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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 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

First Principles Determination of the Potential-of-Zero-Charge in an Alumina-Coated Aluminum/Water Interface Model for Corrosion Applications

The surfaces of most metals immersed in aqueous electrolytes have a several-nanometer-thick oxide/hydroxide surface layer. This gives rise to the existence of both metal|oxide and oxide|liquid electrotlyte interfaces, and makes it challenging to correlate atomic length-scale structures with electrochemical properties such the potential-of-zero-charge (PZC). The PZC has been shown to be correlated the onset potential for pitting corrosion. Here, we conduct large-scale Density Functional Theory and ab initio molecular dynamics to calculate the PZC of a Al(111)| γ -Al 2 O 3 (110)| water double-interface model within the context of aluminum corrosion. By partitioning the multiple interfaces involved into binary components with additive contributions to the overall work function and voltage, we predict the PZC to be -1.53 V vs SHE for this model. Furthermore, we calculate the orbital energy levels of defects like oxygen vacancies in the oxide, which are critical parameters in theories associated with pitting corrosion. We predict that the Fermi level at the PZC lies above the impurity defect levels of the oxygen vacancies, which are therefore uncharged at the PZC. From the PZC estimate, we predict the voltage needed to create oxygen vacancies with net positive charges within a flatband approximation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Improved Application of Local Models to Steel Corrosion in Lead-Bismuth Loops

The corrosion of steels exposed to flowing liquid metals is influenced by local and global conditions of flow systems. The present study improves the previous local models when applied to closed loops by incorporating some global condition effects. In particular the bulk corrosion product concentration is calculated based on balancing the dissolution and precipitation in the entire closed loop. Mass transfer expressions developed in aqueous medium and an analytical expression are tested in the liquid-metal environments. The improved model is applied to a pure lead loop and produces results closer to the experimental data than the previous local models do. The model is also applied to a lead-bismuth eutectic (LBE) test loop. Systematic studies illustrate the effects of the flow rate, the oxygen concentration in LBE, and the temperature profile on the corrosion rate.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Triple-wavelength, narrowband Mg/SiC multilayers with corrosion barriers and high peak reflectance in the 25-80 nm wavelength region

We have developed new, corrosion-resistant Mg/SiC multilayer coatings which can be used to efficiently and simultaneously reflect extreme ultraviolet (EUV) radiation in single or multiple narrow bands centered at wavelengths in the spectral region from 25 to 80 nm. Corrosion mitigation is achieved through the use of partially amorphous Al-Mg thin layers. Three different multilayer design concepts were developed and deposited by magnetron sputtering and the reflectance was measured at near-normal incidence in a broad spectral range. Unprotected Mg/SiC multilayers were also deposited and measured for comparison. They were shown to efficiently reflect radiation at a wavelength of 76.9 nm with a peak reflectance of 40.6% at near-normal incidence, the highest experimental reflectance reported at this wavelength for a narrowband coating. The demonstration of multilayer coatings with corrosion resistance and multiplewavelength EUV performance is of great interest in the development of mirrors for space-borne solar physics telescopes and other applications requiring long-lasting coatings with narrowband response in multiple emission lines across the EUV range.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2012

Corrosion Mechanisms in Chloride and Carbonate Salts

State of the art thermal solar power uses nitrate salts as the heat transfer fluid. In order to increase operating efficiencies high turbine inlet temperatures must be achieved. Different heat transfer fluids must be considered for this purpose. Corrosion mechanisms in chloride and carbonate based salts were reviewed to better understand the practical implications for using either salt in a solar system. Chloride salts are susceptible to high corrosion rates in the presence of moisture and oxygen. Purification techniques are extremely important to ensure good materials performance, as formation of protective layers does not readily occur. Carbonate salts form protective oxide layers, which inhibit corrosion from the base alloy. Large amounts of work have been done for molten carbonate fuel cells, but different atmospheric conditions and temperature (>650°C) will merit investigation for solar applications.

