Publications

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This progress report describes work performed during FY21 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of canister materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In FY21, modeling and experimental work was performed that further defined our understanding of the potential chemical and physical environment present on canister surfaces at both marine and inland sites. Research also evaluated the relationship between the environment and the rate, extent, and morphology of corrosion, as well as the corrosion processes that occur. Finally, crack growth rate testing under relevant environmental conditions was initiated.

This report summarizes the current actives in FY21 related to the effort by Sandia National Laboratories to identify and test coating materials for the prevention, mitigation, and repair of spent nuclear fuel dry storage canisters against potential chloride-induced stress corrosion cracking. This work follows up on the details provided in Sandia National Laboratories FY20 report on the same topic, which provided a detailed description of the specific coating properties desired for application and implementation on spent nuclear fuel canisters, as well as provided detail into several different coatings and their applicability to coat spent nuclear fuel canisters. In FY21, Sandia National Laboratories has engaged with private industry to create a Memorandum of Understanding and established a collaborative R&D program building off the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of private industry. The resulting Memorandum of Understanding included four companies to date (Oxford Performance Materials, White Horse R&D, Luna Innovations, and Flora Coating) proposing six different coating technologies (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin, silane-based polyurethane hybrid with and without a Znrich primer, and a quasi-ceramic sol-gel polyurethane hybrid) to be tested, evaluated, and optimized for their potential use for this application. This report provides a detailed description of each of the coating systems proposed by the participating industry partners. It also provides a description of the planned experimental actives to be performed by Sandia National Laboratories including physical tests, electrochemical tests, and characterization methods. These analyses will be used to identify specific ways to further improve coating technologies toward their application and implementation on spent nuclear fuel canisters. In FY21, Sandia National Laboratories began baseline testing of the base metal material in according with activities of the Memorandum of Understanding. In FY22, Sandia National Laboratories will receive coated coupons from each of the participating industry partners and begin characterization, physical, and electrochemical testing following the test plan described herein.

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Stress corrosion cracking (SCC) is an important failure degradation mechanism for storage of spent nuclear fuel. Since 2014, Sandia National Laboratories has been developing a probabilistic methodology for predicting SCC. The model is intended to provide qualitative assessment of data needs, model sensitivities, and future model development. In fiscal year 2021, improvement of the SCC model focused on the salt deposition, maximum pit size, and crack growth rate models.

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Corrosion-resistant welded alloys are frequently used as a leak-tight boundary in critical applications that require confinement of hazardous and/or radioactive substances, including an increasing population of spent nuclear fuel (SNF) canisters. The behavior of residual stresses generated as a result of irregular elastic–plastic deformation during processes such as welding is one of today's key issues to a full understanding of the aging mechanisms that may compromise the confinement boundary. Whether such processes and any subsequent weld repairs, not subjected to post-weld heat treatment, would negatively affect the initial material by introducing through-thickness tensile stresses remains an open question. Here we report the first residual stress measurements using neutron diffraction on the welded joints of a SNF canister. We found significant tensile residual stresses in the as welded sample, indicating that initiation and through-thickness growth of cracks may be possible. Following repair, we observed a stress redistribution and introduction of beneficial compressive stresses. We anticipate our results will improve understanding of confinement susceptibility to aging and guide improvements in repair techniques.

