Publications

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This report documents the mockup specifications and manufacturing processes; the initial cutting of the mockup into three cylindrical pieces for testing and the measured strain changes that occurred during the cutting process; and the planned weld residual stress characterization activities and the status of those activities.

Long-term corrosion experiments have been performed on Alloy 22 (UNS N06022), in a series of heated brines formulated to represent evaporatively concentrated ground water, to evaluate the long-term corrosion performance of the material. These solutions included 0.5 M NaCl, in addition to two simulated concentrated ground water solutions. Under conditions where Alloy 22 was anticipated to be passive, the corrosion rate was found to be vanishingly small (i.e., below the resolution of the weight-loss technique used to quantify corrosion in this study). However, under low pH conditions where Alloy 22 was anticipated to be active, or more specifically, where the chromium oxide passive film was not thermodynamically stable, the corrosion rate was appreciable. Furthermore, under such conditions the corrosion rate was observed to be a strong function of temperature, with an activation energy of 72.9±1.8 kJ/mol. Time of Flight-Secondary Ion Mass Spectroscopy analysis of the oxide layer revealed that, while sulfur was present within the oxide for all test conditions, no accumulation was observed at or near the metal/oxide interface. These observations confirm that inhibition of passive film formation via sulfur accumulation does not occur during the corrosion of Alloy 22.

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This study has evaluated the technical feasibility of direct disposal in a geologic repository, of commercial spent nuclear fuel (SNF) in dual-purpose canisters (DPCs) of existing designs. The authors, representing several national laboratories, considered waste isolation safety, engineering feasibility, thermal management, and postclosure criticality control. The 3-year study concludes that direct disposal is technically feasible for most DPCs, depending on the repository host geology. Postclosure criticality control, and thermal management strategies that allow permanent disposal within 150 years, are two of the most challenging aspects. This document summarizes technical results from a series of previous reports, and describes additional studies that can be done especially if site-specific information becomes available from one or more prospective repository sites.

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This progress report describes work being done at Sandia national Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions.

This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

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In July, 2014, the Electric Power Research Institute and industry partners sampled dust on the surface of an unused canister that had been stored in an overpack at the Hope Creek Nuclear Generating Station for approximately one year. The foreign material exclusion (FME) cover that had been on the top of the canister during storage, and a second recently-removed FME cover, were also sampled. This report summarizes the results of analyses of dust samples collected from the unused Hope Creek canister and the FME covers. Both wet and dry samples of the dust/salts were collected, using SaltSmart(TM) sensors and Scotch-Brite(TM) abrasive pads, respectively. The SaltSmart(TM) samples were leached and the leachate analyzed chemically to determine the composition and surface load per unit area of soluble salts present on the canister surface. The dry pad samples were analyzed by X-ray fluorescence and by scanning electron microscopy to determine dust texture and mineralogy; and by leaching and chemical analysis to deter mine soluble salt compositions. The analyses showed that the dominant particles on the canister surface were stainless steel particles, generated during manufacturing of the canister. Sparse environmentally - derived silicates and aluminosilicates were also present. Salt phases were sparse, and consisted of mostly of sulfates with rare nitrates and chlorides. On the FME covers, the dusts were mostly silicates/aluminosilicates; the soluble salts were consistent with those on the canister surface, and were dominantly sulfates. It should be noted that the FME covers were w ashed by rain prior to sampling, which had an unknown effect of the measured salt loads and compositions. Sulfate salts dominated the assemblages on the canister and FME surfaces, and included Ca-SO₄ , but also Na-SO₄, K-SO₄, and Na-Al-SO₄. It is likely that these salts were formed by particle - gas conversion reactions, either prior to, or after, deposition. These reactions involve reaction of carbonate, chloride, or nitrate salts with at mospheric SO₂, sulfuric acid, or a mmonium sulfate to form sulfate minerals. The Na-Al-SO₄ phase is unusual, and may have formed by reaction of Na-Al containing phases in aluminum smelter emissions with SO₂ , also present in smelter emissions. An aluminum smelter is located in Camden, NJ, 40 miles NE of the Hope Creek Site.

