Effects of liquid and vapor cesium on structural materials
Literature survey on corrosive effects of liquid and vapor cesium on structural materials, and compatibility of cesium as working fluid for Rankine cycle space power plant
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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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A Multicontinuum-Theory-Based Approach to the Analysis of Fiber-Reinforced Polymer Composites with Degraded Stiffness and Strength Properties Due to Moisture Absorption
Marine energy generation technologies such as wave and tidal power have great potential in meeting the need for renewable energy in the years ahead. Yet, many challenges remain associated with marine-based systems because of the corrosive environment. Conventional materials like metals are subject to rapid corrosive breakdown, crippling the lifespan of structures in such environments. Fiber-reinforced polymer composites offer an appealing alternative in their strength and corrosion resistance, but can experience degradation of mechanical properties as a result of moisture absorption. An investigation is conducted to test the application of a technique for micromechanical analysis of composites, known as multicontinuum theory and demonstrated in past works, as a mechanism for predicting the effects of prolonged moisture absorption on the performance of fiber-reinforced composites. Experimental tensile tests are performed on composite coupons with and without prolonged exposure to a salt water solution to obtain stiffness and strength properties. Multicontinuum theory is applied in conjunction with micromechanical modeling to deduce the effects of moisture absorption on the behavior of constituent materials within the composites. The results are consistent with experimental observations when guided by known mechanisms and trends from previous studies, indicating multicontinuum theory as a potentially effective tool in predicting the long-term performance of composites in marine environments.
36 MATERIALS SCIENCE↗
Refractory Materials for Flame Deflector Protection System Corrosion Control: Flame Deflector Protection System Life Cycle Cost Analysis Report
A 20-year life cycle cost analysis was performed to compare the operational life cycle cost, processing/turnaround timelines, and operations manpower inspection/repair/refurbishment requirements for corrosion protection of the Kennedy Space Center launch pad flame deflector associated with the existing cast-in-place materials and a newer advanced refractory ceramic material. The analysis compared the estimated costs of(1) continuing to use of the current refractory material without any changes; (2) completely reconstructing the flame trench using the current refractory material; and (3) completely reconstructing the flame trench with a new high-performance refractory material. Cost estimates were based on an analysis of the amount of damage that occurs after each launch and an estimate of the average repair cost. Alternative 3 was found to save $32M compared to alternative 1 and $17M compared to alternative 2 over a 20-year life cycle.
Calle, Luz Marina↗
Corrosion Testing of Brazed Space Station IATCS Materials
Increased nickel concentrations in the IATCS coolant prompted a study of the corrosion rates of nickel-brazed heat exchangers in the system. The testing has shown that corrosion is occurring in a silicon-rich intermetallic phase in the braze filler of coldplates and heat exchangers as the result of a decrease in the coolant pH brought about by cabin carbon dioxide permeation through polymeric flexhoses. Similar corrosion is occurring in the EMU de-ionized water loop. Certain heat exchangers and coldplates have more silicon-rich phase because of their manufacturing method, and those units produce more nickel corrosion product. Silver biocide additions did not induce pitting corrosion at silver precipitate sites.
Pohlman, Matthew J.↗
FY23 Progress Report: SRNL Analysis of ICCWR LCM and WAMS data for Corrosion and Cracking
Algorithms for Machine Learning (ML) and data analysis for the 3013 Surveillance Program have been developed in an ongoing collaborative effort by the Savannah River National Laboratory (SRNL) and the University of South Carolina (USC). The objective of the algorithms is to automate the identification of corrosion and crack formation in the Inner Container Closure Weld Region (ICCWR) of the canister system used to store Pu-bearing material. Data for corrosion and cracking is collected from large binary files generated by a Laser Confocal Microscope (LCM), the Wide Area 3D Measurement System (WAMS), and in a recent proposal, by a Scanning Electron Microscope (SEM). The ML software uses the physical attributes in the data files (e.g., one or all of: height, color, and grayscale values as functions of position in a plane projection) to detect the presence of surface corrosion and cracking after being trained on similar data with the features to be detected labeled.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
ECLSS Universal Waste Management System (UWMS) Metal Materials Compatibility Study - Elgiloy: Electrochemical and Crevice Corrosion Evaluation
Electrochemical evaluations were conducted on samples of Elgiloy in support of the materials compatibility program for the Environmental Control and Life Support System (ECLSS). Results were compared alongside three previously tested metals: Titanium 6Al-4V, Inconel 625, and Hastelloy C276. Titanium exhibited the best results of all the metals, indicating superior corrosive nobility and galvanic protection properties. Elgiloy demonstrated anti-corrosion behavior which appeared to be superior to the two nickel-chromium alloys. For this current effort, the results have clearly shown that Elgiloy possesses excellent corrosion protection properties and is galvanically compatible with the other three metals in the subject test media.