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This report describes plans for dust sampling and analysis for the multi-year Canister Deposition Field Demonstration. The demonstration will use three commercial 32PTH2 NUHOMS welded stainless steel storage canisters, which will be stored at an ISFSI site in Advanced Horizontal Storage Modules. One canister will be unheated; the other two will have heaters to achieve canister surface temperatures that match, to the degree possible, spent nuclear fuel (SNF) loaded canisters with heat loads of 10 kW and 40 kW. Surface sampling campaigns will take place on a yearly or bi-yearly basis. The goal of the planned dust sampling and analysis is to determine important environmental parameters that impact the potential occurrence of stress corrosion cracking on SNF dry storage canisters. Specifically, the size, morphology, and composition of the deposited dust and salt particles will be quantified, as well as the soluble salt load per unit area and the rate of deposition, as a function of canister surface temperature, location, time, and orientation. Sampling locations on the canister surface will nominally include 25 locations, corresponding to 5 circumferential locations at each of the 5 longitudinal locations. At each sampling location, a 2x2 sampling grid (containing 4 sample cells) will be painted onto the metal surface. During each sampling campaign, two samples at each sampling location will be collected, in a specific routine to measure both periodic (yearly or bi-yearly) and cumulative deposition rates. For each sample, a wet and a dry sample will be collected. Wet samples will be analyzed to determine the composition of the soluble salt fraction and to estimate salt loading per unit area. Dry samples will be analyzed to assess particle size, morphology, mineralogy, and identity (e.g. for floral/faunal fragments). The data generated by this proposed sampling plan will provide detailed information on dust and salt aerosol deposits on spent nuclear fuel canister surfaces. The anticipated results include information regarding particle compositions, size distributions, and morphologies, in addition to particle deposition rates as a function of canister surface location, orientation, time, and temperature. The information gathered during the Canister Deposition Field Demonstration is critical for ongoing efforts to develop a detailed understanding of the potential for stress corrosion cracking on SNF dry storage canisters

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Maximum pit sizes were predicted for dilute and concentrated NaCl and MgCl2 solutions as well as sea-salt brine solutions corresponding to 40% relative humidity (RH) (MgCl2-rich) and 76% RH (NaCl-rich) at 25 °C. A quantitative method was developed to capture the effects of various cathode evolution phenomena including precipitation and dehydration reactions. Additionally, the sensitivity of the model to input parameters was explored. Despite one's intuition, the highest chloride concentration (roughly 10.3 M Cl−) did not produce the largest predicted pit size as the ohmic drop was more severe in concentrated MgCl2 solutions. Therefore, the largest predicted pits were calculated for saturated NaCl (roughly 5 M Cl−). Next, it was determined that pit size predictions are most sensitive to model input parameters for concentrated brines. However, when the effects of cathodic reactions on brine chemistry are considered, the sensitivity to input parameters is decreased. Although there was not one main input parameter that influenced pit size predictions, two main categories were identified. Under similar chloride concentrations (similar RH), the water layer thickness (WL), and pit stability product, (i·x)sf, are the most influential factors. When varying chloride concentrations (RH), changes in WL, the brine specific cathodic kinetics on the external surface (captured in the equivalent current density (ieq)), and conductivity (κo) are the most influential parameters. Finally, it was noted that dehydration reactions coupled with precipitation in the cathode will have the largest effect on predicted pit size, and cause the most significant inhibition of corrosion damage.

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Cathodic kinetics in magnesium chloride (MgCl2) solutions were investigated on platinum (Pt) and stainless steel 304 L (SS304 L). Density, viscosity, and dissolved oxygen concentration for MgCl2 solutions were also measured. A 2-electron transfer for oxygen reduction reaction (ORR) on Pt was determined using a rotating disk electrode. SS304 L displayed non-Levich behavior for ORR and, due to ORR suppression and buffering of near surface pH by Mg-species precipitation, the primary cathodic reaction was the hydrogen evolution reaction (HER) in saturated MgCl2. Furthermore, non-carbonate precipitates were found to be kinetically favored. Implications of HER are discussed through atmospheric corrosion and stress corrosion cracking.

In September of 2020, dust samples were collected from the surface of spent nuclear fuel (SNF) dry storage canisters during an inspection at an inland Independent Spent Fuel Storage Installation. The purpose of the sampling was to assess the composition and abundance of the soluble salts present on the canister surface, information which provides a metric for potential corrosion risks. The samples were delivered to Sandia National laboratories for analysis. At Sandia, the soluble salts were leached from the dust and quantified by ion chromatography. In addition, subsamples of the dust were taken for scanning electron microscope analysis to determine the texture and mineralogy of the dust and salts. The results of those analyses are presented in this report.

In October of 2020, dust samples were collected from the surface of spent nuclear fuel (SNF) dry storage canisters during an inspection at an inland Independent Spent Fuel Storage Installation, the second inland site at which surface deposits have been sampled. The purpose of the sampling was to assess the composition and abundance of the soluble salts present on the canister surface, information which provides a metric for potential corrosion risks. The samples were delivered to Sandia National laboratories for analysis. At Sandia, the soluble salts were leached from the dust and quantified by ion chromatography. In addition, subsamples of the dust were taken for scanning electron microscopy to determine the texture and mineralogy of the dust and salts. The results of those analyses are presented in this report.

