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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 397 records · Page 22

Complexity of segregation behavior at localized deformation sites formed while in service in a 316 stainless steel baffle-former bolt

Here, post-irradiation evaluation was performed on a 316 stainless steel baffle former bolt harvested after 40 years of service in a pressurized water reactor. Microstructure analysis revealed the presence of defect-free dislocation channels and strain-induced twins, indicative of loading at a stress level close to yield stress at least once while in service. Primary radiation-induced Ni/Si precipitates were destroyed during channel and twin formation, and secondary, significantly coarser Ni/Si precipitates formed along newly formed Σ3 boundaries during the continued irradiation. Additionally, an elevated phosphorus level was observed inside the strain-induced twin. Complex chemistry inside the strain-induced feature may overlap with dislocation pileups and impact localized corrosion, material long-term performance, and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Anti-Biofouling Surface Treatments for Improved Condenser Performance for Coal-Based Power Plants

Condenser performance is a critical area for improving the economic efficiency of coal-based power facilities, since more than 50% of generated heat is lost during condenser operation. Decline in condenser performance due to fouling of condenser heat transfer tubes greatly has a significant and detrimental effect on power plant efficiency and controlling biofilm formation remains a major operational challenge. Biofilm formation in condenser tubes decreases the effective diameter for cooling water and increases the surface roughness, which increases the hydraulic resistance of each tube. In addition, biofouling can result in the degradation of the condenser tubing by facilitating corrosion. Therefore, this project aims to develop novel surface treatments and secondarily applied remediation components to mitigate biofilm growth on condenser tube surfaces used in coal-fueled power plants.

01 COAL, LIGNITE, AND PEAT↗

Modeling Environmental Effects on Ventilated Spent Fuel Storage Systems

This report describes newly developed external environment wind effects models of spent nuclear fuel (SNF) dry storage systems. The primary purpose of these wind effects models is to better understand particle deposition on SNF canisters in the context of chloride-induced stress corrosion cracking. The goal of this effort is to further the understanding and improve the wind effects model of the Nuclear Horizontal Modular Storage (NUHOMS ® ) Advanced Horizontal Storage Module–High Seismic (AHSM-HS) storage system in support of the planned Canister Deposition Field Demonstration (CDFD) project (Durbin et al. 2021, Suffield et al. 2021). The steps in this wind effects investigation begin with a validation case of comparing experimental data with a STAR-CCM+ computational fluid dynamics (CFD) model of the Belowground Vertical Dry Cask Simulator (BVDCS) with the external environment explicitly modeled. Next, a test case is performed with the AHSM-HS models comparing solar-loading strategies for a standalone and wind effects model in STAR-CCM+ and a standalone model in ANSYS Mechanical Parametric Design Language (APDL). The final portion of this study compares results of the validation exercise with applications of wind effects models for two horizontal storage systems, a site specific NUHOMS ® horizontal storage module and a NUHOMS ® AHSM-HS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Irradiation damage reduces alloy corrosion rate via oxide space charge compensation effects

Radiation effects in materials often compound and accelerate other detrimental phenomena such as embrittlement, oxidation and creep. However, irradiation can also decrease the oxidation rate, for instance with ZrNb alloys nuclear fuel cladding. In this study, we rationalize this observation on Zr-0.5Nb alloy by introducing a mechanism based on oxide space charge modification, resulting from irradiation enhanced Nb clustering. Here, this mechanism is investigated using a multiscale approach: from the macroscale, to determine post-irradiation oxidation kinetics, to the atomic scale, using in-situ atom probe tomography sample oxidation, to observe elemental solute redistribution across the oxide/metal interface. The mechanism is further supported by high resolution transmission electron microscopy characterization and density functional theory calculations. A point defect model is proposed to account for oxide space charge effects and their changes under irradiation. This integrated, multiscale experimental and modeling approach challenges the current paradigm on irradiation effects and how they can potentially improve materials performance in extreme environments.

