Two corrosion resistant nickel-based alloys, 625 and C22, have been selected by the U.S. Department of Energy as candidate materials for the inner container of high-level radioactive waste packages. The susceptibility of these materials to localized corrosion was evaluated by measuring the repassivation potential as a function of solution chloride concentration and temperature using cyclic potentiodynamic polarization and lead-in-pencil potential step test methods. At intermediate Cl-concentrations, e.g., 0.028- 0.4 M, the repassivation potential of alloy 625 is greater than that for alloy 825 and is dependent on the Cl- concentration. However, at higher concentrations, the repassivation potential is slightly less than that for alloy 825 and is weakly dependent on Cl-concentration. The repassivation potentials for alloy C-22 under all test conditions are considerably higher than those of either alloy 625 or 825 and are in the range where oxygen evolution is expected to occur.
The U.S. Department of Energy (DOE) is responsible for sitting, constructing and operating a repository for the permanent disposal of high-level nuclear waste (HLW). Several HLW container designs have been considered by DOE over the history of the Yucca Mountain repository programl. The current reference case design consists of two metallic containers, or barriers, one placed inside the other. The primary candidate for the outer barrier is a corrosion-allowance materials, ASTM A5 16 Grade 55 steel. The inner barrier will be fabricated from one of several nickel-based corrosion-resistant materials including alloys 825,625, and C22. More recently, however, alloy 625 has been proposed as the primary material for the inner barrier, replacing alloy 825, DOE?s former primary candidate. The radioactive decay heat generated by the HLW and other factors such as radiation, the presence of man-made materials, etc., are expected to modify the composition of the groundwater that would eventually contact the waste packages upon cooling after a period of hundreds to thousands of years. Degradation of the outer barrier is expected to occur primarily by aqueous corrosion in the presence of modified groundwater on the surface of the container. Once the outer barrier is breached, the inner barrier is expected to be affected by localized corrosion.
The U.S. Nuclear Regulatory Commission has been conducting confirmatory research to develop the technical bases required for judging the adequacy of the HLW container design and material selection. As part of this research program, numerous experimental investigations have been conducted to address the effect of the neat-field repository environment on the localized corrosion of inner container materials. In previous work, Sridhar et al.3-9 have used electrochemical techniques to examine the effects of various water chemistries and material conditions on critical potentials for localized corrosion. Their work ultimately showed that the repassivation potential can be used to conservatively estimate the lower bounds of localized corrosion of candidate container materials. The objectives of the current investigation were to determine and evaluate the critical potentials for localized corrosion of alloys 625 and C22 in simulated repository environments. Two electrochemical techniques, cyclic potentiodynamic polarization (CPP) and lead-in-pencil (LIP) potential step test, were used with solutions containing a wide range of chloride concentrations at various temperatures. Test solutions containing chloride ion concentrations up to 4 M were employed to simulate the effects of a concentrated salt solution on the waste package surface that is expected to evolve over time as a res