Patent Number: 
Section: description

Referring now more particularly to FIG. 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This fuel assembly 10 consists of a tubular flow channel 11 of generally square cross-section provided in its upper end with a lifting bale 12, and at its lower end, with a nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at outlet 13 and the lower end of the nose piece is provided with cooling flow openings. An array of fuel elements or rods 14 is enclosed in the channel 11 and supported therein by means of an upper end plate 15 and a lower end plate (not shown due to the lower portion being omitted). The liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges through the upper outlet 13 at an elevated temperature in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized water reactors. The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the composite cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material and accumulation of gases released from the fuel material. A nuclear fuel material retainer means 24 in the form of a helical member if positioned within space 20 to provide restraint against the axial movement of the pellet column, especially during handling and transportation of the fuel element. The fuel element is designed to provide excellent thermal contact between the cladding and the fuel material, a minimum amount of parasitic neutron absorption, and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity. A nuclear fuel element or rod 14 constructed according to the teachings of this invention is shown in a partial section in FIG. 1. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a plurality of fuel pellets or fissionable and/or fertile materials positioned within a structural composite cladding of container 17. In some cases, the fuel pellets may be of various shapes, such as cylindrical pellets or spheres. In other cases, different fuel forms, such as particulate fuel, may be used. The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. The preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide. Referring now to FIG. 2, the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by composite cladding 17, which, in this invention, is also referred to as a composite cladding container, comprised of an outer metallic tubular portion 21 and an inner metallic portion 22. The composite cladding container, generally elongated in shape, encloses the fissile core. A gap 23 is optional between the core and the composite cladding 17 and may or may not be present. The composite cladding container has an external substrate or outer metallic tubular portion 21 selected from conventional cladding materials such as stainless steel and zirconium allays. In the preferred embodiment of this invention, the outer metallic tubular portion 21 is a zirconium alloy such as Zircaloy-2, currently the preferred alloy in the outer portion, Zircaloy-4 and other zirconium alloy improvements. Metallurgically bonded to the outer metallic tubular portion of the container is an inner metallic barrier. This inner metallic barrier is positioned to prevent contact between the nuclear fuel and the outer metallic tubular portion. The inner metallic barrier is comprised of commercially pure zirconium microalloyed with iron (Fe). The amount of Fe is carefully controlled so as not to be under a lower value of 850 parts per million (ppm) and not to exceed an upper value of 2500 ppm. As set forth in U.S. Pat. No. 4,200,492, production of commercially pure zirconium having trace impurities is well-known. The normal range of these impurities include aluminum (Al) in the amounts of 75 ppm or less; boron (B) in the amounts of 0.4 ppm or less; cadmium (Cd) in the amounts of 0.4 ppm or less; carbon (C) in the amounts of 270 ppm or less; chromium (Cr) in the amounts of 200 ppm or less; cobalt (Co) in the amounts of 20 ppm or less; copper (Cu) in the amounts of 50 ppm or less; hafnium (Hf) in the amounts of 100 ppm or less; hydrogen (H) in the amounts of 25 ppm or less; magnesium (Mg) in the amounts of 20 ppm or less; manganese (Mn) in the amounts of 50 ppm or less; molybdenum (Mo) in the amounts of 50 ppm or less; nickel (Ni) in the amounts of 70 ppm or less; niobium (Nb) in the amounts of 100 ppm or less, nitrogen (N) in the amounts of 80 ppm or less; tungsten (W) in the amounts of 100 ppm or less; silicon (Si) in the amounts of 120 ppm or less; tin (Sn) in the amounts of 50 ppm or less; titanium (Ti) in the amounts of 50 ppm or less and uranium in the amounts of 3.5 ppm or less. The prior art practice has treated Fe as a trace element which may be present in the amounts of 1500 ppm or less. The discovery of this invention has been that by controlling the amount of Fe as a microalloyed addition to the commercially pure zirconium, an inner metallic barrier having a beneficial balance between stress corrosion crack resistance and corrosion resistance resulting in an improved alloy can be produced. While iron has been present as a trace or tramp element in the past, the amount of iron has only been controlled as a maximum permissible amount, so that erratic results in performance occurred because there was no appreciation of controlling the composition of the Fe within the limits set forth by the present invention to achieve the improved beneficial balance between stress corrosion crack resistance and corrosion resistance. It has been discovered that the corrosion resistance and the stress corrosion crack resistance of commercially pure zirconium can be balanced by micro alloying it with Fe in the amounts from about 850 ppm to about 2500 ppm. The incidental impurities as set forth above can remain at the levels as previously set forth without adversely affecting the beneficial aspects of the present invention. Referring now to FIGS. 3 and 4, the corrosion resistance of Zr is related to the Fe content present in the Zr. FIG. 3 is a graph of the corrosion resistance of zirconium having varying amounts of Fe content as measured by