Lee, R. E.↗
Electrical research on solar cells and photovoltaic materials
The flat-plate solar cell array program which increases the service lifetime of the photovoltaic modules used for terrestrial energy applications is discussed. The current-voltage response characteristics of the solar cells encapsulated in the modules degrade with service time and this degradation places a limitation on the useful lifetime of the modules. The most desirable flat-plate array system involves solar cells consisting of highly polarizable materials with similar electrochemical potentials where the cells are encapsulated in polymers in which ionic concentrations and mobilities are negligibly small. Another possible mechanism limiting the service lifetime of the photovoltaic modules is the gradual loss of the electrical insulation characteristics of the polymer pottant due to water absorption or due to polymer degradation from light or heat effects. The mechanical properties of various polymer pottant materials and of electrochemical corrosion mechanisms in solar cell material are as follows: (1) electrical and ionic resistivity; (2) water absorption kinetics and water solubility limits; and (3) corrosion characterization of various metallization systems used in solar cell construction.
Orehotsky, J.↗
Development of hydrothermal corrosion model and BWR metal coating for CVD SiC in light water reactors
SiC/SiC fiber composites with CVD SiC overcoats are potential candidates for light water reactor advanced accident tolerant cladding materials. Understanding its corrosion kinetics in Light Water Reactor (LWR) conditions is essential to evaluate the concept's viability. Existing models only account for the temperature and oxygen concentration effect on the hydrothermal corrosion behavior applicable to LWR operating conditions. However, the development of a general corrosion rate for CVD SiC that accounts for the impact of irradiated microstructure, flow rate, electrical resistivity, pH, and surface roughness is critical for the practical realization of SiC/SiC-based cladding concepts. After a rigorous experimental campaign, this work updates the existing hydrothermal corrosion model to predict hydrothermal corrosion in LWRs. Numerical radiation and coolant chemistry analysis for LWRs conducted based on the updated corrosion kinetic models suggests that CVD SiC is likely a viable environmental barrier coating for Pressurized Water Reactors while questionable for Boiling Water Reactors (BWR) when the effect of irradiation damage on SiC corrosion is considered. An effective mitigation strategy for the double-layer metal coating is proposed for BWR applications. The double-layer metal coating comprising a FeCrAl overcoat with an intermediate Cr bond coating was observed to provide a stable protective barrier against SiC dissolution in BWR conditions. Finally, the proposed metal coating was also fully adherent following quench and burst tests.
36 MATERIALS SCIENCE↗
Corrosion testing needs and considerations for additively manufactured materials in nuclear reactors
Additive manufacturing (AM) technologies have developed rapidly in recent years, creating new opportunities and challenges for the nuclear industry; however, adoption requires that their corrosion performance be evaluated. Here, we discuss known reactor-specific corrosion issues for multiple reactor types and engineering concerns such as regulations and standards. A review of corrosion studies conducted on select AM alloys informs a discussion on key bulk and surface factors likely to impact corrosion behaviors. Recommendations to assess corrosion performance for AM materials are provided, including management of the unique nature of as-built AM surfaces and the inherent process variability that occurs for AM components.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Corrosion testing needs and considerations for additively manufactured materials in nuclear reactors
Metal additive manufacturing holds significant promise as an enabling technology for the 21st century nuclear energy industry. Metal additive manufacturing (MAM) can allow the fabrication of novel materials and innovative component designs that are not achievable through conventional manufacturing. Due to its very different fabrication methods, as-fabricated MAM components are characteristically different from conventionally manufactured components. MAM materials exhibit very different microstructures from conventional cast or wrought materials. For example, austenitic stainless steels fabricated by laser powder bed fusion exhibit columnar grain structures, dislocation cell structures, and melt pool fingerprints. In addition, MAM fabrication may result in defects such as porosity, incomplete processing of the feedstock (e.g., lack of fusion in melt-based methods), and oxide inclusions. Heat treatments may further evolve the microstructure, microsegregation, and stresses within the component. Furthermore, MAM components have a rough surface with feature sizes on the order of the feedstock material, as opposed to smooth surfaces resulting from conventional machining and forming operations. As a result, the corrosion behavior of MAM components will likely be significantly different from that of conventionally formed components. Corrosive environments for structural materials within advanced reactor environments include molten fluoride and chloride salts, liquid sodium and lead-bismuth, and high-temperature helium. The Advanced Materials and Manufacturing Technologies program within the Department of Nuclear Energy in the United States Department of Energy is assessing the unique concerns of MAM component corrosion and testing methodologies in advanced nuclear reactor environments. We discuss these concerns and testing strategies in this paper.