1/186—High-melting or refractory metals or alloys based thereon of zirconium or alloys based thereon
    • B—PERFORMING OPERATIONS; TRANSPORTING
    • B22—CASTING; POWDER METALLURGY
    • B22D—CASTING OF METALS; CASTING OF OTHER SUBSTANCES BY THE SAME PROCESSES OR DEVICES
    • B22D7/00—Casting ingots, e.g. from ferrous metals
    • B22D7/005—Casting ingots, e.g. from ferrous metals from non-ferrous metals

Definitions

  • the present invention relates to a zirconium alloy having excellent corrosion resistance and creep resistance and a method of manufacturing the same and, more particularly, to a zirconium alloy composition and annealing conditions, suitable for use in nuclear fuel cladding tubes and spacer grids for light and heavy water reactor nuclear power plants.
  • Zirconium alloys having a low neutron absorption cross-section, superior corrosion resistance and mechanical properties, have been widely used for decades as materials for nuclear fuel cladding tubes, nuclear fuel assembly spaer grids, and internal structures in nuclear reactors.
  • Zircaloy-2 (Sn: 1.20 to 1.70 wt %, Fe: 0.07 to 0.20 wt %, Cr: 0.05 to 1.15 wt %, Ni: 0.03 to 0.08 wt %, O: 900 to 1500 ppm, Zr: balance) and Zircaloy-4 (Sn: 1.20 to 1.70 wt %, Fe: 0.18 to 0.24 wt %, Cr: 0.07 to 1.13 wt %, O: 900 to 1500 ppm, Ni: ⁇ 0.007 wt %, Zr: balance) are widely used in nuclear industry.
  • U.S. Pat. No. 4,649,023 discloses a zirconium alloy composed essentially of 0.5 to 2.0 wt % of Nb and 0.9 to 1.5 wt % of Sn, and including 0.09 to 0.11 wt % of any one selected from among Fe, Cr, Mo, V, Cu, Ni and W, and 0.1 to 0.16 wt % of O, and the balance of Zr. Also, there is disclosed a method of manufacturing a product in which precipitates having a small size of 80 nm or less are uniformly distributed in a matrix using the above alloy.
  • U.S. Pat. No. 5,648,995 discloses a cladding tube using a zirconium alloy comprising 0.8 to 1.3 wt % of Nb, 50 to 250 ppm of Fe, 1600 ppm or less of 0, and 120 ppm or less of Si.
  • This alloy is annealed at 600 to 800° C., extruded, and subjected to cold rolling four to five times.
  • intermediate annealing between the cold rolling processes is performed in the temperature range of 565 to 605° C. for 2 to 4 hr, and final annealing is performed at 580° C., thereby manufacturing a nuclear fuel cladding tube.
  • the alloy composition is configured such that the amounts of Fe and O are limited to 250 ppm or less and 1000 to 1600 ppm, respectively.
  • U.S. Pat. No. 6,325,966 discloses an alloy having superior corrosion resistance and mechanical properties, composed essentially of 0.15 to 0.25 wt % of Nb, 1.10 to 1.40 wt % of Sn, 0.35 to 0.45 wt % of Fe, and 0.15 to 0.25 wt % of Cr, and including 0.08 to 0.12 wt % of any one selected from among Mo, Cu, and Mn, 1000 to 1400 ppm of O, and the balance of Zr.
  • an Zr—Nb alloy from which Sn is removed and to which P, Ta and the like are added, and which is controlled in terms of composition and annealing temperatures, may improve creep resistance while significantly increasing corrosion resistance, thus culminating in the present invention.
  • an object of the present invention is to provide a zirconium alloy composition and final annealing conditions, in which Sn, which negatively affects corrosion resistance, is removed and Nb, P, Ta and the like are added to maintain creep resistance, thus ensuring optimal annealing conditions while improving corrosion resistance and creep resistance.