Disposal of large, heat-generating waste packages containing the equivalent of 21 pressurized water reactor (PWR) assemblies or more is among the disposal concepts under investigation for a future repository for spent nuclear fuel (SNF) in the United States. Without a long (>200 years) surface storage period, disposal of 21-PWR or larger waste packages (especially if they contain high-burnup fuel) would result in in-drift and near-field temperatures considerably higher than considered in previous generic reference cases that assume either 4-PWR or 12-PWR waste packages (Jové Colón et al. 2014; Mariner et al. 2015; 2017). Sevougian et al. (2019c) identified high-temperature process understanding as a key research and development (R&D) area for the Spent Fuel and Waste Science and Technology (SFWST) Campaign. A two-day workshop in February 2020 brought together campaign scientists with expertise in geology, geochemistry, geomechanics, engineered barriers, waste forms, and corrosion processes to begin integrated development of a high-temperature reference case for disposal of SNF in a mined repository in a shale host rock. Building on the progress made in the workshop, the study team further explored the concepts and processes needed to form the basis for a high-temperature shale repository reference case. The results are described in this report and summarized..

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This progress report describes work performed during FY20 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY20 further defined our understanding of the potential chemical and physical environment present on canister surfaces, evaluated the relationship between the environment and the resultant corrosion that occurs, and initiated crack growth rate testing under relevant environmental conditions. In FY20, work to define dry storage canister surface environments included several tasks. First, collection of dust deposition specimens from independent spent fuel storage installation (ISFSI) site locations helped to establish a more complete understanding of the potential chemical environment formed on the canister. Second, the predicted evolution of canister surface relative humidity RH) values was estimated using ISFSI site weather data and the horizontal canister thermal model used by the SNL probabilistic SCC model. These calculations determined that for typical ISFSI weather conditions, seasalt deliquescence to produce MgCl2-rich brines could occur in less than 20 years at the coolest locations on the canister surface, and, even after nearly 300 years, conditions for NaCl deliquescence (75% RH) are not reached. This work illustrates the importance of understanding the stability of MgCl2-rich brines on the heated canister surface, and the potential impact of brine composition on corrosion processes, including pitting and stress corrosion cracking. In an additional study, the description of the canister surface environment was refined in order to define more realistic corrosion testing environments including diurnal cycles, soluble salt chemistries, and inert mineral particles. The potential impacts of these phenomena on canister corrosion are being evaluated experimentally. Finally, work over the past few years to evaluate the stability of magnesium chloride brines continued in FY20. MgCl2 degassing experiments were carried out, confirming that MgCl2 brines slowly degas HCl on heated surfaces, converting to less deliquescent magnesium hydroxychloride phases and potentially leading to brine dryout.

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This report summarizes the results of a literature survey on coatings and surface treatments that are used to provide corrosion protection for exposed metal surfaces. The coatings are discussed in the context of being used on stainless steel spent nuclear fuel (SNF) dry storage canisters for potential prevention or repair of corrosion and stress corrosion cracking. The report summarizes the properties of different coating classes, including the mechanisms of protection, their physical properties, and modes of degradation (thermal, chemical, radiological). Also discussed are the current standard technologies for application of the coatings, including necessary surface pretreatments (degreasing, rust removal, grinding) and their effects on coating adhesion and performance. The coatings are also classified according their possible use for in situ repair; ex situ repair, requiring removal from the overpack; and ex situ prevention, or application prior to fuel loading to provide corrosion protection over the lifetime of the canister.

Here, for the first time, we demonstrate the use of an in situ spectroelectrochemical Raman technique to explore simulated atmospheric corrosion scenarios with a variable boundary layer thickness (δ). The effects of solution flow rate on oxygen concentration and δ were explored. It was found solution regeneration is necessary to prevent oxygen depletion in the Raman cell. It was further shown that by increasing the solution flow rate, the effective δ decreases and allows for the investigation of atmospheric corrosion scenarios. Finally, the technique developed was utilized to explore the effect of precipitation on the cathodic behavior of SS304L in dilute MgCl₂. During cathodic polarization, evidence supports previous observations that magnesium hydroxide species are kinetically favored over the thermodynamically predicted magnesium carbonate.