This document provides supporting technical rationale in draft form as input for: STAD Performance Specification Requirements Rationale (planned to be issued as FCRD-NFST-2015- 000106). The supporting rationale addresses those parts of the storage, transportation and disposal (STAD) canister performance specification that pertain to geologic disposal. These parts include service lifetime, repository thermal performance, postclosure criticality, and materials selection. Each draft performance specification is presented, and those pertaining to geologic disposal are elaborated with brief sections labeled Rationale. More detail is provided in appendices that address: long-term performance of borated stainless steel, use of borated stainless steel in existing designs, recommendations for additional corrosion studies, stress corrosion cracking of canister shell materials, and a generic postclosure safety case for STAD canisters.

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For long-term storage, spent nuclear fuel (SNF) is placed in dry storage cask systems, commonly consisting of welded stainless steel containers enclosed in ventilated cement or steel overpacks. At near-marine sites, failure by chloride-induced stress corrosion cracking (SCC) due to deliquescence of deposited salt aerosols is a major concern. This paper presents a preliminary probabilistic performance assessment model to assess canister penetration by SCC. The model first determines whether conditions for salt deliquescence are present at any given location on the canister surface, using an abstracted waste package thermal model and site-specific weather data (ambient temperature and absolute humidity). As the canister cools and aqueous conditions become possible, corrosion is assumed to initiate and is modeled as pitting (initiation and growth). With increasing penetration, pits convert to SCC and a crack growth model is implemented. The SCC growth model includes rate dependencies on temperature and crack tip stress intensity factor. The amount of penetration represents the summed effect of corrosion during time steps when aqueous conditions are predicted to occur. Model results and sensitivity analyses provide information on the impact of model assumptions and parameter values on predicted storage canister performance, and provide guidance for further research to reduce uncertainties.

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This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface of in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters. At both sites, the surface dust could be divided into fractions generated by manufacturing processes and by natural processes. The fraction from manufacturing processes consisted of variably-oxidized angular and spherical particles of stainless steel and iron, generated by machining and welding/cutting processes, respectively. Dust from natural sources consisted largely of detrital quartz and aluminosilicates (feldspars and clays) at both sites. At Hope Creek, soluble salts were dominated by sulfates and nitrates, mostly of calcium. Chloride was a trace component and the only chloride mineral observed by SEM was NaCl. Chloride surface loads measured by the Saltsmart™ sensors were very low, less than 60 mg m^–2 on the canister top, and less than 10 mg m^–2 on the canister sides. At Diablo Canyon, sea-salt aggregates of NaCl and Mg-SO₄, with minor K and Ca, were abundant in the dust, in some cases dominating the observed dust assemblage. Measured Saltsmart™ chloride surface loads were very low (<5 mg m^–2); however, high canister surface temperatures damaged the Saltsmart™ sensors, and, in view of the SEM observations of abundant sea-salts on the package surfaces, the measured surface loads may not be valid. Moreover, the more heavily-loaded canister tops at Diablo Canyon were not sampled with the Saltsmart™ sensors. The observed low surface loads do not preclude chloride-induced stress corrosion cracking (CISCC) at either site, because (1) the measured data may not be valid for the Diablo Canyon canisters; (2) the surface coverage was not complete (for instance, the 45º offset between the outlet and inlet vents means that near-inlet areas, likely to have heavier dust and salt loads, were not sampled); and (3) CISCC has been experimentally been observed at salt loads as low as 5-8 mg/m². Experimental efforts at SNL to assess corrosion of interim storage canister materials include three tasks in FY14. First, a full-diameter canister mockup, made using materials and techniques identical to those used to make interim storage canisters, was designed and ordered from Ranor Inc., a cask vendor for Areva/TN. The mockup will be delivered prior to the end of FY14, and will be used for evaluating weld residual stresses and degrees of sensitization for typical interim storage canister welds. Following weld characterization, the mockup will be sectioned and provided to participating organizations for corrosion testing purposes. A test plan is being developed for these efforts. In a second task, experimental work was carried out to evaluate crevice corrosion of 304SS in the presence of limited reactants, as would be present on a dustcovered storage canister. This work tests the theory that limited salt loads will limit corrosion penetration over time, and is a continuation of work carried out in FY13. Laser confocal microscopy was utilized to assess the volume and depth of corrosion pits formed during the crevice corrosion tests. Results indicate that for the duration of the current experiments (100 days), no stifling of corrosion occurred due to limitations in the amount of reactants present at three different salt loadings. Finally, work has been carried out this year perfecting an instrument for depositing sea-salts onto metal surfaces for atmospheric corrosion testing purposes. The system uses an X-Y plotter system with a commercial airbrush, and deposition is monitored with a quartz crystal microbalance. The system is capable of depositing very even salt loadings, even at very low total deposition rates.

