Alloys or metallic compositions – Zirconium or hafnium base
Reexamination Certificate
2001-11-01
2004-11-02
Carone, Michael J. (Department: 3641)
Alloys or metallic compositions
Zirconium or hafnium base
C420S423000, C148S672000, C376S260000, C376S261000, C376S409000, C376S410000, C376S457000
Reexamination Certificate
active
06811746
ABSTRACT:
TECHNICAL FIELD
The present invention relates to a zirconium alloy having excellent corrosion resistance and mechanical properties and a method for preparing a nuclear fuel cladding tube by zirconium alloy. More particulary, the present invention is directed to a zirconium alloy comprising Zr-aNb-bSn-cFe-dCr-eCu (a=0.05-0.4 wt %, b=0.3-0.7 wt %, c=0.1-0.4 wt %, d=0-0.2 wt % and e=0.01-0.2 wt %, provided that Nb+Sn=0.35-1.0 wt %), and to a method for preparing a zirconium alloy nuclear fuel cladding tube, comprising melting a metal mixture comprising zirconium and alloying elements to obtain an ingot, forging the ingot at &bgr; phase range, &bgr;-quenching the forged ingot in water after a solution heat-treatment at 1015-1075° C., hot-working the quenched ingot at 600-650° C., cold-working the hot-worked ingot in three to five times with intermediate vacuum annealing, and final vacuum annealing the cold-worked ingot at 460-540° C.
BACKGROUND ART OF THE INVENTION
In the past, zirconium alloys have found widespread use in nuclear reactor applications, including nuclear fuel rod cladding, nuclear fuel assembly grids and reactor core components, of a pressurized water reactor (PWR) and a boiling water reactor (BWR). Of zirconium alloys developed up to now, zircaloy-2(Sn 1.20-1.70 wt %, Fe 0.07-0.20 wt %, Cr 0.05-1.15 wt %, Ni 0.03-0.08 wt %, O 900-1500 ppm, Zr the balance) and zircaloy-4 (Sn 1.20-1.70 wt %, Fe 0.18-0.24 wt %, Cr 0.07-1.13 wt %, O 900-1500 ppm, Ni<0.07 wt %, Zr the balance) including Sn, Fe, Cr and Ni have been widely utilized.
In recent years, to improve the operations of atomic reactors, such as by a reduction of cycling cost of nuclear fuel, nuclear fuels for high burnup are considered. In the case that conventional zircaloy-2 and zircaloy-4 are used as nuclear fuel cladding tube materials, many problems including corrosion and poor mechanical strength are caused. and Thus, there is a widely recognized need for development of materials usable as nuclear fuel cladding tubes for high burnup, which are advantageous in terms of excellent corrosion resistance and mechanical strength. Therefore, in the present invention, Sn negatively affecting corrosion resistance of the zirconium alloy is added in a smaller amount, and Nb is additionally added to the alloy, thereby developing a novel zirconium alloy nuclear fuel cladding tube for high burnup, capable of compensating for an increase of corrosion and a decrease of tensile and creep strengths. Corrosion resistance and mechanical properties of the zirconium alloy depend highly on kinds and amounts of the alloying elements. Also, all properties of final products are changed according to preparation process, so that the products should be prepared by optimal process.
For conventional Nb and Sn-containing zirconium alloys and methods for preparing nuclear fuel cladding tubes thereof, U.S. Pat. No. 6,125,161 refers to a method for preparing a zirconium alloy nuclear reactor fuel cladding, the alloy comprising Sn 0.2-0.7 wt %, Fe 0.18-0.6 wt %, Cr 0.07-0.4 wt %, Nb 0.05-1.0 wt %, N<60 ppm and Zr the balance, and Sn 0.2-0.7 wt %, Fe 0.18-0.6 wt %, Cr 0.07-0.4 wt %, Nb 0.05-1.0 wt %, Ta 0.01-0.1 wt %, N<60 ppm and Zr the balance. As for non-Ta added alloy, an accumulated annealing parameter (&Sgr;A) is differently determined on a basis of 0.5 wt % of Nb. That is to say, when Nb content ranges from 0.05 to to 0.5 wt %, said parameter is limited to the range of −20<log &Sgr;A
i
<−15 and −18−10X
Nb
<log &Sgr;A
i
<−15−3.75(X
Nb
−0.2). Meanwhile, when Nb exceeds 0.5 wt %, said parameter is limited to −20<log &Sgr;A
i
<−18−2(X
Nb
−0.5).
