Gustavo A. Cragnolino

An Electrochemical Approach to Predicting Long-Term Localized Corrosion of Corrosion-Resistant High-Level Waste Container Materials

Abstract Lifetime predictions of containers used for the disposal of high-level radioactive waste (HLW) are necessary to determine the overall performance of a HLW repository. Failure of the contai...

Applicability of Repassivation Potential for Long-Term Prediction of Localized Corrosion of Alloy 825 and Type 316L Stainless Steel

Abstract Repassivation potential (Erp) was investigated for use as a parameter in the long-term prediction of pitting resistance of two Fe-Ni-Cr-Mo alloys that are candidate materials for high-leve...

Localized Corrosion Susceptibility of Alloy 22 in Chloride Solutions: Part 1—Mill-Annealed Condition

Abstract Alloy 22 (UNS N06022) is a Ni-Cr-Mo-W alloy developed for corrosion resistance in a variety of aggressive environments. Because of its excellent corrosion resistance, Alloy 22 was selected as a candidate container material for the disposal of high-level nuclear waste at the proposed repository at Yucca Mountain, Nevada. The localized corrosion susceptibility of mill-annealed Alloy 22 was assessed in chloride-containing solutions at temperatures ranging from 60°C to 150°C by comparing the crevice corrosion repassivation potentials with corrosion potentials measured in separate tests. Crevice corrosion repassivation potentials were found to be strongly dependent on temperature, chloride concentration, and nitrate-to-chloride molar concentration ratio. Corrosion potentials were dependent on solution pH but independent of chloride concentration.

Localized Corrosion Susceptibility of Alloy 22 in Chloride Solutions: Part 2—Effect of Fabrication Processes

Abstract The effect of fabrication processes on the localized corrosion susceptibility of Alloy 22 ([UNS N06022] 56Ni-22Cr-13.5Mo-3W-4Fe), a candidate container material for the disposal of high-level nuclear waste in the proposed repository at Yucca Mountain, Nevada, was assessed in chloride-containing solutions at temperatures ranging from 60°C to 95°C. The results of tests covering a range of metallurgical conditions, including as-welded, welded plus solution-annealed, and thermally aged materials, were compared with those for the as-received mill-annealed alloy. The localized corrosion susceptibility of Alloy 22 was determined by comparing the crevice corrosion repassivation potentials with corrosion potentials measured in separate tests. The crevice corrosion susceptibility was found to be dependent on temperature, chloride concentration, and the nitrate-to-chloride molar concentration ratio. Welding and short-term thermal aging decreased the localized corrosion resistance of Alloy 22 compared with t...

Long-term prediction of localized corrosion of alloy 825 in high-level nuclear waste repository environments

Abstract Long-term prediction of localized corrosion of high-level nuclear waste container materials is a necessary step in the performance assessment of the engineered barrier system. Localized corrosion of corrosion-resistant materials may occur if the containers are exposed to chloride (Cl−)-containing ground water at elevated temperatures. Potentiostatic tests conducted on alloy 825 (UNS N08825), a candidate container material, have shown that the potential at which localized corrosion can be initiated is a function of surface conditions and exposure time. The initiation potentials for localized corrosion measured in short-term tests with polished specimens overestimated the long-term initiation potentials by as much as 300 mV. On the other hand, the repassivation potential obtained from short-term tests on deep pits can be used to predict the initiation of localized corrosion in long-term tests.

Effect of Simulated Groundwater Chemistry on Stress Corrosion Cracking of Alloy 22

Abstract Slow strain rate tests were performed to evaluate the effect of simulated groundwater chemistry variations on the stress corrosion cracking (SCC) susceptibility of mill-annealed Alloy 22 (...

