Patent Number: 052788822
Section: summary

BACKGROUND OF THE INVENTION This invention relates to alloys for use in light water nuclear reactor (LWR) core structural components and fuel cladding. More particularly, this invention relates to a zirconium alloy for such use which exhibits superior ductility, creep strength, and corrosion resistance after irradiation. Still more particularly, this invention relates to a zirconium alloy with improved creep strength, corrosion resistance, and low neutron absorption cross section by controlling its alloy composition to within particular ranges, and especially including oxygen in particularly high ranges, thus to assist in reducing hydrogen uptake of the proposed alloy. DESCRIPTION OF THE PRIOR ART Zirconium alloys are used in the fuel assembly structural components of nuclear reactors, such as in fuel rod cladding, guide or thimble tubes, grid strips, instrument tubes, and so forth because of their low neutron cross section, good corrosion resistance in high pressure/high temperature steam and water, good mechanical strength and fabricability. Zirconium alloys, particularly those commonly known as Zircaloy-2 and Zircaloy-4 have been used in light water reactor cores because of their relatively small capture cross section for thermal neutrons. The addition of 0.5 to 2.0 percent by weight niobium and up to 0.25 percent of a third alloying element to these zirconium alloys for purposes of corrosion resistance in the reactor core is suggested in U.S. Pat. No. 4,649,023 as part of a teaching of producing a microstructure of homogeneously dispersed fine precipitates of less than about 800 angstroms. The third alloying element is a constituent such as iron, chromium, molybdenum, vanadium, copper, nickel and tungsten. Pellet-clad interaction (PCI) resistance is sought in U.S. Pat. Nos. 4,675,153 and 4,664,831 by use of zirconium-based alloys including "zirconium-2.5 w/o niobium". The latter teaching also refers to "Zr-Nb alloys containing about 1.0 to 3.0 w/o Nb". In these patents, oxygen is present "below about 350 ppm of said alloy". U.S. Pat. No. 4,648,912 teaches improving high temperature corrosion resistance of an alpha zirconium alloy body by rapidly scanning the surface of the body with a laser beam. The alloy treated included zirconium-niobium alloys. Thus, it has been found by various investigators in the prior art literature that the addition of niobium to a zirconium alloy for use in light water reactors will reduce hydrogen uptake from waterside corrosion, stabilize alloying element and oxygen-irradiation defect complexes, and make the alloy more resistant to annealing of irradiation damage. It is also reported by investigators that niobium will enhance work hardenability of irradiated Zircaloy but that an addition of niobium above the 1 percent level will not result in further additional benefit in mechanical properties. An improved ductile irradiated zirconium alloy is described in U.S. Pat. No. 4,879,093 issued to an inventor in this application. The alloy has a stabilized microstructure which minimizes loss of alloy ductility required to resist release of fission gases and to handle spent fuel safely. The alloy retains a reasonable corrosion resistance in both pressurized water reactors (PWR) and boiling water reactors (BWR) because of its optimum intermetallic precipitate average particle size. The alloy of the '093 patent is based on an alpha phase zirconium-tin-niobium or alpha phase zirconium-tin-molybdenum alloy having characteristics as shown in Table 1 of that patent with niobium, if present, in a range of from a measurable amount up to 0.6 percent by weight. The molybdenum, if present, is in a range of from a measurable amount up to 0.1 percent by weight. The zirconium-tin system is known as "Zircaloy" and, typically, if Zircaloy-4, for example, would also have 0.18 to 0.24 percent by weight iron, 0.07 to 0.13 percent by weight chromium, oxygen in the range of from 1000 to 1600 ppm, 1.2 to 1.7 percent by weight tin, and the remainder zirconium. U.S. Pat. No. 4,992,240 discloses another zirconium alloy containing on a weight basis, 0.4 to 1.2% tin, 0.2 to 0.4% iron, 0.1 to 0.6% chromium, not higher than 0.5% of niobium, and balance zirconium, wherein the sum weight proportions of tin, iron and chromium is in the range of 0.9 to 1.5%. Oxygen, according to FIG. 4 of the '240 patent, is about 1770 ppm to 1840 ppm. Niobium is apparently optional, and silicon is not reported. Recent trends in the nuclear industry include shifts toward higher coolant temperatures to increase the thermal efficiency and toward higher fuel discharge burnups to increase the fuel utilization. Both the higher coolant temperatures and discharge burnups tend to increase the in-reactor corrosion and hydrogen uptake of the zirconium alloys. The high levels of neutron fluence and simultaneous hydrogen pickup degrade the ductility of zirconium alloys. For these more demanding service conditions, it is therefore necessary to improve the corrosion resistance and irradiated ductility of zirconium alloys. Accordingly, it is a continuing problem in this art to develop a zirconium alloy having superior ductility after irradiation; good corrosion resistance, especially independent of processing history; reduced hydrogen absorption by the alloy; and a significant solid solution alloy strength. It is another continuing general problem in this art to improve the corrosion resistance and irradiated ductility of zirconium alloys used in fuel assembly structural components in nuclear reactors. It is another continuing general problem in this art to provide a zirconium alloy which has superior creep resistance, superior corrosion resistance, and low neutron absorption cross section by the selection of alloying elements in particular ranges. It is another continuing general problem in this art to provide a zirconium alloy with selected alloying elements to assist in reducing hydrogen uptake of the alloy.