Patent Number: 046648810
Section: description

DETAILED DESCRIPTION OF THE INVENTION A transverse section through a cladding tube 1 in accordance with the present invention is shown in FIG. 1. The cladding tube is composed of an outer layer, or tubular member, 10 bonded to an inner layer, or tubular member, 100. The outer layer 10 is composed of a first zirconium alloy having excellent resistance to in reactor aqueous corrosion, high strength, and a low creep rate. This first zirconium alloy is preferably a Zircaloy-2 alloy, a Zircaloy-4 alloy or a zirconium-niobium alloy such as zirconium-2.5 w/o niobium. The inner layer is composed of a second zirconium base alloy. This second alloy has been designed by the present inventors to have a combination of in pile resistance to crack propagation caused by pellet cladding interaction effects, as well as resistance to aqueous corrosion. Preferably the inner layer has a wall thickness between about 0.003 and 0.0045 inches. The composition ranges of this second alloy are shown in Table I, and have been selected based on the following theory. The presence of tin in conjunction with iron and chromium, and the optional addition of nickel provides enhanced aqueous corrosion resistance to zirconium. However, tin and oxygen are solid solution strengtheners of zirconium. Iron, chromium and nickel provide some additional strengthening through the formation of Zr (Fe, Ni, Cr) precipitates. The foregoing elements also decrease the creep rate of zirconium and decrease the ability of zirconium to anneal out neutron irradiation defects at the ambient reactor operating temperature, thereby increasing irradiation hardening of the material. In summary then, while tin, iron, nickel and chromium tend to improve the aqueous corrosion properties of zirconium, they also should tend to be detrimental to zirconium's ability to stop the propagation of PCI related cracks. The present inventors now submit that by limiting the tin content to 0.1 to 0.6 weight percent, and the oxygen content to less than about 350 ppm and more preferably less than 250 ppm, that the creep rate and stress relaxation rates of these alloys should be high enough to provide significantly enhanced and effective resistance to PCI crack propagation compared to the commercial zirconium alloy making up the outer portion of the cladding according to the present invention. It is further believed that when the tin content is held to about 0.2 to 0.6 weight percent in the present invention, an optimum combination of high creep rates, low neutron irradiation hardening, and aqueous corrosion resistance will be obtained. In the preferred range of about 0.2 to 0.6 wt. % tin, and more preferably 0.3 to 0.5, the creep rates of our alloy under BWR operating conditions should be comparable to that of zirconium containing less than 350 ppm oxygen, resulting in a barrier having the crack propagation resistance of zirconium, but with essentially the same aqueous corrosion resistance as commercial Zircaloy-2 or Zircaloy-4. In addition, the tin, iron, chromium and nickel contents of our alloy cause its recrystallized grain size to be significantly finer than that observed in zirconium. It is further preferred that nitrogen, which can have an adverse effect on both aqueous corrosion resistance and PCI crack propagation resistance, be limited to less than 65 ppm and more preferably less than 40 ppm. It is preferred that all other incidental impurities listed in Table 1 of ASTM B350-80 meet the requirements shown there for alloy 60802 or 60804 which are as follows, maximum impurities, in wt.%: ______________________________________ Al 0.0075 B 0.00005 Cd 0.00005 C 0.0270 Co 0.0020 Cu 0.0050 Hf 0.010 H 0.0025 Mg 0.0020 Mn 0.0050 Mo 0.0050 Ni (when not an alloying element) 0.0070 N 0.0065 Si 0.0120 Ti 0.0050 W 0.010 U 0.00035 ______________________________________ The total amount of incidental impurities (including oxygen and nitrogen) is preferably held to less than 1500 ppm and most preferably less than 1000 ppm to minimize the cumulative adverse effect incidental impurities can have on irradiation hardening. Table I of ASTM B350-80 is hereby incorporated by reference. It should be understood that the cladding chemistry requirements set forth in this application may be met by performing chemical analyses at the ingot stage of manufacture for all alloying elements and impurities, and subsequently at an intermediate stage of manufacture, such as near the coextrusion stage, for