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The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

Description: Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pellet contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress effect. Kim et ...
Date: June 21, 2001
Creator: Kammenzind, B.F., Eklund, K.L. and Bajaj, R.
Partner: UNT Libraries Government Documents Department

Effects of neutron irradiation on deformation behavior of nickel-base fastener alloys

Description: This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x10{sup 20} n/cm{sup 2} at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranular failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed.
Date: July 1, 1999
Creator: Bajaj, R.; Mills, W.J.; Kammenzind, B.F. & Burke, M.G.
Partner: UNT Libraries Government Documents Department

The corrosion behavior of hafnium in high-temperature-water environments

Description: The high-temperature-water corrosion performance of hafnium is evaluated. Corrosion kinetic data are used to develop correlations that are a function of time and temperature. The evaluation is based on corrosion tests conducted in out-of-pile autoclaves and in out-of-flux locations of the Advanced Test Reactor (ATR) at temperatures ranging from 288 to 360 C. Similar to the corrosion behavior of unalloyed zirconium, the high-temperature-water corrosion response of hafnium exhibits three corrosion regimes: pretransition, posttransition, and spalling. In the pretransition regime, cubic corrosion kinetics are exhibited, whereas in the posttransition regime, linear corrosion kinetics are exhibited. Because of the scatter in the spalling regime data, it is not reasonable to use a best fit of the data to describe spalling regime corrosion. Data also show that neutron irradiation does not alter the corrosion performance of hafnium. Finally, the data illustrate that the corrosion rate of hafnium is significantly less than that of Zircaloy-2 and Zircaloy-4.
Date: October 1, 1999
Creator: Rishel, D.M.; Smee, J.D. & Kammenzind, B.F.
Partner: UNT Libraries Government Documents Department

Hydrogen pickup and redistribution in alpha-annealed Zircaloy-4

Description: Zircaloy-4, which is widely used as a core structural material in Pressurized-Water Reactors (PWR), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and hydrides precipitate after the Zircaloy-4 matrix becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4. To study hydrogen pickup and concentration, a postirradiation nondestructive radiographic technique for measuring hydrogen concentration was developed and qualified. Experiments on hydrogen pickup were conducted in the Advanced Test Reactor (ATR). Ex-reactor tests were conducted to determine the conditions for which hydrogen would dissolve, migrate, and precipitate. Finally, a phenomenological model for hydrogen diffusion was indexed to the data. This presentation describes the equipment and the model, presents the results of experiments, and compares the model predictions to experimental results.
Date: June 1, 1996
Creator: Kammenzind, B.F.; Franklin, D.G.; Duffin, W.J. & Peters, H.R.
Partner: UNT Libraries Government Documents Department

The long range migration of hydrogen through Zircaloy in response to tensile and compressive stress gradients

Description: Zircaloy-4, which is used widely as a core structural material in pressurized water reactors (PWRs), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and zirconium hydride phases precipitate after the Zircaloy-4 lattice becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4, degrading its mechanical performance as a structural material. Because hydrogen can move rapidly through the Zircaloy-4 lattice, the potential exists for large concentrations of hydride to accumulate in local regions of a Zircaloy component remote from its point of entry into the component. Much has been reported in the literature regarding the long range migration of hydrogen through Zircaloy under concentration gradients and temperature gradients. Relatively little has been reported, however, regarding the long range migration of hydrogen under stress gradients. This paper presents experimental results regarding the long range migration of hydrogen through Zircaloy in response to both tensile and compressive stress gradients. The importance of this driving force for hydrogen migration relative to concentration and thermal gradients is discussed.
Date: November 1, 1998
Creator: Kammenzind, B.F.; Berquist, B.M.; Bajaj, R.; Kreyns, P.H. & Franklin, D.G.
Partner: UNT Libraries Government Documents Department

Lifetime embrittlement of reactor core materials

Description: Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10{sup 24} n/M{sup 2} (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K{sub IC} due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K{sub IC} plus its lower hydrogen absorption characteristics compared with Zircaloy-2.
Date: August 1, 1994
Creator: Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G. & White, C.J.
Partner: UNT Libraries Government Documents Department