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Enhanced steam oxidation resistance of uranium nitride nuclear fuel pellets
Here, the steam oxidation resistance of UN and UN-(20 vol%)ZrN fuel pellets is evaluated to enhance understanding of steam corrosion mechanisms in advanced nuclear fuel materials. In situ neutron diffraction shows the modified UN fuel pellets form a (U 0.77 ,Zr 0.23 )N solid-solution and the sole crystalline oxidation product detected in bulk is (U 0.77 ,Zr 0.23 )O 2 . U 2 N 3 is not detected in significant quantities during the steam oxidation of UN or (U 0.77 ,Zr 0.23 )N and stable lattice parameters show that hydriding does not take place. Steam oxidation rates, obtained via sequential Rietveld refinement show how (U 0.77 ,Zr 0.23 )N has a higher activation energy (79 ± 1 kJmol -1 vs. 50 ± 5 kJmol -1 ), higher onset temperature (430 °C vs. 400 °C) and slower reaction rates for steam oxidation up to 616 °C, than pure UN. Throughout, both UN and (U 0.77 ,Zr 0.23 )N exhibit linear (non-protective) oxidation kinetics, signifying that degradation of the fuel pellets is caused by the evolution of gaseous products at the interface followed by oxide scale spallation. This quantitative and mechanistic understanding of material degradation enables better defined operating regimes and points towards (U,Zr)N solid solutions as a promising strategy for the design of advanced nuclear fuel materials with enhanced steam corrosion resistance.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Viability of Cathodic Protection for Preventing Corrosion of Stainless Steel 316H in Molten LiF-NaF-KF
Molten fluoride salts are candidate heat transfer fluids in a number of applications such as generation IV molten salt nuclear reactors and concentrated solar power plants. However, a chief concern in the design of these systems is the corrosion of structural materials that come in contact with these molten salts. Redox control methods such as the purification of salt, the addition of active elements, and applied electrochemical potential can be efficient methods for preventing the corrosion of structural materials in molten fluoride salts. Applied electrochemical potential as a redox control method for application in molten fluoride salts has rarely been explored. This study seeks to understand the viability of impressed current cathodic protection (CP) at various currents as a redox control method to prevent corrosion of stainless steel 316H in molten LiF-NaF-KF (FLiNaK) salt. Results show that application of CP can be an effective method to prevent corrosion of SS316H in molten FLiNaK salt, but the applied current will have to be optimized to prevent undesirable side effects such as reduction of salt constituents, salt deposition on electrodes, etc.
Materials Science↗
Elucidating the corrosion mechanism of commercial Ni-based superalloys in UCl3 containing-chloride molten salt systems
Elucidating the role of UCl3 in the corrosion mechanism of Ni-based superalloys exposed to chloride molten salts Trishelle Copeland-Johnson1, Michael Woods1, Ruchi Gakhar1, Daniel J. Murray1, Guoping Cao1, Lingfeng He1 1Idaho National Laboratory, Idaho Falls, ID, United States The United States Department of Energy (DOE) aims to diversify the domestic energy portfolio towards more sustainable options, including implementation of molten salt reactor (MSR) technology. Chloride molten salts have been investigated as an appropriate MSR coolant and fuel because their relatively inexpensive, abundant, and exhibit favorable thermophysical properties. However, the corrosivity of chloride molten salts have not been extensively studied, especially with the inclusion of actinide products, such as UCl3. Accordingly, the development of nuclear structural materials with excellent corrosion performance is critical to the successful implementation of MSRs, particularly from a mechanistic perspective. In this investigation, we attempt to elucidate the interfacial corrosion mechanism between Ni-based structural materials, such as Inconel 617, and UCl3-containing salt systems through a multi-modal advanced characterization approach, including electron microscopy techniques. The findings from this investigation will expand the knowledgebase of chloride molten salt corrosion of MSR structural materials for strategic property-to-performance design.