Localized Corrosion Susceptibility of Alloy 22 as a Waste Package Container Material

Alloy 22 is the material preferred by the U.S. Department of Energy for the waste package outer container for geological disposal of high-level radioactive waste at the proposed repository site in Yucca Mountain, Nevada. The susceptibility of Alloy 22 to localized corrosion is an important consideration in the evaluation of the waste package behavior and the assessment of the overall performance of the proposed repository. The effects of the environment chemical composition and temperature on localized corrosion susceptibility were examined by measuring the repassivation potential for crevice corrosion in chloride-containing solutions at temperatures ranging from 80 to 150°C. The effect of potentially inhibiting anionic species, such as nitrate, was also determined. In addition to the mill annealed material, tests were conducted on both welded and thermally aged material to evaluate microstructural effects related to container fabrication processes. The resistance of Alloy 22 to localized corrosion decreased with increasing temperature and chloride concentration. Welding and thermal aging also decreased the localized corrosion resistance of the alloy.

Effects of Environment on Localized Corrosion of Copper-Based, High-Level Waste Container Materials

Abstract Effects of environmental factors on localized corrosion of copper (Cu)-based materials that are candidates for use in high-level nuclear waste containers were examined. Effects of bicarbonate (HCO3−), chloride (Cl−), and sulfate (SO42−), which are present in the ground water near the proposed Yucca Mountain site for a high-level waste repository, were studied. Localized corrosion was observed only at low temperatures in environments involving combinations of high HCO3− with high Cl− or high SO42−. Uniform corrosion was observed at low concentrations of HCO3− ( 80°C, irrespective of the Cl− and SO42− concentrations. Localized corrosion observed in the Cu-based alloys was much shallower than in stainless steels and nickel (Ni)-based alloys and tended to broaden with time under potentiostatic conditions. Oxygen-free Cu (CDA-102...

Environmental Factors in the Stress Corrosion Cracking of Type 316L Stainless Steel and Alloy 825 in Chloride Solutions

Abstract This paper describes the experimental studies conducted to date to investigate the stress corrosion cracking (SCC) of candidate container materials for the proposed high-level nuclear waste repository at Yucca Mountain, Nevada. The effects of environmental variables, such as chloride (Cl−) concentration, the addition of thiosulfate (S2O32−), and temperature on the SCC susceptibility of type 316L (UNS S31603) stainless steel (SS) and alloy 825 (UNS N08825, Ni-29% Fe-22% Cr-3.0% Mo-2.0% Cu-1.0% Ti) were studied at temperatures from 95°C to 120°C. Results of slow strain rate tests (SSRT) at various potentials were compared to those obtained under constant deflection conditions using U-bend specimens to determine the existence of a critical potential for SCC. While not conclusive, results generated thus far have been consistent with the hypothesis that the repassivation potential (Erp) for localized corrosion is also the critical potential for SCC in these environments. It was confirmed that alloy 82...

An approach to the assessment of high-level radioactive waste containment — II: radionuclide releases from an engineered barrier system

Abstract This paper is the second in a series describing the models used in the Engineered Barrier System Performance Assessment Code (EBSPAC) to represent processes that govern the failure of waste packages (WPs) and the release of radionuclides from the engineered barrier system (EBS). These models are specifically adapted to the US Department of Energy (DOE) WP design, adopted in 1996, for the proposed high-level radioactive waste (HLW) repository at Yucca Mountain (YM). The design consists of a double-wall overpack composed of two concentric containers of different metallic materials in a horizontal drift emplacement. EBSPAC was developed to deterministically evaluate the performance of the engineered barriers and to be used as the source term module in the Center for Nuclear Waste Regulatory Analyses (CNWRA)/Nuclear Regulatory Commission (NRC) Total-system Performance Assessment (TPA) code. EBSPAC has two distinct parts. The part dealing with the radionuclide release subsequent to WP failure is the focus of this paper in which various models (i.e. dry-air oxidation and aqueous dissolution of spent fuel (SF), gaseous and aqueous release of radionuclides) are presented, whereas modeling of the WP failure is described in a companion paper. An example problem is presented to illustrate computational results obtained with the code analyzing the influence of several critical input parameters for the source term related to the repository and EBS designs and resulting environmental conditions. The source term calculations are confined to the radionuclides being released just outside of the WP. Both gaseous and aqueous release calculations are performed using models in which radionuclide decay, in-growth of daughter products in the chains, degradation process of SF, temporal variation of inventory in the WP, and spatial variations in the properties of the surrounding material are included. The degree of complexity varies from model to model as necessary simplifications are made, while ensuring conservatism.