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After an exposition of the materials used in DPCs and the factors controlling material corrosion in disposal environments, a survey is given of the corrosion rates, mechanisms, and products for commonly used stainless steels. Research needs are then identified for predicting stability of DPC materials in disposal environments. Stainless steel corrosion rates may be low enough to sustain DPC basket structural integrity for performance periods of as long as 10,000 years, especially in reducing conditions. Uncertainties include basket component design, disposal environment conditions, and the in-package chemical environment including any localized effects from radiolysis. Prospective disposal overpack materials exist for most disposal environments, including both corrosion allowance and corrosion resistant materials. Whereas the behavior of corrosion allowance materials is understood for a wide range of corrosion environments, demonstrating corrosion resistance could be more technically challenging and require environment-specific testing. A preliminary screening of the existing inventory of DPCs and other types of canisters is described, according to the type of closure, whether they can be readily transported, and what types of materials are used in basket construction.

This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to establish corrosion rates and component lifetimes. Finally, it is unlikely that the aluminum-based neutron absorber materials that are commonly used in existing DPCs would survive for 10,000 years in disposal environments, because the aluminum will act as a sacrificial anode for the steel. We recommend additional testing of borated and Gd-bearing stainless steels, to establish general and localized corrosion resistance in repository-relevant environmental conditions.

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This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions.

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Results reported here continue to support the FY13 conclusion that direct disposal of DPCs is technically feasible, at least for some DPCs, and for some disposal concepts (geologic host media). Much of the work performed has reached a point where site-specific information would be needed for further resolution. Several activities in FY14 have focused on clay/shale media because of potential complications resulting from low thermal conductivity, limited temperature tolerance, and the need to construct hundreds of kilometers of emplacement drifts that will remain stable for at least 50 years. Technologies for rapid excavation and liner installation have significantly advanced in the past 20 years. Tunnel boring machines are the clear choice for large-scale excavation. The first TBM excavations, including some constructed in clay or shale media, are now approaching 50 years of service. Open-type TBMs are a good choice but the repository host formation would need to have sufficient compressive strength for the excavation face to be self-supporting. One way to improve the strength-stress relationship is to reduce the repository depth in soft formations (e.g., 300 m depth). The fastest construction appears to be possible using TBMs with a single-pass liner made of pre-fabricated concrete segments. Major projects have been constructed with prefabricated segmented liner systems, and with cast-in-place concrete liners. Cost comparisons show that differences in project management and financing may be larger cost factors than the choice of liner systems. Costs for large-scale excavation and construction in clay/shale media vary widely but can probably be limited to $10,000 per linear meter, which is similar to previous estimates for repository construction. Concepts for disposal of DPC-based waste packages in clay/shale media are associated with thermal management challenges because of the relatively low thermal conductivity and limited temperature tolerance. Peak temperature limits of 100°C or lower for clay-rich materials have been selected by some international programs, but a limit above 100°C could help to shorten the duration of surface decay storage and repository ventilation. The effects of locally higher peak temperatures on repository performance need to be evaluated (in addition to the effects at lower temperatures). This report describes a modeling approach that couples the TOUGH2 and FLAC3D codes to represent thermally driven THM processes, as a demonstration of the types of models needed.

Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

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Report RWEV-REP-001, Analysis of Postclosure Groundwater Impacts for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High Level Radioactive Waste at Yucca Mountain, Nye County, Nevada was issued by the DOE in 2009 and is currently being updated. Sandia National Laboratories (SNL) provided support for the original document, performing calculations and extracting data from the Yucca Mountain Performance Assessment Model that were used as inputs to the contaminant transport and dose calculations by Jason Associates Corporation, the primary developers of the DOE report. The inputs from SNL were documented in LSA-AR-037, Inputs to Jason Associates Corporation in Support of the Postclosure Repository Supplemental Environmental Impact Statement. To support the updating of the original Groundwater Impacts document, SNL has reviewed the inputs provided in LSA-AR-037 to verify that they are current and appropriate for use. The results of that assessment are documented here.