36 - MATERIALS SCIENCE↗
Compatibility of Type 304H Stainless Steel in Static Molten FLiBe for Inertial Confinement Fusion Reactors: Role of Impurities and Redox Control
Molten fluoride salts, such as FLiBe (LiF-BeF2), are promising candidates for tritium breeding and heat transfer in fusion reactors, but corrosion of structural materials remains a major challenge. This study investigates the corrosion behavior of austenitic stainless steel 304H in purified and NiF2-containing FLiBe at 500°C and 600°C, focusing on the effects of impurities and redox control. Exposure to purified FLiBe resulted in the concurrent depletion of Cr, Mn, and Fe, with corrosion at 500 °C dominated by the combined oxide formation and elemental dissolution, while at 600°C elemental depletion was predominant. The addition of a controlled NiF2 impurity significantly accelerated corrosion at both temperatures, demonstrating the sensitivity of 304H to the salt redox state. Beryllium additions were effective in mitigating corrosion for both baseline and NiF2-containing FLiBe; minimal depletion of Cr, Mn, and Fe occurred with Be additions as low as 2.5 mg (147 wppm), and no NiBe intermetallics formed at 5 mg (294 wppm), indicating that small Be inventories can provide substantial protection without deleterious phase formation. Thermodynamic equilibrium and coupled thermodynamic-kinetic analyses at the salt-alloy interface suggested low corrosion rates in systems with limited hydrogen fluoride (HF) generation, highlighting the importance of salt redox control. Estimates for a Be addition rate were calculated for the HYLIFE-II fusion reactor that can mitigate corrosion-induced degradation, assuming complete conversion of tritium to tritium fluoride (TF). Overall, 304H shows reasonable compatibility with FLiBe under optimized redox conditions. These results provide quantitative guidance for material selection and salt management in fusion blanket and heat exchanger systems and motivate validation under flowing, nonisothermal, and irradiated conditions.
Pillai, Rishi [ORNL] (ORCID:0000000243688197)↗
Preliminary Design and Testing of Candidate Coating Materials Applied to Inconel Bubblers for Improved Corrosion Resistance
The US Department of Energy (DOE) Office of Environmental Management Technology Development (EM-TD) has requested Savannah River National Laboratory (SRNL) evaluate ceramic coating materials that can extend the service life of Hanford melter bubblers. The proposed Task will focus on the development of ceramic coating materials for Inconel bubblers to reduce corrosion and increase the service life of the bubblers.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Surprisingly high irradiation-induced defect mobility in Fe 3 O 4 as revealed through in situ transmission electron microscopy
Radiation-induced defects are expected to change the corrosion kinetics when a material is exposed to a corrosive environment. Characterization of the defects forming under irradiation and their ability to move/diffuse is therefore key to understanding how they impact the corrosion kinetics and possibly mechanisms. For that matter, Fe/Fe 3 O 4 interfaces were irradiated using 1 MeV Kr 2+ at temperatures of 273 K (25 °C), 573 K (300 °C) and 773 K (500 °C) to a maximum dose of 10 dpa (as measured in the Fe matrix) in situ in a transmission electron microscope. The evolution of the microstructure was followed as a function of the dose and temperature. The formation of defects was followed in the metal and the oxide as well as at the metal/oxide interface and characterized in terms of size and density. Additionally, the dynamic observation of the spatial distribution of the defects allowed for a qualitative assessment of the mobility of defects in the oxide as a function of temperature, revealing evidence for enhanced mobility even more so than in the Fe matrix. Mechanisms of induced defect mobility in the spinel are discussed to account for the observations.
36 MATERIALS SCIENCE↗
SNAP-8 materials report for July - December 1964
SNAP-8 reactor materials development program - lubricant, coolant, corrosion loop, and structural materials evaluation
STRUCTURAL MATERIAL↗
Multi–length scale characterization of point defects in thermally oxidized, proton irradiated iron oxides
A key for the success of safe nuclear power generation system is to consider structural materials that are economical, meet mechanical property needs, possess good corrosion resistance, and are radiation tolerant. Nevertheless, fundamental insights that elucidate the details of radiation damage on materials corrosion performance are lacking. This includes the behavior of surface oxides which often regulate corrosion. For example, it is unclear how non-equilibrium point defects, oxide structure, mass transport in oxides, and subsequent oxidation behavior are altered by the radiation. Here, in this work, some of the effects of proton irradiation on the attributes of point defects, iron oxide microstructures, and the physical nature of the oxidation product were correlated with corrosion behavior. Iron oxides, fabricated by thermal oxidation in air at 400°C and 800°C for 1 h, were subjected to 200 keV, 0.03 dpa (displacements per atom) of proton irradiation, and subjected to corrosion reactivity assessment using AC and DC electrochemical methods. Experimental methods that target materials information at different length scales, such as positron annihilation spectroscopy (atomistic), transmission electron microscopy (mesoscopic), and electrochemical methods (macroscopic), were coupled to shed light on the impact of radiation-induced defect modifications and structural changes in oxides on corrosion reactivity which ultimately affects durability in harsh environments.
36 MATERIALS SCIENCE↗