  • the present invention provides a zirconium alloy, comprising: 1.1 to 1.2 wt % of Nb, 0.01 to 0.2 wt % of P, 0.2 to 0.3 wt % of Fe, and the balance of Zr.
  • P is added in an amount of 0.02 to 0.07 wt %.
  • the zirconium alloy further comprises 0.01 to 0.15 wt % of Ta in order to improve corrosion resistance and creep resistance.
  • Ta is added in an amount of 0.03 to 0.1 wt %.
  • the present invention provides a method of manufacturing a zirconium alloy, comprising the steps of: (1) melting a mixture comprising 1.1 to 1.2 wt % of Nb, 0.01 to 0.2 wt % of P, 0.2 to 0.3 wt % of Fe, and the balance of Zr, thus preparing an ingot; (2) subjecting the ingot prepared in step (1) to solution heat treatment at 1,000 to 1,050° C. ( ⁇ -phase range) for 30 to 40 min and then to ⁇ -quenching using water; (3) preheating the ingot treated in step (2) at 630 to 650° C.
  • step (3) subjecting the material hot-rolled in step (3), to primary intermediate vacuum annealing at 570 to 590° C. for 3 to 4 hr and then to primarily cold-rolled at a reduction ratio of 30 to 40%; (5) subjecting the material primarily cold-rolled in step (4), to secondary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and then to secondarily cold-rolled at a reduction ratio of 50 to 60%; (6) subjecting the material secondarily cold-rolled in step (5), to tertiary intermediate vacuum annealing at 560 to 580° C.
  • step (6) subjecting the material tertiarily cold-rolled in step (6), to final vacuum annealing at 440 to 650° C. for 7 to 9 hr.
  • step (1) P is added in an amount of 0.02 to 0.07 wt %, and in step (7), the final vacuum annealing temperature is 460 to 600° C., thereby optimizing corrosion resistance and creep resistance.
  • the mixture further comprises 0.01 to 0.15 wt % of Ta, thereby further increasing corrosion resistance.
  • Ta is added in an amount of 0.03 to 0.1 wt %, and in step (7), the final vacuum annealing temperature is 460 to 530° C., thereby maximizing corrosion resistance and creep resistance.
  • P is compacted in order to prevent precipitation thereof before melting the mixture in step (1).
  • the zirconium alloy is configured such that Sn is completely removed and the kinds and amounts of added elements, such as P, Ta and the like, and final annealing conditions are controlled, thus exhibiting corrosion resistance superior to that of Zircaloy-4 and high creep resistance. Therefore, this zirconium alloy can be effectively utilized in nuclear fuel cladding tubes and the like inside reactor cores for light and heavy water reactor nuclear power plants.
  • FIG. 1 is a graph illustrating the weight gain over time in corrosion testing of the zirconium alloy according to the present invention.
  • FIG. 2 is a graph illustrating the creep strain in creep testing of the zirconium alloy according to the present invention.
  • the present invention addresses a zirconium alloy, comprising: 1.1 to 1.2 wt % of Nb, 0.02 to 0.05 wt % of P, 0.2 to 0.3 wt % of Fe, and the balance of Zr.
  • the present invention addresses a zirconium alloy, comprising: 1.1 to 1.2 wt % of Nb, 0.02 wt % of P, 0.2 to 0.3 wt % of Fe, and the balance of Zr.
  • the present invention addresses a zirconium alloy, comprising: 1.1 to 1.2 wt % of Nb, 0.05 wt % of P, 0.2 to 0.3 wt % of Fe, and the balance of Zr.
  • the present invention addresses a zirconium alloy, comprising: 1.1 to 1.2 wt % of Nb, 0.05 wt % of P, 0.03 to 0.04 wt % of Ta, 0.2 to 0.3 wt % of Fe, and the balance of Zr.
  • the present invention addresses a zirconium alloy, comprising: 1.1 to 1.2 wt % of Nb, 0.05 wt % of P, 0.09 to 0.1 wt % of Ta, 0.2 to 0.3 wt % of Fe, and the balance of Zr.