This paper summarizes modeling of waste container degradation in performance assessments conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. As understanding of the Yucca Mountain disposal system increased, modeling of container degradation evolved from a component of the source term in 1984 to a separate module describing both container and drip shield degradation in 2008. A thermal module for evaluating the influence of higher heat loads from more closely packed, large waste packages was also introduced. In addition, a module for evaluating drift chemistry was added in later PAs to evaluate the potential for localized corrosion of the outer barrier of the waste container composed of Alloy 22, a highly corrosion-resistant nickel-chromium-tungsten-molybdenum alloy. The uncertainty of parameters related to container degradation contributed significantly to the estimated uncertainty of performance measures (cumulative release in assessments prior to 1995 and individual dose, thereafter). © 2013 Elsevier Ltd.

The chemical potential of water may play an important role in adsorption and capillary condensation of water under multiphase conditions at geologic CO2 storage sites. Injection of large volumes of anhydrous CO 2 will result in changing values of the chemical potential of water in the supercritical CO2 phase. We hypothesize that the chemical potential will at first reflect the low concentration of dissolved water in the dry CO2. As formation water dissolves into and is transported by the CO2 phase, the chemical potential of water will increase. We present a pore-scale model of the CO2-water interface or menisci configuration based on the augmented Young-Laplace equation, which combines adsorption on flat surfaces and capillary condensation in wedge-shaped pores as a function of chemical potential of water. The results suggest that, at a given chemical potential for triangular and square pores, liquid water saturation will be less in the CO2-water system under potential CO2 sequestration conditions relative to the air-water vadose zone system. The difference derives from lower surface tension of the CO2-water system and thinner liquid water films, important at pore sizes <1 × 10 -6 m, relative to the air-water system. Water movement due to capillary effects will likely be minimal in reservoir rocks, but still may be important in finer grained, clayey caprocks, where very small pores may retain water and draw water back into the system via adsorption and capillary condensation, if dry-out and then rewetting were to occur. © 2014. American Geophysical Union. All Rights Reserved.

A primary concern with dry storage of spent nuclear fuel is chloride-induced stress corrosion cracking, caused by deliquescence of salts deposited on the stainless steel canisters. However, limited access through the ventilated overpacks and high surface radiation fields impede direct examination of cask surfaces for CISCC, or sampling of surface deposits. Predictive models for CISCC must be able to predict the occurrence of a corrosive chemical environment (a chloride-rich brine formed by dust deliquescence) at specific locations (e.g. weld zones) on the canister surface. The presence of a deliquescent brine is controlled by the relative humidity (RH), which is a function of absolute humidity and cask surface temperature. This requires a thermal model that includes the canister and overpack design, canister-specific waste heat load, and passive cooling by ventilation. Brine compositions vary with initially-deposited salt assemblage, reactions with atmospheric gases, temperature, and the relative rates of salt deposition and reaction; predicting brine composition requires site-specific compositional data for atmospheric aerosols and acid gases. Aerosol particle transport through the overpack and deposition onto the canister must also be assessed. Initial field data show complex variability in the amount and composition of deposited salts as a function of canister surface location.

A primary concern with dry storage of spent nuclear fuel is chloride-induced stress corrosion cracking, caused by deliquescence of salts deposited on the stainless steel canisters. However, limited access through the ventilated overpacks and high surface radiation fields impede direct examination of cask surfaces for CISCC, or sampling of surface deposits. Predictive models for CISCC must be able to predict the occurrence of a corrosive chemical environment (a chloride-rich brine formed by dust deliquescence) at specific locations (e.g. weld zones) on the canister surface. The presence of a deliquescent brine is controlled by the relative humidity (RH), which is a function of absolute humidity and cask surface temperature. This requires a thermal model that includes the canister and overpack design, canister-specific waste heat load, and passive cooling by ventilation. Brine compositions vary with initially-deposited salt assemblage, reactions with atmospheric gases, temperature, and the relative rates of salt deposition and reaction; predicting brine composition requires site-specific compositional data for atmospheric aerosols and acid gases. Aerosol particle transport through the overpack and deposition onto the canister must also be assessed. Initial field data show complex variability in the amount and composition of deposited salts as a function of canister surface location.