the interstitial elements, oxygen, hydrogen and nitrogen. Chemical analysis of the final size cladding is not required. The invention will be further clarified by the following examples which are intended to be purely exemplary of the present invention. Two alloys having the nominal compositions shown in Table II are melted by arc melting the required alloying additions with commercially available zirconium. TABLE II ______________________________________ Nominal Ingot Composition of Inner Layer Material Alloy A Alloy B ______________________________________ Sn 0.5 w/o 0.4 Fe 0.06 w/o 0.18 Cr 0.06 w/o .06 Ni 0.03 w/o impurity O .about.50-150 ppm .about.50-150 ppm Zr remainder, with remainder, with incidental impurities incidental impurities totalling less than totalling less than about 1500 ppm about 1500 ppm (including oxygen) (including oxygen) ______________________________________ The ingots formed are then fabricated by conventional Zircaloy primary fabrication techniques, including a beta solution treatment step, into tubular starting components for the inner layer. Tubular Zircaloy starting components for the outer layer are conventionally fabricated from ingots meeting the requirements of ASTM B350-80 for grade R60802 or R60804 and having an oxygen content between about 900 and 1600 ppm. These tubular starting components, for both the inner and outer layers, may have a cold worked, hot worked, alpha annealed, or beta quenched microstructure. The inside diameter surface of the outer layer starting component, as well as the outside diameter surface of the inner layer starting component are then machined to size, such that the clearance between the components when nested inside of each other is minimized. After machining, the components are cleaned to remove, as nearly as possible, all surface contamination from the surfaces to be bonded. After cleaning, the component surfaces to be bonded are preferably maintained under clean room conditions until they are welded together. Recontamination of the surfaces to be bonded is thereby minimized. The components are then nested inside of each other, and the annulus formed at the interface of the adjacent components is vacuum electron beam welded shut, such that a vacuum is maintained in the annulus after welding both ends of the nested components. At this stage, the unbonded tube shell assembly is ready to be processed according to the known extrusion, cold pilgering and annealing processes utilized to fabricate cladding tubes made completely of Zircaloy. Conventional Zircaloy lubricants, cleaning, straightening, and surface finishing techniques may be used in conjunction with any of the processes, both conventional and new, described in copending application Ser. Nos. 343,788 and 343,787 both filed on Jan. 29, 1982 (now continuation application Ser. Nos. 571,123 and 571,122, respectively, both filed on Jan. 13, 1984). which are hereby incorporated by reference. All of the foregoing fabrication processes will result in autogeneous, complete and continuous metallurgical bonding of the layers, except for minor, insignificant areas of unavoidable bond-line contamination. Surface beta treatment, either by laser or induction heating, as described in U.S. patent application Ser. No. 343,788 while not required to practice the present invention is clearly preferred. When used, such treatment would be performed either between the next to last and last cold pilgering passes or just prior to the next to last cold pilger pass. In either case it is preferred that the tube have had an intermediate anneal as well as being straightened, if necessary, prior to surface beta treatment. After surface beta treatment all intermediate, as well as the final anneals, should be performed below 600.degree. C. and more preferably below 550.degree. C. Most preferably, the final anneal is performed at about 500.degree. C. These low temperature anneals are used to preserve the enhanced corrosion resistance imparted by the beta surface treatment. While the surface beta treatment produces a Widmanstatten microstructure in only about the outer 10 to 40% of the wall thickness of the beta surface treated intermediate size tube, it is to be understood that enhanced aqueous corrosion resistance produced by such treatment is not confined to that area but preferably extends throughout the outer layer, as well as the inner layer and is retained after cold pilgering and annealing. Most preferably the aqueous corrosion resistance of the