36 MATERIALS SCIENCE↗

Deliquescence of Eutectic LiCl-KCl Diluted with NaCl for Interim Waste Salt Storage

Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, such operations must be performed under controlled atmospheric conditions, and waste salts require careful storage to avoid deliquescence and corrosion of container materials. This study investigated a potential processing path for reducing the degree of deliquescence through dilution to varying extents with NaCl. Here, the hydration behavior of LiCl-KCl salts diluted with NaCl was evaluated in terms of mass gain due to water absorption, degree of deliquescence (including first appearances of standing water), and evidence of corrosion to stainless steel containers in a humid air environment (40°C, 20% relative humidity). In this humid air environment, pure eutectic LiCl-KCl exhibited a 50 mass % increase due to water absorption and showed evidence of standing water after 24 h. Waste salt diluted with NaCl required loadings of 89 mass % NaCl in order to prevent deliquescence and exhibited a 3 mass % increase due to water absorption. After periodic observation for 48 h, standing water was observed near all ingots with the exception of the 89 mass % NaCl samples. Dilution with 89% NaCl was also found to reduce evidence of corrosion when stored in stainless steel crucibles. While dilution with NaCl greatly decreases steady-state hydration, the storage volume is increased ~10× through this procedure.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Use of in Situ Synchrotron Techniques to Probe the Oxidized Surface of Molybdenum Nitride Oxygen Reduction Electrocatalysis

The development of active and stable earth-abundant catalysts for the oxygen reduction reaction (ORR) is needed for widespread, economic development of fuel cell technologies. Designing and optimizing these non-platinum group metals is challenging, however, because they are susceptible to composition and structure changes, including dissolution, oxidation, and corrosion, both in air and under reaction conditions. To identify the active surface, and thus understand the properties that affect activity, the catalyst surface must be characterized in situ. Herein, we utilize a grazing incidence electrochemical cell to investigate in situ composition and morphology changes of a molybdenum nitride (Mo-N) thin film catalyst using grazing incidence x-ray absorption spectroscopy (GI-XAS) and x-ray reflectivity (XRR). In rotating ring disk electrode measurements, we find that the activity, selectivity, stability, and capacitance of the Mo-N catalyst is dependent on the maximum potential to which it has been exposed. Specifically, the overpotential required to reach -2 mA cm-2geo decreases by over 90 mV when the maximum potential is increased from 0.3 to 0.8 V vs RHE (Figure 1). Because the Mo-N oxidizes rapidly in air, ex situ characterization methods including x-ray photoelectron spectroscopy and time-of-flight secondary ion mass spectrometry can provide only limited insight into these in situ catalyst changes. Using in situ GI-XAS measurements at applied potentials between 0.3 and 0.9 V vs RHE, however, we are able to determine that the surface of the film oxidizes and becomes more amorphous when exposed to increasingly higher potentials (Figure 1). Furthermore, the surface remains oxidized on the order of several hours when returned to "ORR relevant potentials" (< 0.6 V vs RHE), indicating that this surface-oxidized nitride is the active surface for ORR. Using in situ XRR measurements, we find that there is no change in surface roughness at potentials below 0.7 V vs RHE, but the film roughens significantly at 0.8 V vs RHE, correlating with ex situ measurements of Mo dissolution at this potential (Figure 1). We therefore conclude that the intrinsic activity of the Mo-N catalyst increases when exposed to potentials up to 0.7 V vs RHE, while above that potential activity enhancements are due to the exposure of more active sites through dissolution. The in situ electrochemical surface-sensitive x-ray characterization as used here is a promising methodology for understanding and leveraging surface dynamics to improve the performance of non-traditional catalysts.

Kreider, Melissa↗

Method of manufacturing a reinforced nuclear fuel cladding using an intermediate thermal deposition layer

A method is described that includes the steps of making a thin walled Zr alloy tube, loading nuclear fuel pellets into the tube, compressing the tube onto the fuel pellets to substantially reduce free space around the fuel pellets, positioning end plugs at each of two ends of the tube, filling the tube with a heat transferring gas, and coating the compressed tube with a corrosion resistant material using a thermal deposition process, such as cold spray, before inserting the tube into a pre-formed SiC composite cover having at least one closed end.