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For the interim storage of used nuclear fuel, the storage casks/containers will be exposed to conditions under which considerable dust and/or atmospheric aerosols may be deposited on the surface. These dust layers may contain a sizeable portion of water soluble salts, particularly in marine environments where many interim storage systems are located. These soluble salts will deliquesce if sufficient moisture is present, resulting in the formation of potentially corrosive brine on the material surface. Experimental results have illustrated that some stainless steels, such as 304SS (a common material of construction for interim storage containers) can and will undergo localized corrosion in elevated temperature conditions where a chloride rich brine has formed on the surface. Results presented here illustrate that it is possible that stifling of localized attack will result when limited reactant is present, but additional analysis is necessary before a definite conclusion can be made.

Once sufficiently cool, spent nuclear fuel is stored in dry storage cask systems, most commonly consisting of welded stainless steel containers enclosed in ventilated concrete or steel overpacks. As the United States does not currently have a viable disposal pathway for SNF, these containers may be required to perform their waste isolation function for many decades beyond their original design criteria. Failure by stress corrosion cracking due to deliquescence of deposited salt aerosols is a major concern. Parameters controlling deliquescence include the temperature and RH at the waste package surface, and the composition of deposited salts. The timing and duration of deliquescence under in situ conditions is poorly defined, because of uncertainties in thermal history, the large variability in temperatures over the storage container surface, and uncertainties in the composition of deposited salts. Storage installations in near-marine environments are of greatest concern because of exposure to significant quantities of chloride-rich sea salt aerosols. Published stainless steel corrosion studies with sea salt and sea salt components suggest that conditions conducive to localized corrosion initiation and propagation may exist on the surface of SNF storage containers in such environments at some point in their extended service life, and furthermore, that stress corrosion cracking may occur over a broad range of potentially relevant conditions. However, the studies were carried out with heavy salt loads and limited gas flow, which may limit the beneficial effects of brine/atmosphere exchange (e.g., acid degassing, CO2 exchange, degassing and thermal decomposition of ammonium phases). Gas exchange with the atmosphere will modify brine pH and chloride content, and will modify the deliquescent salt assemblage through precipitation of Ca and Mg carbonates, potentially limiting brine volumes or resulting in dryout. Nitrate-rich inland salt aerosols are considered less corrosive, but may have higher levels of potentially reactive pollutants. Moreover, the compositions of inland salt deposits on hot storage containers may have greater uncertainty, as ammonium- and nitrate-rich salt assemblages are subject to thermal decomposition and potential reactions with organics. For both inland and near-marine sites, little information is available on the dust/salt deposition rates, or the quantity of salt present on existing storage container surfaces. A sampling program for in situ dust deposits on current storage containers will provide critical compositional data for new stress corrosion cracking studies, and will allow evaluation of the applicability of existing studies of stainless steel stress corrosion cracking under conditions of dust deliquescence.

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In the supercritical CO2-water-mineral systems relevant to subsurface CO2 sequestration, interfacial processes at the supercritical fluid-mineral interface will strongly affect core- and reservoir-scale hydrologic properties. Experimental and theoretical studies have shown that water films will form on mineral surfaces in supercritical CO2, but will be thinner than those that form in vadose zone environments at any given matric potential. The theoretical model presented here allows assessment of water saturation as a function of matric potential, a critical step for evaluating relative permeabilities the CO2 sequestration environment. The experimental water adsorption studies, using Quartz Crystal Microbalance and Fourier Transform Infrared Spectroscopy methods, confirm the major conclusions of the adsorption/condensation model. Additional data provided by the FTIR study is that CO2 intercalation into clays, if it occurs, does not involve carbonate or bicarbonate formation, or significant restriction of CO2 mobility. We have shown that the water film that forms in supercritical CO2 is reactive with common rock-forming minerals, including albite, orthoclase, labradorite, and muscovite. The experimental data indicate that reactivity is a function of water film thickness; at an activity of water of 0.9, the greatest extent of reaction in scCO2 occurred in areas (step edges, surface pits) where capillary condensation thickened the water films. This suggests that dissolution/precipitation reactions may occur preferentially in small pores and pore throats, where it may have a disproportionately large effect on rock hydrologic properties. Finally, a theoretical model is presented here that describes the formation and movement of CO2 ganglia in porous media, allowing assessment of the effect of pore size and structural heterogeneity on capillary trapping efficiency. The model results also suggest possible engineering approaches for optimizing trapping capacity and for monitoring ganglion formation in the subsurface.

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