Stress-corrosion-crack initiation and growth-rate studies on titanium grade 7 and alloy 22 in concentrated groundwater
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NTRODUCTION
ALLOY 22 (UNS N06022) is the current reference material for the outer barrier in the high-level nuclear waste package for the Yucca Mountain Project. The waste package is an essential element of the engineered barrier system, and the ability to provide very long waste-package lifetimes that can be predicted with confidence is a central factor in the release rate of radionuclides. Stress-corrosion cracking (SCC), along with general and localized corrosion, are the most likely degradation modes for the waste-package materials. While aggressive environments that give rise to pitting and crevice corrosion may also induce SCC, it has proven incorrect to assume that because the material is highly resistant to localized corrosion, it is also stress-corrosion resistant.[1,2] Increasingly careful SCC studies reveal that many materials once thought to be immune do exhibit stress-corrosioncrack growth under constant stress-intensity-factor conditions.[1,2] While no crack initiation and only very slow or no sustained crack growth occurs in Alloy 22 in these studies, the Project approach is to preclude outer surface tensile stresses by utilizing an effective stress-mitigation process.[3] Demonstrating and predicting corrosion damage and associated waste-package lifetimes depends primarily on characterizing the local environment that forms on the waste package. This is particularly important at higher temperatures (above 75 °C), where the heat flux through the waste package is higher, the environments more concentrated, and the material susceptibility to corrosion degradation highest. Because of the radioactive decay heat generated within the waste packages, there is a resulting heat flux across each waste package P.L. ANDRESEN, Senior Scientist, G.M. CATLIN, Lead Professional, and P.W. EMIGH, Technician, are with GE Global Research, Schenectady, NY 12309. Contact e-mail: [email protected] L.M. YOUNG, Scientist, is formerly with GE Global Research. G.M. GORDON is with AREVA (Framatome ANP), Las Vegas, NV 89144. This article is based on a presentation made in the symposium “Effect of Processing on Materials Properties for Nuclear Waste Disposition,” November 10–11, 2003, at the TMS Fall meeting in Chicago, Illinois, under the joint auspices of the TMS Corrosion and Environmental Effects and Nuclear Materials Committees. METALLURGICAL AND MATERIALS TRANSACTIONS A
and adjacent drift wall, which results in the waste package always being somewhat hotter than its surrounding environment. It is reasonable to assume that water reaches the emplacement drift and that the higher surface area of the tunnel walls controls the air temperature and, once the temperature drops sufficiently, maintains relative humidity near 100 pct. Thus, any liquid that forms on the waste package must concentrate sufficiently to account for the temperature differential between the emplacement drift wall and the waste package. Whether from dripping/splashing ground water, contaminants from handling, or rock dust and atmospheric aerosols (during con
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