Lahoda, Edward J.↗

Recent developments in catalyst-related PEM fuel cell durability

Cost and durability remain the two major barriers to the widespread commercialization of polymer electrolyte membrane fuel cell (PEMFC)-based power systems, especially for the most impactful but challenging fuel cell electric vehicle (FCEV) application. Commercial FCEVs are now on the road; however, their PEMFC systems do not meet the cost targets established by the U.S. Department of Energy, primarily due to the high platinum loading needed on the cathode to achieve the requisite performance and lifetime. Furthermore, while the activities of a number of commercial Pt-based alloy cathode catalysts exceed the beginning-of-life (BOL) targets, these activities, and the overall cathode performance, degrade via a variety of mechanisms described herein. Degradation is mitigated in current FCEVs by utilizing a cathode catalyst with a lower BOL activity (e.g., much lower transition metal alloy content and larger BOL nanoparticle size), necessitating higher catalyst loadings, and through the utilization of system controls that avoid conditions known to exacerbate degradation processes, such as limiting the fuel cell stack voltage range. The design and development of active and robust materials and eliminating the need for vehicle mitigation strategies would greatly simplify the operating system, allowing for greater transient operation, avoiding large hybridization, and curtailing of fuel cell power. Although system mitigation strategies have provided the near-term pathway for FCEV commercialization, material-specific solutions are required to further reduce costs and improve operability and efficiency. Future material developments should focus on stabilization of the electrode structure and minimization of the catalyst particle susceptibility to dissolution caused by oxide formation and reduction over PEMFC cathode-relevant operating potentials plus minimization of support corrosion. Ex situ accelerated stress tests have provided insight into the processes responsible for material and performance degradation and will continue to provide useful information on the relative stability of materials and benchmarks for robust and stable materials-based solutions not requiring system mitigation strategies to achieve adequate lifetime.

25 ENERGY STORAGE↗

Multiscale Modeling of Silicon Carbide Cladding for Nuclear Applications: Thermal Performance Modeling

The complex multiscale and anisotropic nature of silicon carbide (SiC) ceramic matrix composite (CMC) makes it difficult to accurately model its performance in nuclear applications. The existing models for nuclear grade composite SiC do not account for the microstructural features and how these features can affect the thermal and structural behavior of the cladding and its anisotropic properties. In addition to the microstructural features, the properties of individual constituents of the composites and fiber tow architecture determine the bulk properties. Models for determining the relationship between the individual constituents’ properties and the bulk properties of SiC composites for nuclear applications are absent, although empirical relationships exist in the literature. Here, a hierarchical multiscale modeling approach was presented to address this challenge. This modular approach addressed this difficulty by dividing the various aspects of the composite material into separate models at different length scales, with the evaluated property from the lower-length-scale model serving as an input to the higher-length-scale model. The multiscale model considered the properties of various individual constituents of the composite material (fiber, matrix, and interphase), the porosity in the matrix, the fiber volume fraction, the composite architecture, the tow thickness, etc. By considering inhomogeneous and anisotropic contributions intrinsically, our bottom-up multiscale modeling strategy is naturally physics-informed, bridging constitutive law from micromechanics to meso-mechanics and structural mechanics. The effects that these various physical attributes and thermo-physical properties have on the composite’s bulk thermal properties were easily evaluated and demonstrated through the various analyses presented herein. Since silicon carbide fiber-reinforced SiC CMCs are also promising thermal–structural materials with a broad range of high-end technology applications beyond nuclear applications, we envision that the multiscale modeling method we present here may prove helpful in future efforts to develop and construct reinforced CMCs and other advanced composite nuclear materials, such as MAX phase materials, that can service under harsh environments of ultrahigh temperatures, oxidation, corrosion, and/or irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Analysis of Dust Samples Collected from an Inland ISFSI Site (''Site B'').

