Publications

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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.

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.

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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.

This progress report describes work done in FY19 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 FY19 refined our understanding of the chemical and physical environment on canister surfaces and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs.

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This report discusses several possible sources of water that could persist in SNF dry storage canisters through the drying cycle. In some cases, the water is trapped in occluded geometries in the cask such as dashpots or damaged fuel. Persistence of water or ice in such locations seems unlikely, given the high heat load of the canistered fuel; this is especially true in the case of vacuum drying, where a strong driver exists to remove water vapor from the headspace of such occluded geometries. Water retention in Boral® core material is a known problem, that has in the past resulted in the need for much extended drying times. Since the shift to slightly higher porosity "blister resistant" Boral®, water drainage appears to be less of a problem. However, high surface areas for the Boral® core material will provide a trap for significant amounts of adsorbed water, at least some of which is certain to survive the drying process. Moreover, if corrosion within the cores produces hydrous aluminum corrosion products, these may also survive.

The phonon, infrared, and Raman spectroscopic properties of zirconium tungsten phosphate, Zr2(WO4)(PO4)2 (space group Pbcn, IT No. 60; Z = 4), have been extensively investigated using density functional perturbation theory (DFPT) calculations with the Perdew, Burke, and Ernzerhof exchange-correlation functional revised for solids (PBEsol) and validated by experimental characterization of Zr2(WO4)(PO4)2 prepared by hydrothermal synthesis. Using DFPT-simulated infrared, Raman, and phonon density-of-state spectra combined with Fourier transform infrared and Raman measurements, new comprehensive and extensive assignments have been made for the spectra of Zr2(WO4)(PO4)2, resulting in the characterization of its 29 and 34 most intense IR- and Raman-active modes, respectively. DFPT results also reveal that ν1(PO4) symmetric stretching and ν3(PO4) antisymmetric stretching bands have been interchanged in previous Raman experimental assignments. Negative thermal expansion in Zr2(WO4)(PO4)2 appears to have very limited impact on the spectral properties of this compound. This work shows the high accuracy of the PBEsol exchange-correlation functional for studying the spectroscopic properties of crystalline materials using first-principles methods.

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This study was initiated to quantify and characterize the uncertainty associated with the degradation mechanisms impacting normal dry storage operations for used nuclear fuel (UNF) and normal conditions of transport in support of the Spent Fuel and Waste Science & Technology Campaign (SFWST) and its effectiveness to rank the data needs and parameters of interest. This report describes the technical basis and guidance resulting from the development of software to perform uncertainty quantification (UQ) by developing and describing a holistic model that integrates the various processes controlling Atmospheric Stress Corrosion Cracking (ASCC) in the specific context of Interim Spent Fuel Storage Installations (ISFSIs). These processes include the daily and annual cycles of temperature and humidity associated with the environment, the deposition of chloride-containing aerosol particles, pit formation, pit-to-crack transition, and crack propagation.

The High Burn-Up Demonstration Project was recently initiated by the Department of Energy (DOE) to evaluate the effects of fuel drying and long term dry storage on high burn-up spent nuclear fuel. As part of the project, samples of the He backfill gas were collected 5 hours, 5 days, and 12 days after completion of drying. The samples provide information on the state of the fuel at closure, and on the environment within the cask. At Sandia National Laboratories, the samples were analyzed by gamma-ray spectroscopy to quantify fission product gases and by gas mass spectrometry to quantify bulk and trace gases; water content was measured via humidity probe. Gamma-ray spectroscopy results indicated no detectible ⁸⁵Kr, indicating no failed fuel rods were present after drying. Mass spectrometry indicated build-up of CO₂ to 930 ppmv over two weeks, attributed to oxidation of organic compounds (possibly vacuum grease or vacuum pump oil) within the cask. H₂, generated by either radiolysis or metal corrosion, also increased up to —500 ppmv. Water contents in the cask were higher than anticipated, increasing to —17,400 ppmv ±10% after 12 days. Measuring water content proved challenging, and possible improvements to the method for future analyses are proposed.

The DOE and industry collaborators have initiated the high burn-up demonstration project to evaluate the effects of drying and long-term dry storage on high burn-up fuel. Fuel was transferred to a dry storage cask, which was then dried using standard industry vacuum-drying techniques and placed on a storage pad to be opened and the fuel examined in 10 years. Helium fill gas samples were collected 5 hours, 5 days, and 12 days after closure. The samples were analyzed for fission gases (85Kr) as an indicator of damaged or leaking rods, and then analyzed to determine water content and concentrations of other trace gases. Gamma-ray spectroscopy found no detectible 85Kr. Sample water contents proved difficult to measure, requiring heating to desorb water from the inner surface of the sampling bottles. Final results indicated that water in the cask gas phase built up over 12 days to 17,400 ppmv ±10%, equivalent to ∼100 ml of water within the cask gas phase. Trace gases were measured by direct gas mass spectrometry. Carbon dioxide built up over two weeks to 930 ppmv, likely due to breakdown of hydrocarbon contaminants (possibly vacuum pump oil) in the cask. Hydrogen built up to nearly 500 ppmv. and may be attributable to water radiolysis and/or to metal corrosion in the cask.

We have investigated cubic zirconium tungstate (ZrW2O8) using density functional perturbation theory (DFPT), along with experimental characterization to assess and validate computational results. Cubic zirconium tungstate is among the few known materials exhibiting isotropic negative thermal expansion (NTE) over a broad temperature range, including room temperature where it occurs metastably. Isotropic NTE materials are important for technological applications requiring thermal-expansion compensators in composites designed to have overall zero or adjustable thermal expansion. While cubic zirconium tungstate has attracted considerable attention experimentally, a very few computational studies have been dedicated to this well-known NTE material. Therefore, spectroscopic, mechanical and thermodynamic properties have been derived from DFPT calculations. A systematic comparison of the calculated infrared, Raman, and phonon density-of-state spectra has been made with Fourier transform far-/mid-infrared and Raman data collected in this study, as well as with available inelastic neutron scattering measurements. The thermal evolution of the lattice parameter computed within the quasi-harmonic approximation exhibits negative values below the Debye temperature, consistent with the observed negative thermal expansion characteristics of cubic zirconium tungstate, α-ZrW2O8. These results show that this DFPT approach can be used for studying the spectroscopic, mechanical and thermodynamic properties of prospective NTE ceramic waste forms for encapsulation of radionuclides produced during the nuclear fuel cycle.