  • the present invention addresses a method of manufacturing the zirconium alloy, comprising the steps of: (1) melting a mixture of zirconium alloy elements, thus preparing an ingot; (2) subjecting the ingot prepared in step (1) to solution heat treatment at 1,000 to 1,050° C. ( ⁇ -phase range) for 30 to 40 min and then to ⁇ -quenching using water; (3) preheating the ingot treated in step (2) at 630 to 650° C. for 20 to 30 min and subjecting the ingot to hot rolling at a reduction ratio of 60 to 65%; (4) subjecting the material hot-rolled in step (3), to primary intermediate vacuum annealing at 570 to 590° C.
  • step (6) subjecting the material primarily cold-rolled in step (4), to secondary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and then to secondarily cold-rolled at a reduction ratio of 50 to 60%; (6) subjecting the material secondarily cold-rolled in step (5), to tertiary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and then to tertiarily cold-rolled at a reduction ratio of 30 to 40%; and (7) subjecting the material tertiarily cold-rolled in step (6), to final vacuum annealing.
  • step (1) 1.2 wt % of Nb, 0.02 to 0.05 wt % of P, 0.03 to 0.1 wt % of Ta, 0.2 wt % of Fe, and the balance of Zr were subjected to VAR (Vacuum Arc Remelting), thus forming an ingot.
  • VAR Vauum Arc Remelting
  • the Zr that was used is zirconium sponge (Reactor Grade ASTM B349), and the added elements, such as Nb, P, Ta, Fe and the like, have a high purity of 99.99% or more.
  • this process was repeated about three times, and the alloy was melted under the condition that the chamber for VAR was maintained at a vacuum level of 10 ⁇ 5 torr or less, thus forming an ingot. Unlike the other alloy elements, P was melted after being compacted, in order to prevent precipitation and segregation.
  • cooling was carried out inert gas environment such as argon.
  • step (2) for ⁇ -solution heat treatment and ⁇ -quenching solution heat treatment was performed for 30 min at 1,000 to 1,050° C., corresponding to the ⁇ -phase range, and then, water cooling at a rate of about 300° C./sec or more was performed. This process was performed to homogenize the alloy composition in the formed ingot and to uniformly distribute the size of SPP (Secondary Phase Particles) in the matrix.
  • SPP Single Phase Particles
  • the ingot was clad with a 1 mm thick stainless steel plate and was then spot welded.
  • step (3) the ⁇ -quenched sample was subjected to hot rolling.
  • the sample was preheated at 630 to 650° C. for about 20 to 30 min, and was then rolled at a reduction rate of about 60 to 65%. If the processing temperature falls out of the above range, it is difficult to obtain the rolled material suitable for use in subsequent step (4). Also, if the reduction rate of hot rolling is less than 60%, the texture of the zirconium material becomes non-uniform, which lead to undesirably deterioration in hydrogen embrittlement resistance. On the other hand, if the reduction rate is higher than 80%, subsequent processability may become problematic.
  • the material hot-rolled was treated as follows: the clad stainless steel plate was removed, an oxide film and impurities were removed using a pickling solution comprising water, nitric acid and hydrofluoric acid at a volume ratio of 50:40:10, and the remaining oxide film was completely removed using a wire brush in order to facilitate subsequent processing.
  • the intermediate vacuum annealing is carried out at a temperature elevated to a fully recrystallization annealing temperature. If the temperature falls out of the above range, corrosion resistance may deteriorate.
  • the rolled material was subjected to primary cold rolling at a reduction ratio of about 40 to 50% at an interval of about 0.3 mm for each pass.
  • the rolled material was subjected to secondary intermediate vacuum annealing at 570 to 580° C. for about 2 to 3 hr.
  • the rolled material was subjected to secondary cold rolling at a reduction ratio of about 50 to 60% at an interval of about 0.3 mm for each pass.