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Multimodal analysis and characterization of the boehmite layer formed on AA6061 before and after alkaline etching

Low-enriched uranium (LEU) alloyed with 10% Mo (U-10Mo) is being considered as a promising alternative to oxide-based dispersion fuel with high-enriched uranium for use in research reactors. The configuration of this proposed LEU monolithic LEU fuel plate consists of a U-10Mo plate-type fuel foil with a 25 µm zirconium interlayer barrier clad with an aluminum alloy (AA6061). In certain research reactors, the clad AA6061 is coated with a boehmite layer to prevent corrosion. The boehmite layer has a high-pH passivation range, which makes it resistant to oxidation. Boehmite is usually formed on the AA6061 surface by autoclave processing. Before the boehmite layer is added, the surface of the AA6061 is cleaned using techniques such as polishing and wet etching. In this study, we use multimodal analysis to examine how pretreatment of AA6061 using polishing followed by alkaline etching affects the chemical composition of the boehmite layer. X-ray photoelectron microscopy (XPS), transmission electron microscopy (TEM), and x-ray diffraction (XRD) were used to study the chemical changes in the boehmite layer caused by alkaline etching pretreatment. XPS provides quantitative analysis for the Al:O ratio as well as oxidation states present on the surface, which suggests slight oxidation of the boehmite surface after alkaline etching of the AA6061 surface. We further explored this suggested oxidation of the boehmite surface using high-resolution transmission electron microscopy with selected area electron diffraction (SAED) and grazing incidence x-ray diffraction (GI-XRD), which suggested only a small amount of aluminum oxide at the surface. The multimodal analysis and imaging yielded new insights for optimizing boehmite growth on AA6061 for research reactors.

36 MATERIALS SCIENCE↗

The Use of Microelectrodes in Molten Salt Electrochemistry

Molten salts have attracted considerable interest as essential media for advanced high-temperature technologies, including molten salt reactors, thermal energy storage, high-temperature electrolysis, and pyrochemical processing. Their ability to remain stable in liquid form at elevated temperatures, combined with favorable thermophysical properties and wide electrochemical windows, makes them highly suitable for applications involving heat transfer, energy storage, and hightemperature electrochemical processing. However, despite these advantages, molten salts present challenges due to their chemically reactive nature at high-temperatures, especially in the presence of oxidizing impurities. Salt chemistry can fluctuate through interactions with impurities over time, or fuel burnup in the case of molten salt reactors, often leading to the dissolution of metal species. This dynamic environment not only results in complex redox behavior but also promotes corrosion, which is rarely uniform and frequently manifests as localized degradation driven by structural materials’ compositional differences, electrochemical imbalances, and microstructural susceptibilities.

Kim, Changkyu [University of Wisconsin-Madison, WI↗

Estimation of Respirable Aerosol Release Fractions through Stress Corrosion Crack-Like Geometries

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the significant backfill pressurization of the canister, up to approximately 800 kPa. This pressure differential offers a relatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the carrier gas flow rates and aerosol transmission properties were evaluated for an engineered microchannel with characteristic dimensions similar to those of an SCC. The microchannel was formed by mating two gage blocks with a slot orifice measuring 28.9 μm (0.0011 in.) tall by 12.7 mm (0.500 in.) wide by 8.86 mm (0.349 in.) long (flow length). Surrogate aerosols of cerium oxide, Ce0 2 , were seeded and mixed inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to demonstrate a new capability to characterize SCCs under well-controlled boundary conditions. Separate modeling efforts are also underway that will be validated using these data. The test apparatus and procedures developed in this study can be easily modified for the evaluation of more complex SCC-like geometries including laboratory-grown SCC samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Continued Investigations of Respirable Release Fractions for Stress Corrosion Crack-Like Geometries

