Patent Number: 047755083
Section: description

DETAILED DESCRIPTION OF THE INVENTION As shown in the FIGURE, and in accordance with the present invention, a composite fuel cladding tube 1 is provided having two concentric layers, each composed of a different zirconium base alloy. The outer layer 10 is composed of a conventional high strength zirconium base alloy known for its excellent corrosion resistance in aqueous environments. This first alloy may be either Zircaloy-2 or Zircaloy-4. The Zircaloy-2 or 4 utilized preferably conforms to the chemistry specification published in ASTM B350-80 Table 1 for UNS 60802 (Zircaloy-2) or UNS 60804 (Zircaloy-4). In addition the oxygen content of these alloys should be between 900 and 1600 ppm. Metallurgically bonded to and located within the outer layer is a second cylindrical layer 20 having the composition shown in Table I, below. TABLE I ______________________________________ Preferred Preferred Composition Composition A Composition B (wt. %) (wt. %) (wt. %) ______________________________________ Sn .1-.3 .1-.3 .1-.3 Fe .05-.2 .05-.2 .05-.2 Nb .05-.4 .05-.4 .05-.4 Ni .03-.1 &lt;70 ppm .03-.1 Cr total Ni + Cr .03-.1 &lt;200 ppm O 300-1200 ppm 300-700 ppm 300-700 ppm Fe + Ni + Cr &lt;.25 &lt;.25 &lt;.25 Zr Balance* Balance* Balance* ______________________________________ *Zirconium is essentially the balance except for impurities (other than oxygen), which are maintained below 2000 ppm. This inner layer has been provided to give the fuel cladding tube improved resistance to the propagation of PCI related cracks in pile. The alloy selected for this layer (as shown in Table I) contains minimal amounts of tin, iron, niobium and nickel (as noted in the table chromium may be substituted for some or all of the nickel) in order to assure that the aqueous corrosion resistance of the inner layer is at least substantially the same as the corrosion resistance of the Zircaloy outer layer. Upper limits have been provided for these elements to assure that the inner layer material maintains sufficient ductility during in pile usage to stop the propagation of PCI related cracks. At the levels shown in the table the total iron, nickel and chromium contents, as well as their individual values, have been limited to assure that the amount of precipitates formed by these elements is not excessive, thereby minimizing any adverse effects these elements may have on PCI related performance, while providing a sufficient level of precipitates to assure the desired aqueous corrosion resistance. In one of the preferred compositions shown in Table I chromium may completely replace the nickel in the inner layer composition. For applications such as in heavy water reactors, this low nickel composition is preferred since chromium has a significantly lower thermal neutron capture cross section compared to that of nickel. At the levels specified for tin and niobium, these elements in addition to enhancing aqueous corrosion resistance also provide some solid solution strengthening. It is critical that the niobium content be kept below 0.4 wt.% in order to minimize niobium containing precipitates. In order to provide greater assurance in this regard it is preferred that the maximum niobium content be no greater than 0.2 wt.%. Increasing oxygen increases the hardness of the inner layer alloy and is believed to adversely affect the ability of the layer to resist PCI crack propagation in pile. Oxygen is therefore kept below 1200 ppm. Preferably the oxygen content of the inner layer is between about 300 to 1000 ppm, and more preferably between 300 and 700 ppm. The lower limit on oxygen content has been selected on the basis that any further improvement in PCI performance obtained by decreasing the oxygen further is believed to be limited and therefore cannot be justified in view of the significant additional costs involved in reducing the oxygen content further. While it has been noted that the total impurities in the inner layer are maintained below 2000 ppm, it is preferred that it be below 1500 ppm and that individual impurity contents be within the maximum levels specified by ASTM B350-80 Table 1 UNS R60001, where applicable. ASTM B350-80, in its entirety, is hereby incorporated by reference. Electron beam melting of the zirconium starting material to be used in the inner layer alloy, may be performed to reduce total impurity content. The thickness of the inner layer 20 is less than the thickness of the outer layer 10, and is preferably about 0.002 to about 0.006 and more preferably about 0.003 to 0.005 inches. The outer layer 20 forms the bulk of the cladding and provides the cladding with its required mechanical properties. The required thickness of this outer layer may thus be determined by conventional procedures used by those of ordinary skill in the art of nuclear fuel element design. Complete metallurgical bonding between the inner and outer layer is preferably obtained by a combination of hot working, annealing and cold working steps. The invention will be further clarified by the following example which is intended to be purely exemplary of the present invention. Melt an alloy having the nominal composition shown in Table II by consumable electrode vacuum arc melting the required alloying additions with commercially available zirconium. Arc melting is preferably performed at least twice. 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 alloying elements and impurities, and subsequently, at an intermediate stage of manufacture, such as near the co-extrustion stage, for the interstitial elements, oxygen, hydrogen and nitrogen. Chemical analysis of the final size cladding is not required. TABLE II ______________________________________ Nominal Composition of Inner Layer Material ______________________________________ Sn 0.2 wt. % Fe 0.1 wt. % Nb 0.1 wt. % Ni 0.05 wt. % Cr 0.05 wt. % O 300 ppm Zr remainder, with incidental impurities ______________________________________ Fabricate the resulting ingot 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 micro-structure. 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. 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, and in U.S. Pat. No. 4,450,016 which are all hereby incorporated by reference. All of the foregoing fabrication processes will result in complete and continuous metallurgical bonding of the layers, except for minor, insignificant areas of unavoidable bond-line contamination. Beta treatment, either by laser or induction heating, while not required to practice the present invention, is preferred. When used, such treatment would be performed either between the next to last and last cold pilgering passes preferably as a surface treatment (as described in U.S. patent application 343,788) or just prior to the next to last cold pilger pass preferably as a through wall beta treatment. After beta treatment all intermediate, as well as the final anneals, should be performed below about 600.degree. C. and more preferably at or below about 550.degree. C. These low temperature anneals are used to preserve the enhanced corrosion resistance imparted by the beta treatment. Most preferably, the aqueous corrosion resistance of the outer layer and inner layer are characterized by a grey or 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 beta treatment has been used, the final anneal, after the final cold pilgering pass, may be one in which the zirconium alloy inner layer is stress relieved (i.e. without significant recrystallization), partially recrystallized, or fully recrystallized. Where a full recrystallization final anneal is performed, the resulting average equiaxed grain size is no larger than about 1/4, and more preferably between about 1/10 and 1/30, 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, and finishing steps are performed. The lined cladding is loaded with fissile fuel material. Preferably the fuel materials used are in the form of cylindrical pellets and may have chamferred edges and/or concavedly dished ends. Preferably those pellets are composed of UO.sub.2 and are about 95% dense. The uranium in these pellets may be enriched or natural uranium. These pellets may also contain a burnable absorber such as gadolinium oxide or a boron containing compound. The resulting fuel element may be one of any of the known commercial pressurized water, boiling water, or heavy water reactor designs, preferably containing helium within the sealed fuel rod. 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.