outer layer and inner layer are characterized by a substantially black, adherent corrosion film and a weight gain of less than about 200 mg/dm.sup.2 and more preferably less than about 100 mg/dm.sup.2 after a 24 hour 500.degree. C., 1500 psi steam test. Whether or not surface beta treatment has been used, the final anneal, after the final cold pilgering pass, is one in which the zirconium alloy inner layer is at least substantially fully recrystallized, to a grain size which is no larger than about 1/10, and more preferably between about 1/10 and 1/20, the inner layer wall thickness and the Zircaloy outer layer has been at least fully stress relief annealed. After the final anneal, conventional Zircaloy tube cleaning, straightening, final sizing and finishing steps are performed. As finished, the lined cladding is ready for loading with fissile fuel material. A preferred embodiment of a hermetically sealed boiling water reactor fuel rod is shown in FIGS. 2 and 3. As shown in FIG. 3, the fuel rod 300 utilizes the cladding 1 according to the present invention. This cladding has an outer layer 10 preferably of Zircaloy 2 or 4 metallurgically bonded to an inner layer 100 about 0.003 inches thick and composed of, for example, either alloy A or B as previously described. The overall wall thickness of the cladding is preferably about 0.029 to 0.032 inches thick. Contained within the cladding 1 are generally cylindrical fuel pellets 400 having a diameter which is preferably about 0.008 inches smaller than the inside diameter of the cladding 1 in accordance with the present invention. In a most preferred embodiment of the fuel rod 300 in accordance with the present invention, the fuel pellets 400 have been sintered to about 95% of their theoretical density and have an outside diameter of about 0.39 inches and a height of about 0.47 inches. As shown in FIG. 3 the ends 410 of each enriched pellet have been concavedly dished to minimize relative axial expansion of the hot center portion of the fuel pellet 400 in use. The edges 420 of each pellet 400 have been chamferred. The fuel pellets 400 preferably include enriched UO.sub.2 pellets, enriched UO.sub.2 +Gd.sub.2 O.sub.3 pellets, and natural UO.sub.2 pellets. Mixed oxide, UO.sub.2 +PUO.sub.2, pellets may also be used. The enriched pellets preferably contain uranium which has been enriched to include about 2.8 to 3.2 weight percent U.sub.235. As shown in FIG. 2, the fuel pellets 400 are preferably stacked into three zones within the cladding tube 1. The bottom zone A is comprised of UO.sub.2 pellets containing natural uranium. The bottom pellet in this zone abuts against the bottom Zircaloy end plug 200 which has been previously welded to the cladding tube 1. The middle portion B of the fuel pellet stack preferably makes up about at least 80% of the fuel pellet stack length and contains the aforementioned enriched uranium pellets. Enriched pellets containing about 3 to 5 weight percent gadolinium oxide (Gd.sub.2 O.sub.3) may be substituted for all or part of the enriched pellets in this zone. The top zone C of the fuel pellet stack is comprised of UO.sub.2 pellets containing natural uranium. In a preferred embodiment, the length of zone A and C are equal, and together comprise less than 20% of the fuel pellet stack length. The top pellet in the top zone C is in pressurized abutment with a spring which is compressively held between the top pellet and the top Zircaloy end cap 220 thereby forming a void space or plenum 230. Top end 220 is circumferentially welded to the cladding 1. The welded top 220 and bottom 200 end plugs in conjunction with the cladding 1 form a hermetically sealed container around the fuel pellets 400, and spring 210. The void space or plenum 230 is in communication with the clearance spaces 450 left between the pellets and the inside surface 9 of the cladding (see FIG. 3). The clearance spaces 450, 460 and void space 230 have been filled with a high purity, inert atmosphere having high thermal conductivity. Preferably, this atmosphere is high purity helium pressurized to about 2 to 5 atmospheres, and most preferably about 3 atmospheres (STP). Other embodiments of the present invention will be apparent to those skilled in the art from a consideration of this specification or practice of the invention disclosed herein. It is intended that the specification and examples be considered as illustrative only, with the true scope and spirit of the invention being indicated by the following claims.