Metal treatment – Stock – Titanium – zirconium – or hafnium base
Reexamination Certificate
2000-08-15
2002-03-12
Sheehan, John (Department: 1742)
Metal treatment
Stock
Titanium, zirconium, or hafnium base
C420S422000
Reexamination Certificate
active
06355118
ABSTRACT:
FIELD OF THE INVENTION
The invention relates to a metallurgical process involving zirconium alloys, and more particularly to a process for treating zirconium alloys to immunize and improve resistance of such alloys to nodular corrosion when exposed to high pressure steam.
BACKGROUND OF THE INVENTION
Nuclear reactors utilize water/steam as a coolant for the reactor as well as a source of energy to power steam turbines to thereby provide electrical energy. Nuclear reactors typically have their nuclear fissionable material contained in sealed cladding tubes, generally of a zirconium alloy, for isolation of the nuclear fuel from the water/steam. Zirconium and its alloys are widely used as nuclear fuel cladding since they advantageously possess low neutron absorption cross-sections, and at temperatures below about 398 C (the approximate core temperature of an operating nuclear reactor), are non-reactive and importantly possess high corrosion resistance relative to other metal alloys in the presence of de-mineralized water or steam. Two widely used zirconium alloys (“Zircaloys”) are “Zircaloy-2” and “Zircaloy-4”, trade names of Westinghouse Electric Corporation for zirconium alloys of the above chemical compositions. Zircaloy-2, a Zr—Sn—Ni—Fe—Cr alloy, is generally comprised (by weight) of approximately 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium and 0.05-0.08% nickel. Zircaloy-4 has essentially no nickel, and about 0.2% iron, but is otherwise substantially similar to Zircaloy-2. Zircaloy-2 has enjoyed widespread use and continues to be used at present in nuclear reactors. Zircaloy-4 was developed as an improvement to Zircaloy-2 to reduce problems with hydriding, which causes Zircaloy-2 to become brittle when cooled to ambient temperatures (ie. when the reactor is shut down) after absorbing hydrogen at higher temperatures.
Zirconium alloys are among the best corrosion resistant materials when exposed to steam at reactor operating temperatures (less than 398 C, typically 290 C) in the absence of radiation from nuclear fission reactors. The corrosion rate in absence of neutron bombardment is very low and the corrosion product is a uniform, black ZrO
2
oxide film/layer which forms on exterior surfaces of Zircaloy exposed to high temperature steam (uniform corrosion). The black oxide layer of ZrO
2
usually contains a small (non-stoichiometric) excess of zirconium, and as such, it contains excess electrons giving it a black or gray color. It is also highly adherent to zirconium or Zircaloy surfaces exposed to steam.
Despite such relatively high corrosion resistance, when Zircaloys are used as cladding and exposed to high neutron flux in nuclear reactors, corrosion rates are generally increased, and cladding corrosion does become a potential problem in Pressurized Water Reactors (PWR's) and particularly Boiling Water Reactors (BWR's), where corrosion occurs in two formats, namely increased uniform corrosion as mentioned above, and alternatively, a second form, namely, nodular corrosion. Nodular corrosion is a highly undesirable, white, stoichiometric ZrO
2
oxide layer (“bloom”) which forms on the surface of the cladding. It tends to form as small patches (“nodules” or “pustules”) on the surface of Zircaloys. Today, it is increasingly common to operate nuclear reactors at high “burn-up” (ie. to nearly complete consumption of the nuclear fuel). Under these conditions, the cladding is exposed to neutron flux for longer periods, which generally tends to increase the severity of nodular corrosion. Such increased nodular corrosion not only shortens the service life of the tube cladding (since when concentrated nodular corrosion acts in conjunction with certain contaminants—such as copper ions—localized spalling and ultimately penetration of the cladding can occur), but also produces a detrimental effect on the efficient operation of the reactor. In particular, the white ZrO
2
, being less adherent than black ZrO
2
, is prone to spalling or flaking away from the tube and entering into the reactor water, with detrimental effects. On the other hand, if the white nodular corrosion product does not spall away but remains on the tubing, a decrease in rapidity of heat transfer through the Zircaloy tube into the water cooling medium occurs when the less-dense white ZrO
2
oxide layer covers an increasingly large portion of the Zircaloy tube exterior surface, and the reactor becomes less thermally efficient. Thus, nodular corrosion can become a significant problem for Zircaloy cladding in situations where Zircaloy tube cladding is left in the nuclear reactor for longer periods in conditions of high “burn-up”.
Zircaloys used in cladding for nuclear fuel rods are generally subject during their manufacture to a variety of heat treatments and anneals during the formation of the tubular cladding. It is known that the various heat treatments and quenching procedures used in forming a Zircaloy billet, and the various anneals and cold-working thereafter to form the Zircaloy tube cladding, all have an effect on the particular Zircaloy tubing's ability to resist nodular corrosion, with some Zircaloys able to withstand nodular corrosion better than others despite both being of identical chemical composition. For example, fine grained equiaxed &agr; Zircaloy-2, heated to 1010 C and slow-cooled at a rate of 18 C/hr. to 600 C and thereafter quenched, exhibits a high susceptibility to nodular corrosion under the standard steam test (510 C, 1500 psig, 24 hr.). Paradoxically, the same material, if simply quenched from 1010 C, or if heated to only 950 C and cooled at the same rate of 18 C/hr. to 600 C and thereafter quenched, exhibits high resistance to corrosion under the same standard steam test.
The actual physical changes in the structural properties of zirconium alloys during manufacturing processes of nuclear fuel tubing made therefrom were little understood, and it was therefore, prior to this invention, difficult to conceive of the best ways to immunize such fuel tubing to nodular corrosion. U.S. application Ser. No. 09/050,214 by the same inventor, filed Mar. 30, 1998 entitled “Method for Determining Corrosion Susceptibility of Nuclear Fuel Cladding to Nodular Corrosion”, the subject matter of which is herein incorporated by reference, discloses that &agr; Zircaloy-2 with very small precipitates, formed by having been heated to 1010 C and quenched, exhibits high resistance to nodular corrosion. Unfortunately, some research has suggested that small precipitates in the Zircaloy metal matrix can increase the danger of crack propagation in the cladding axial direction [see for example, U.S. patent application Ser. No. 08/052,793 entitled “Zircaloy Tubing Having High Resistance to Crack Propagation” (now U.S. Pat. No. 5,519,748), and U.S. patent application Ser. No. 08/052,791 entitled “Method of Fabricating Zircaloy Tubing Having High Resistance to Crack Propagation” (now U.S. Pat. No. 5,437,747), both assigned to the assignee hereof]. Thus, while zirconium alloy tubing possessing excellent resistance to nodular corrosion may be manufactured, it is frequently necessary to add further annealing heat treatments to achieve other further objectives, such as to reduce the incidence of axial splitting of a Zircaloy-2 tubing. Unfortunately, up until the present invention and the understanding of the concept of critical temperature T
c
disclosed in U.S. application Ser. No. 09/050,214, filed Mar. 30, 1998, entitled “Method for Determining Corrosion Susceptibility of Nuclear Fuel Cladding to Nodular Corrosion”, such other anneal processes often had detrimental effects on the ability of such zirconium alloy tubing to withstand nodular corrosion. In fact, until the present invention, it was little understood why some annealing processes actually have the effect of sensitizing the tubing to nodular corrosion. It was thus unknown, prior to this invention, how to reliably retain the benefits of a zirconium alloy possessing high resistance to nodular corrosion when further subjecting such tubin
General Electric Company
Johnson Noreen C.
Santandrea Robert P.
Sheehan John
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