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the commercial practice of using relatively high backfill pressures (up to approximately 800 kPa) in the canister to enhance internal natural convection. This pressure differential offers a comparatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the rates of gas flow and aerosol transmission of a spent fuel surrogate through an engineered microchannel with dimensions representative of an SCC were evaluated experimentally using coupled mass flow and aerosol analyzers. The microchannel was formed by mating two gage blocks with a linearly tapering slot orifice nominally 13 μm (0.005 in.) tall on the upstream side and 25 μm (0.0010 in.) tall on the downstream side. The orifice is 12.7 mm (0.500 in.) wide by 8.89 mm (0.350 in.) long (flow length). Surrogate aerosols of cerium oxide, CeO 2 , were seeded and mixed with either helium or air inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to demonstrate a new capability to characterize SCCs under well-controlled boundary conditions. Modeling efforts were also initiated that evaluate the depletion of aerosols in a commercial dry storage canister. These preliminary modeling and ongoing testing efforts are focused on understanding the evolution in both size and quantity of a hypothetical release of aerosolized spent fuel particles from failed fuel to the canister interior and ultimately through an SCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Quantification of Aerosol Transmission through Stress Corrosion Crack-Like Geometries

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the commercial practice of using more significant backfill pressures in the canister, up to approximately 800 kPa. This pressure differential offers a relatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the rates of gas flow and aerosol transmission of a spent fuel surrogate through an engineered microchannel with dimensions representative of an SCC were evaluated experimentally using coupled mass flow and aerosol analyzers. The microchannel was formed by mating two gage blocks with a linearly tapering slot orifice nominally 13 μm (0.005 in.) tall on the upstream side and 25 μm (0.0010 in.) tall on the downstream side. The orifice is 12.7 mm (0.500 in.) wide by 8.86 mm (0.349 in.) long (flow length). Surrogate aerosols of cerium oxide, CeO 2 , were seeded and mixed with either helium or air inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to add to previous testing that characterized SCCs under well-controlled boundary conditions through the inclusion of testing improvements that establish initial conditions in a more consistent way. These ongoing testing efforts are focused on understanding the evolution in both size and quantity of a hypothetical release of aerosolized spent fuel particles from failed fuel to the canister interior and ultimately through an SCC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Testing of Microchannels and Lab-Grown Stress Corrosion Cracks for Quantification of Aerosol Transmission

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the commercial practice of using higher backfill pressures in the canister, up to approximately 800 kPa, compared to their horizontal counterparts. This pressure differential offers a relatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the rates of gas flow and aerosol transmission of a spent fuel surrogate through an engineered microchannel with dimensions representative of an SCC were evaluated experimentally using coupled mass flow and aerosol analyzers. The microchannel was formed by mating two gage blocks with a linearly tapering slot orifice nominally 13 μm (0.005 in.) tall on the upstream side and 25 μm (0.0010 in.) tall on the downstream side. The orifice is 12.7 mm (0.500 in.) wide by 8.86 mm (0.349 in.) long (flow length). Surrogate aerosols of cerium oxide, CeO 2 , were seeded and mixed with either helium or air inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to add to previous testing that characterized SCCs under well-controlled boundary conditions through the inclusion of testing improvements that establish initial conditions in a more consistent way. While the engineered microchannel has dimensions similar to actual SCCs, it does not reproduce the tortuous path the aerosol laden flow would have to traverse for eventual transmission. SCCs can be rapidly grown in a laboratory setting given the right conditions, and initial characterization and clean-flow testing has begun on lab grown crack samples provided to Sandia National Laboratories (SNL). Many such samples are required to produce statistically relevant transmission results, and SNL is developing a procedure to produce samples in welded steel plates. These ongoing testing efforts are focused on understanding the evolution in both size and quantity of a hypothetical release of aerosolized spent fuel particles from failed fuel to the canister interior and ultimately through an SCC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Preliminary HEA Solid Phase Processing Development Report

The development of high temperature fuel cladding materials to withstand a variety of extreme environments have received much attention. Several potential materials systems that have been identified for the fuel systems and core structural materials application in advanced reactor systems are ferritic/martensitic steel (e.g., HT9), austenitic stainless steels (e.g., 316 LN), oxide-dispersion strengthened steels (e.g., 12 YWT), Ni-based alloys and ceramic-based composites depending on the type of the reactors. Though these material systems have promising properties conducive for radiation-resistant performance, they suffer beyond the design-limit from one or more damage processes such as void swelling, radiation embrittlement, phase instability, corrosion, and limited creep life.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