Kruizenga, Alan Michael↗
DOE OSTI2013

Technical Letter Report on the Cracking of Irradiated Stainless Steels in Low-Corrosion-Potential Environments

Irradiation-assisted stress corrosion cracking (IASCC) is a major degradation mechanism for reactor internal components of austenitic stainless steels (SSs) in light water reactors. Crack growth rate (CGR) and fracture toughness J-R curve tests were performed on four irradiated SS specimens in high-purity water with low dissolved oxygen and in a simulated pressurized water reactor (PWR) environment. The samples had been previously irradiated from ~5 to 8 dpa in the BOR-60 reactor. The materials were cold-worked (CW) 316 and solution-annealed (SA) 304L SSs, which are commonly used in PWR core internals. Cyclic and constant-load CGRs were obtained from these samples to assess their cracking behaviors in low-corrosion-potential environments. A Ti-stabilized specimen, CW 316-Ti SS, was found to be highly vulnerable to IASCC, and its cracking susceptibility increased with neutron dose between ~5 and 8 dpa. The CW 316 and SA 304L SSs showed moderate cracking susceptibilities in the low-corrosion potential environments up to ~7 dpa. Periodic partial unloading (PPU) had a significant impact on the crack growth behavior of all irradiated specimens, and the measured CGRs were more than one order of magnitude higher in the test periods with than without PPU. A “stair-step” crack growth behavior was also observed for all irradiated specimens during the test periods with PPU, suggesting a strong effect of dynamic loading on IASCC. The SA 304L SS showed the best fracture resistance, and the CW 316 SS was the most brittle among the tested specimens. The fracture toughness J values of the two CW 316-Ti SS specimens were nearly identical at ~5 and ~8 dpa.

Chen, Y.↗
DOE OSTI2014

Corrosion report for the U-Mo fuel concept

The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A systematic multiscale modeling and experimental approach to protect grain boundaries in magnesium alloys from corrosion

A multiscale modeling Internal State Variable (ISV) constitutive model was developed that captures the fundamental structure-property relationships. The macroscale ISV model used lower length scale simulations (Butler-Volmer and Electronics Structures results) in order to inform the ISVs at the macroscale. The chemomechanical ISV model was calibrated and validated from experiments with magnesium (Mg) alloys that were investigated under corrosive environments coupled with experimental electrochemical studies. Because the ISV chemomechanical model is physically based, it can be used for other material systems to predict corrosion behavior. As such, others can use the chemomechanical model for analyzing corrosion effects on their designs.

36 MATERIALS SCIENCE↗
DOE OSTI2017

FY16 Status Report: SNF Interim Storage Canister Corrosion and Surface Environment Investigations

This progress report describes work done in FY15 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. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Corrosion Monitors for Embedded Evaluation

We have developed and characterized novel in-situ corrosion sensors to monitor and quantify the corrosive potential and history of localized environments. Embedded corrosion sensors can provide information to aid health assessments of internal electrical components including connectors, microelectronics, wires, and other susceptible parts. When combined with other data (e.g. temperature and humidity), theory, and computational simulation, the reliability of monitored systems can be predicted with higher fidelity.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

Surveillance and Monitoring Program Full-Scale Experiments to Evaluate the Potential for Corrosion in 3013 Containers

A set of six long-term, full-scale experiments were initiated to determine the type and extent of corrosion that occurs in 3013 containers packaged with chloride-bearing plutonium oxide materials. The materials were exposed to a high relative humidity environment representative of actual packaging conditions for the materials in storage. The materials were sealed in instrumented, inner 3013 containers with corrosion specimens designed to test the corrosiveness of the environment inside the containers under various conditions. This report focuses on initial loading conditions that are used to establish a baseline to show how the conditions change throughout the storage lifetime of the containers.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modeling Radionuclide Releases from TRISO Particles by Simultaneous Diffusion Through and Corrosion of the Silicon Carbide Barrier Layer

TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 μm thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 °C). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10 -20 m 2 /s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10 -21 m 2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

SNL Research into Stress Corrosion Cracking of SNF Dry Storage Canisters (FY19 Status Report)

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Crack Growth Rate and Large Plate Demonstration of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report describes the body of work involving testing and analyses performed in FY18 and FY19 at the Savannah River National Laboratory and at Korea University to determine chloride-induced stress corrosion crack (CISCC) growth rates in weldments of austenitic stainless steel plate material. The primary work involved: several separate crack growth rate test campaigns with new testing setups using laboratory-scale corrosion specimens; analyses to estimate the stress intensity factors for flaws postulated to be present in the residual stress fields of axial and circumferential weldments in a prototypic dry storage canister; and establishment of a “large plate” (approximately 51×46 cm with thickness 16 mm and weight about 30 kg) CISCC test that uses a plate with a circumferential weldment harvested from the Sandia National Laboratories’ full-size mockup dry storage canister. The tests for crack growth rate were conducted on ASTM E1861 bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse the brine to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The crack growth rate test results matched previous experimental data in the open literature. These literature data had been compiled by SNL and EPRI to develop a temperature-dependent (K-insensitive) crack growth rate model to support flaw disposition in the ASME Boiler and Pressure Vessel Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.” The knowledge of the salt, temperature, humidity conditions for CISCC including the threshold stress intensity factor, $K_{ISCC}$, were used to setup the large plate test in which a set of partthrough- wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a constant 75% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC. The visual results at the surface of the plate show corrosion product staining (brown rust appearance) but with no CISCC emanating from the machined notches after approximately 5 months exposure to date. The plate test will continue in FY20 and phased-array ultrasonic testing and other NDE techniques will be applied to identify and characterize CISCC that may emanate from the machined defects. The notches and any CISCC from them will be sectioned at the conclusion of the test to verify NDE characterization, and to develop effective through-wall crack growth rates from cracks that would likely be branched.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