  • the rolled material was subjected to tertiary intermediate vacuum annealing at 570 to 580° C. for 2 to 3 hr.
  • the rolled material was subjected to tertiary cold rolling at a reduction ratio of about 30 to 40% at an interval of about 0.3 mm for each pass.
  • the rolled material was finally annealed in a high vacuum of 10 ⁇ 5 torr or less.
  • Zircaloy-4 As a commercially available zirconium alloy for use in nuclear power plants, Zircaloy-4 was used.
  • Each of the zirconium alloys of Examples 1 to 12 was manufactured into a sheets through the above manufacturing process, which was then fabricated a corrosion test sample having a size of 20 mm ⁇ 20 mm ⁇ 1.0 mm, followed by stepwise mechanical polishing using #400 to #1200 SiC abrasive paper.
  • the sample was pickled using a solution comprising water, nitric acid and hydrofluoric acid at a volume ratio of 50:40:10, sonicated with acetone, and then completely dried in an oven for 24 hr or longer.
  • the surface area and the initial weight of the alloy were measured before the alloy was loaded into an autoclave.
  • the loaded sample was subjected to corrosion testing for 260 days using a static autoclave at 360° C. in an 18.6 MPa pure water atmosphere.
  • the corrosion testing results were evaluated depending on 1) when P was added in amounts of 0.02 wt % and 0.05 wt % in the absence of Ta, and 2) when Ta was added in amounts of 0.03 wt % and 0.1 wt % in the presence of 0.05 wt % of P. As such, both 1) and 2) were tested at all of three final annealing temperatures of 460° C., 520° C. and 580° C.
  • Example 4 When the amount of Ta was 0.1 wt %, corrosion resistance was significantly increased in Example 4 at a final annealing temperature of 460° C. and Example 8 at a final annealing temperature of 520° C. When the amount of Ta was 0.03 wt %, corrosion resistance was insignificantly increased.
  • corrosion resistance was increased when Ta was added in an amount of 0.01 wt % to 0.15 wt %, and was remarkably increased when Ta was added in an amount of 0.03 wt % to 0.1 wt %.
  • Each of the zirconium alloys of Examples 1 to 4 was manufactured into a sheets through the above manufacturing process, which was then formed into a creep test sample.
  • a Zircaloy-4 sheet sample of Comparative Example 2 was manufactured through the same process by simulating the commercially available cladding tube of Comparative Example 1.
  • the final annealing temperature of Comparative Example 2 was set to 460° C., as in Examples 1 to 4 and Comparative Example 1, and creep testing was carried out.
  • the creep testing was performed at 350° C. under a predetermined load of 120 MPa for 120 hr, and the results thereof were compared with those of Comparative Example 2. The results are shown in Table 4 below.

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WO2019162876A1 (es) 2018-02-21 2019-08-29 Comisión Nacional De Energía Atómica (Cnea) Aleaciones de circonio con resistencia a la corrosión y temperatura de servicio mejoradas para usar en el revestimiento del combustible y las partes estructurales del núcleo de un reactor nuclear
CN113316489A (zh) * 2019-12-26 2021-08-27 Tvel股份公司 一种制造锆合金管状制品的方法
US20220184706A1 (en) * 2019-04-30 2022-06-16 Westinghouse Electric Company Llc Improved corrosion resistance of additively-manufactured zirconium alloys

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CN115011822B (zh) * 2022-06-13 2023-07-18 国核宝钛锆业股份公司 一种外方内圆锆合金型材的制备方法
CN115652237B (zh) * 2022-08-16 2023-11-24 重庆大学 含三次孪晶的锆合金及其制备方法

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US20220184706A1 (en) * 2019-04-30 2022-06-16 Westinghouse Electric Company Llc Improved corrosion resistance of additively-manufactured zirconium alloys
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JP2018514650A (ja) 2018-06-07
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