weight gain over a period of time in 400xc2x0 C. (750xc2x0 F.) steam as set forth by a modified test based on the ASTM G2 Steam Test. The modification consists of extending the time of the ASTM G2 Stream Test. FIG. 4 is a graph showing the effect of increasing iron content on the corrosion rate of Zr in 400xc2x0 C. steam. Clearly, there is a correlation between Fe content and the corrosion of Zr in 400xc2x0 C. steam. Between about 100 and 400 ppm Fe, the corrosion rate of the Zr drops from about 35 mg/dm2/day to about 15 mg/dm2/day. At about 800 ppm to about 850 ppm Fe the corrosion rate is significantly reduced, exhibiting a decay approaching 0 asymptotically. As indicated by FIG. 3, small increases in iron from about 360 ppm to 940 ppm significantly reduce the corrosion of the Zr. Above about 940 ppm, although increasing the Fe content improves the corrosion resistance of the Zr slightly, the improvement is a diminishing function of increased Fe. It appears that the saturation in corrosion rate improvement occurs at about 1500 ppm. Thus, it is important to maintain the Fe levels in the Zr above the minimum amounts of about 850 ppm and preferably above about 1000 ppm and most preferably above about 1500 ppm in order to take advantage of the corrosion resistance of the microalloyed Fe. As the amount of microalloyed Fe in Zr drops below about 850 ppm, the corrosion resistance of the inner barrier begins to deteriorate dramatically. Thus, it can be seen that the life of an inner barrier having less than the critical amount of Fe will be shortened, resulting in a failure. Referring now to FIG. 5, the resistance of Zr to stress corrosion cracking is also related to the Fe content of the Zr. Resistance to stress corrosion cracking is measured by a test using an expanding mandrel in an iodine environment. The test simulates the ability of a fuel rod to resist the in-reactor failure mechanism called pellet cladding interaction (PCI) and is referred to as PCI resistance testing. The test is performed using a ramp and hold expanding mandrel, with a 4% strain rate at 315xc2x0 C. (662xc2x0 F.). There is no standard industry test, although various tests are used in the industry. The Assignee of the present invention has found that the expanding mandrel test as set forth above, can discriminate between and among varying stress corrosion cracking susceptibility of alloys. When the stress-strain conditions generated by various industry testing techniques reflect actual fuel rod conditions, similar discrimination should result from such tests. FIG. 5 and the data of Table 1 indicate that at and below about 1000 ppm Fe the PCI resistance is perfect, which is to say that no tests produced failure. At about 1500 ppm Fe, there is a transition to decreasing PCI resistance. At about 3000 ppm Fe, about 60% of the tests produced failure, which is an unsatisfactory result. From these two tests, it can be seen that even though cracking resistance is excellent below about 850 ppm Fe, the corrosion resistance at this Fe level is unacceptable. At or above 850 ppm Fe, the corrosion level is acceptable. At about 1500 ppm Fe, the crack resistance begins to deteriorate, trending towards unacceptable levels, even though the corrosion resistance is excellent. Thus, even though there may be some small amount of cracking at about 2000 ppm Fe, it is tolerable since it is not accompanied by corrosion, which would provide a combined failure mechanism, exacerbating the failure of the barrier protection. The inner metallic barrier can have a composition of iron above about 1500 ppm and below about 2000 ppm microalloyed with commercially pure zirconium. Crack resistance continues to deteriorate and it can be seen from the graphs of FIGS. 5 and 6 that it becomes unacceptable in the range of Fe between about 2500 and 3000 ppm. FIG. 6 is a graph that depicts both the corrosion rate and the PCI resistance as a function of the Fe content of the barrier. It can be seen from this graph that Fe in a range of from about 1000 to about 2000 ppm produces both acceptable corrosion rates and tolerable PCI resistance. Referring to Table 1 and FIGS. 3-6, since a slight increase in PCI cracking can be tolerated, another embodiment of the invention will include Fe in the range of about 850-2000 ppm microalloyed with commercially pure Zr. In a preferred embodiment of the invention, Fe is included in the range of about 1000-2000 ppm. The most preferred embodiment includes 1000xc2x1150 ppm Fe-microalloyed with Zr. The Zr microalloyed with Fe will retain the ductility that is desirable in an inner metallic barrier. Because the Fe is controlled in microalloyed amounts, it retains its low neutron absorptivity and good heat transfer characteristics yet is highly resistant to radiation hardening. Ideally, the Zirconium microalloyed with Fe is metallurgically bonded to the outer metallic tubular portion 21, preferably Zircaloy-2, but is recrystallized having a grain size in the range of ASTM 9 to ASTM 12. The inner metallic barrier 22 typically and preferably comprises between about 10% to about 20% of the total thickness of the composite cladding 17. However, the thickness of inner metallic barrier 22 can be varied outside this prescribed thickness range as long as neither the integrity of the outer metallic tubular portion 21 nor the ability of the inner metallic portion 22 to function as a barrier is adversely affected. The composite cladding 17 can be manufactured by any of the well-known prior art methods for bonding an inner barrier of zirconium sponge to an outer barrier of, for example, stainless steel or Zircaloy. Although the present invention has been described in connection with specific examples and embodiments, those skilled in the art will recognize that the present invention is capable of other variations and modifications within its scope. These examples and embodiments are intended as typical of, rather than in any way limiting on, the scope of the present invention as presented in the appended claims.