U.S. Pat. No. 5,838,753 discloses a process for fabricating nuclear fuel rod cladding tube comprising a zirconium alloy, comprising &bgr; quenching a zirconium alloy billet consisting essentially of Nb 0.5-3.25 wt % and Sn 0.3-1.8 wt %, the balance of said alloy being essentially nuclear grade zirconium with incidental impurities, by heating to a temperature in &bgr; range above 950° C. and rapidly quenching the billet to a temperature below the &agr;+&bgr; to &agr; transformation temperature to form a martensitic structure; extruding the &bgr;-quenched billet at a temperature below 600° C. to form a hollow billet; annealing the extruded billet by heating at a temperature up to 590° C.; cold-working said annealed billet; and final annealing said pilgered annealed hollow billet to a temperature up to 590° C. As such, said nuclear fuel rod cladding tube comprises the alloy having a microstructure of &bgr;-Nb second phase precipitates uniformly distributed intragranularly and intergranularly forming radiation resistant second phase precipitates in the alloy matrix so as to result in increased resistance to aqueous corrosion compared to that of zircaloy when irradiated to high fluence. In addition, the &bgr;-quenching step is performed below 250° C. at a rate greater than about 300 K/sec. The second phase precipitates have a limited average size of 80 nm. Also, the alloy further comprising Si 150 ppm or less, C 50-200 ppm and O 400-1000 ppm has the second phase precipitates with a size of 60 nm.
EP 0 198 570 B1 refers to a process for fabricating thin-walled tubing with a thickness of 1 mm or less from a zirconium-niobium alloy containing Nb 1.0-2.5 wt % as homogeneously dispersed finely divided particles, and selected from the group consisting of Cu, Fe, Mo, Ni, W, V, and Cr as a third element, comprising &bgr;-quenching a zirconium-niobium alloy billet; extruding said &bgr;-quenched billet at a temperature no higher than 650° C. to form a tube shell; further deforming said tube shell by cold working the same in a plurality of cold working stages; annealing said tube shell, between each of said stages of cold working, at a temperature below 650° C.; and final annealing the resultant tubing at a temperature below 600° C., so as to produce a microstructure of the material having Nb particles of a size below 80 nm homogeneously dispersed therein. As for the alloy containing only Nb 1-2.5 wt %, annealing of the tube shell is performed at a temperature of from 500 to 600 ° C., and, preferably, at a temperature of about 524° C. for a period of about 7.5 hours. The final annealing is at a temperature below 500° C., and, preferably, at a temperature of about 427° C. for a period of about 4 hours. Following the extruding and prior to the further deforming, the tube shell is &bgr;-annealed by heating the same at a temperature in the range of 850-1050° C. and rapidly cooling the same.
U.S. Pat. No. 5,230,758 discloses that zirconium alloy comprising Nb 0.5-2.0 wt %, Sn 0.7-1.5 wt %, Fe 0.07-0.14 wt %, Cr 0.025-0.08 wt %, Cr-Ni 321 ppm or less, and 0.03-0.14 wt % of at least one of Cr and Ni, and at least 0.12 wt % total of Fe+Cr+Ni, and C 220 ppm or less, and the Zr the balance, is subjected to a post extrusion annealing and a series of fabrication step. Intermediate annealing temperature is 645-704° C. and the alloy is subjected to &bgr; annealing two steps prior to a final sizing.
As mentioned in the above prior arts, research has been carried out on conventional zirconium alloy comprising Nb and Sn for preparing a zirconium alloy nuclear fuel cladding tube for high burnup, with excellent corrosion resistance and improved strength by changing kinds and amounts of the elements to be added, or by adjusting conditions of working and annealing.
SUMMARY OF THE INVENTION
Leading to the present invention, the intensive and thorough research for a zirconium alloy having excellent corrosion resistance and mechanical properties, carried out by the present inventors aiming to avoid the problems encountered in the prior arts, resulted in the finding that Nb+Sn is added in an amount of 0.35-1.0 wt % to a zirconium alloy and then Fe, Cu and Cr are added thereto, whereby corrosion resistance and mechanical properties can be improved.
Accordingly, it is an object of the
Baek Jong Hyuk
Choi Byoung Kwon
Jeong Yong Hwan
Jung Youn Ho
Lee Myung Ho
Bachman & LaPointe P.C.
Carone Michael J.
Korea Atomic Energy Research Institute
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