M4SF-20LL010302042: Process Model for Radionuclide Incorporation into Corrosion Products

This progress report (Level 4 Milestone Number M4SF-20LL010302042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-20LL010302041. The research is focused on actinide and radionuclide sequestration in steel corrosion products. Fuel matrix degradation models suggest that the near field is likely to be reducing at the time of canister breaching, steel corrosion, and radionuclide release, but more oxidizing conditions may prevail in the far field. The incorporation of radionuclides into corrosion phases may limit the rate of radionuclide release by sequestering a portion of the radionuclide source term. For these reasons there is a need to evaluate the incorporation of Pu and other radionuclides into various Fe-oxide phases, and to understand the behavior of coprecipitated phases during mineral recrystallization processes and during re-oxidation events. Radionuclide coprecipitation with Fe minerals may impact long-term repository performance and is an ongoing research focus at Lawrence Livermore National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report provides a description of the testing (in progress at the Savannah River National Laboratory) to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The details of the establishment of this plate test, the interim corrosion results (from visual examination), and the general plan for examination of the plate in CY21 to evaluate crack growth behavior (if any) are provided. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage. A “large plate” (approximately 51 x 46 cm with thickness 16 mm and weight about 30 kg) and containing a circumferential weldment from the mockup is the test specimen. The knowledge of the salt, temperature, humidity conditions for chloride-induced stress corrosion cracking (CISCC) including the threshold stress intensity factor, K iscc , were used to setup the large plate test in which a set of part-through-wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A marine salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a sustained 71-73% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Full loop thermodynamic corrosion inhibition and sensing in molten chloride systems. Final report

The objective of this project is to increase the lifetime of Concentrated Solar Power (CSP) systems by identifying corrosion mitigation methods for heat transfer loops in the temperature range of 500 - 800 °C. Coupling of experimental data with CFD modeling has provided greater insight into corrosion mitigation methods and how they will be applied in molten salt systems with convective flows. The overall goal of the program is to demonstrate techniques that control corrosion while having limited impact on system alloys.

14 SOLAR ENERGY↗
DOE OSTI2021

Additive Manufacturing of Corrosion Resistant UHTC Materials for Chloride Salt-to-sCO2 Brayton Cycle Heat Exchangers

The objectives of the project include: (1) screen high-purity, high density ultra-high-temperature ceramics (UHTCs) for chloride salt corrosion resistance, (2) develop binder-jet additive manufacturing and sintering for ZrB 2 -MoSi 2 blends using literature and experiments to demonstrate feasibility, and (3) evaluate chloride salt corrosion resistance of additively manufactured ZrB 2 -MoSi 2 . These initial objectives support the development of a durable and reliable ultra-compact high-efficiency salt-to-sCO 2 heat exchanger for closed-loop Brayton power cycles based on corrosion resistant, high-strength, thermally conductive UHTC materials and novel designs enabled by additive manufacturing. The UHTC materials are also abrasion resistant so that they could be applicable to all three thermal carrier media pathways identified in the CSP Gen3 Roadmap.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Steel Corrosion Impacts on Radionuclide Immobilization

This progress report (Level 4 Milestone Number M4SF-21LL010302052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-21LL01030205. The research is focused on actinide and radionuclide sequestration in steel corrosion products. Fuel matrix degradation models suggest that the near field is likely to be reducing at the time of canister breaching, steel corrosion, and radionuclide release, but more oxidizing conditions may prevail in the far field. The incorporation of radionuclides into corrosion phases may limit the rate of radionuclide release by sequestering a portion of the radionuclide source term. For these reasons there is a need to evaluate the incorporation of Pu and other radionuclides into various Fe-oxide phases, and to understand the behavior of coprecipitated phases during mineral recrystallization processes and during re-oxidation events. Radionuclide coprecipitation with Fe minerals may impact long-term repository performance and is an ongoing research focus at Lawrence Livermore National Laboratory.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