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The incorporation of uranium, plutonium and technetium in the negative thermal expansion (NTE) α-Zr(WO4)2 has been investigated within the framework of density functional theory (DFT). It is found that the vacancy formation energies of the charged vacancies are overall larger than that of its counterpart neutral Frenkel defects and Schottky defects. DFT calculations suggest that U and Pu substitutions for the Zr site are preferred in α-Zr(WO4)2. In case of Tc substitution, both Tc(IV) for the Zr site and Tc(VII) for the W site are considered under oxygen-poor and oxygen-rich conditions, while Tc(VII) substitution can be improved significantly by including Y2O3 (charge compensation).

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Classical molecular dynamics (MD) simulations were performed to provide a conceptual understanding of the amorphous-crystalline interface for a candidate negative thermal expansion (NTE) material, ZrW2O8. Simulations of pressure-induced amorphization at 300 K indicate that an amorphous phase forms at pressures of 10 GPa and greater, and this phase persists when the pressure is subsequently decreased to 1 bar. However, the crystalline phase is recovered when the slightly distorted 5 GPa phase is relaxed to 1 bar. Simulations were also performed on a two-phase model consisting of the high-pressure amorphous phase in direct contact with the crystalline phase. Upon equilibration at 300 K and 1 bar, the crystalline phase remains unchanged beyond a thin layer of disrupted structure at the crystalline-amorphous interface. Differences in local atomic structure at the interface are quantified from the simulation trajectories.

One of the primary concerns with the long-term performance of storage systems for spent nuclear fuel (SNF) is the potential for corrosion due to deliquescence of salts deposited as aerosols on the surface of the canister, which is typically made of austentic stainless steel. In regions of high residual weld stresses, this may lead to localized stress-corrosion cracking (SCC). The ability to detect and image SCC at an early stage (long before the cracks are susceptible to propagate through the thickness of the canister wall and leaks of radioactive material may occur) is essential to the performance evaluation and licensing process of the storage systems. In this paper, we explore a number of nondestructive testing techniques to detect and image SCC in austenitic stainless steel. Our attention is focused on a small rectangular sample of 1 × 2 in2 with two cracks of mm-scale sizes. The techniques explored in this paper include nonlinear resonant ultrasound spectroscopy (NRUS) for detection, Linear Elastodynamic Gradient Imaging Technique (LEGIT), ultrasonic C-scan, vibrothermography, and synchrotron X-ray diffraction for imaging. Results obtained from these techniques are compared. Cracks of mm-scale sizes can be detected and imaged with all the techniques explored in this study.

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This report describes the potential of a novel class of materials—α-ZrW₂O₈, Zr₂WP₂O₁₂, and related compounds that contract upon amorphization as possible radionuclide waste-forms. The proposed ceramic waste-forms would consist of zoned grains, or sintered ceramics with center- loaded radionuclides and barren shells. Radiation-induced amorphization would result in core shrinkage but would not fracture the shells or overgrowths, maintaining isolation of the radionuclide. In this report, we have described synthesis techniques to produce phase-pure forms of the materials, and how to fully densify those materials. Structural models for the materials were developed and validated using DFPT approaches, and radionuclide substitution was evaluated; U(IV), Pu(IV), Tc(IV) and Tc(VII) all readily substitute into the material structures. MD modeling indicated that strain associated with radiation-induced amorphization would not affect the integrity of surrounding crystalline materials, and these results were validated via ion beam experimental studies. Finally, we have evaluated the leach rates of the barren materials, as determined by batch and flow-through reactor experiments. ZrW₂O₈ leaches rapidly, releasing tungstate while Zr is retained as a solid oxide or hydroxide. Tungsten release rates remain elevated over time and are highly sensitive to contact times, suggesting that this material will not be an effective waste-form. Conversely, tungsten releases rates from Zr₂WP₂O₁₂ rapidly drop, show little dependence on short-term changes in fluid contact time, and in over time, become tied to P release rates. The results presented here suggest that this material may be a viable waste-form for some hard-to-handle radionuclides such as Pu and Tc.