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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Diagnosis of PV Cell Antireflective Coating Degradation Resulting from Hot-Humid, High Voltage Potential Aging

Corrosion of the antireflective coating on a photovoltaic cell ("ARc corrosion") has previously been observed in studies using hot-humid test conditions with external high-voltage (HV) bias. This study primarily focuses on known vulnerable, legacy aluminum back surface field (Al-BSF) cells in mini-modules (MiMos) put through comparative stepped stress tests. Each cell type had MiMos at +1500 V, -1500 V, or unbiased ("Voc") potential, which were sequentially subjected to test conditions of 60deg C/60% relative humidity (RH) for 96 h, as in International Electrotechnical Commission (IEC) Technical Specification 62804-1; 70 deg C/70% RH for 200 h; and 85 deg C/85% RH for 200 h. Characterizations at each step included visual camera and electroluminescence (EL) imaging, colorimetry, and current-voltage (I-V) curve tracing. Final characterizations included: Suns-Voc; spatial mapping of external quantum efficiency (EQE); high-resolution photoluminescence (PL), EL, and dark lock-in thermography (DLIT) imaging. Forensics were performed on extracted cores, including scanning electron microscopy (SEM) with energy-dispersive X-ray spectroscopy (EDS), X-ray photoelectron spectroscopy (XPS), and scanning Auger microscopy (SAM). Forensics were also conducted on MiMos from previous studies that underwent stepped HV aging and separate outdoor aged full-sized modules. ARc corrosion was specifically seen for the glass/encapsulant/cell side of the +1500V (HV+) stressed MiMos and modules. Appearance, color, and reflectance were the most distinguishing characteristics relative to glass corrosion, gridline-corrosion and -delamination as well as other concurrent degradation modes. SEM/EDS and SAM identified conversion of silicon nitride to hydrated silica, hydrous silica, or hydrated amorphous silica, which preferentially occurred at the edges and tips of the pyramidal textured cell surface.

14 SOLAR ENERGY↗
DOE OSTI2024

Experimental Operation of a Prototype 750C Advanced Chloride Molten Salt Bellows Valve

To achieve DOE 2030 SunShot targets that reduce the cost of liquid-based solar by an additional 40% to 70% beyond 2018 costs, a more reliable, highly manufacturable flow valve, capable of achieving operational temperatures of >700°C is required [1]. This paper investigates the development of an innovative high-temperature chloride molten salt valve, with operation up to 750°C. This valve is intended to be employed within Gen 3 CSP liquid-based thermal energy storage (TES) systems as well as Gen 4 modular salt reactor (MSR) technologies. This work details the general design and flow testing of a bellows-seal flow control valve (FCV). This design includes an integrated closed-loop thermal control system to ensure robust design for freeze-thaw cycles. The self-contained thermal management STM system, is unique in the salt valve industry since it is an integrated solution to provide a consistent, repeatable alternative to typical heat tracing. Additionally, the design includes the employment of a novel heat pipe valve stem to facilitate enhanced passive thermal management into the valve assembly. This valve stem heat pipe is designed to facilitate natural circulation within the bonnet to ensure robust operation, during both nominal and transient thermal operation. The valve body and trim will be designed using SS316H, consistent with Flowserve Corporation’s existing product base and is code qualified but will utilize clad material for materials corrosion, manufacturing cost reduction and compatibility to ensure design flexibility. A test campaign was performed in this investigation utilizing a novel 750°C ternary chloride (20%NaCl/40%MgCl2/40%KCl by mol. wt. %) molten salt flow loop. A discussion about the design and installation of the valves within this test bed is provided for this investigation. Valve test results from this study assessed Cv curves as well as multiple actuator cycles, under varying thermodynamic and operational mode conditions, which would be characteristic within a commercial molten salt facility. The results indicate nominal operation for the baseline design, though improved performance and reliability is expected with the full designed FCV.

Armijo, Kenneth (ORCID:0000000346832147)↗
DOE OSTI2024

Crack Growth Rate and Fracture Toughness of Stainless Steel Welds in Light Water Reactor Environment

Austenitic stainless steel welds play a crucial role in maintaining the structural integrity of light water reactors. Within the reactor core, stainless steel welds endure a harsh service environment including elevated temperature, corrosive coolant, and intense neutron irradiation. Long-term exposure to such a severe service environment could lead to performance degradation. To ensure the safe and economic operation of light water reactors, this study performed crack growth rate and fracture toughness tests on irradiated and thermally aged weld samples harvested from decommissioned reactor components. In a low-corrosion-potential environment, stainless steel welds showed resistance to stress corrosion cracking, and the measured crack propagation rates were comparable to those of wrought stainless steels under similar test conditions. However, a marked decline in fracture toughness was observed in weld samples, particularly in irradiated samples. Fracture instability was observed at ~0.8 dpa, significantly lower than those observed in wrought stainless steels. Further research at higher doses is needed to evaluate the dose dependence of irradiation-induced embrittlement, and to understand the interplay between thermal aging and neutron irradiation in stainless steel welds.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

SAM Code Enhancements for Modeling of Liquid Metal-Cooled Fast Reactor Concepts

The SAM code is under development and supported by DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. These advanced reactor concepts incorporate novel and improved approaches to achieve safety and economic feasibility. This report summarizes two major efforts in addressing the modeling gaps in SAM for liquid-metal-cooled fast reactor (LMFR) applications, i.e. thermal mixing and stratification phenomena in large pools and corrosion-oxidation of components in flowing lead. A new one-dimensional model for thermal mixing and stratification effects in large pools and enclosures is developed and implemented. Thermal mixing and stratification occur when fluid enters a pool at a temperature different than the bulk fluid itself, a scenario often encountered during transients in pool-type systems. These phenomena are critical for the safety of reactors, impacting phenomena like natural circulation, which is essential for passive cooling. The improved model in SAM addresses limitations of state-of-the-art approaches by combining one-dimensional (1D) channels, representing the coolant jet flow, with lumped-parameter zero-dimensional (0D) pools, representing the rest of coolant in the tank. Energy exchange between the 1D jet and the 0D pools is based on heat transfer correlations calibrated against 3D simulations. It is verified that this model can handle various flow configurations, including hot jets in colder pools, cold jets in hotter pools, and the presence of features like ceilings, free surfaces, and obstacles. Additionally, validation against experimental data demonstrates the ability of the model to capture mixing and stratification effects in a wide range of conditions. The flexibility and improved accuracy of the new model make it a valuable tool for reactor safety analysis, allowing for the simulation of different geometries encountered in advanced reactors. A system-level corrosion modeling capability is developed and implemented in SAM to support Lead Fast Reactor (LFR) development. Although the initial focus of this capability will be on LFR application, this can later be expanded to include other liquid metals such as Lead-Bismuth Eutectic (LBE) and PbLi. This report summarizes the common corrosion mitigation strategies and outlines the progress on implementing and validating a corrosion-oxidation model in SAM. Verification and validation of the corrosion-oxidation portion of the model was performed using analytical solution and measured data from samples tested in the non-isothermal pumped lead loop at IPPE Obninsk. The iron transport and corrosion/precipitation portion of the model was assessed using an analytical model and measured corrosion depths from a natural convection lead loop experiment performed at CEA. It is demonstrated that the model implemented in SAM performed well in these assessments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Experimental Investigations into the Corrosion of Alloy 625 Using NaCl-PuCl3 Molten Salt in a Natural Circulation Microloop

Molten salt reactors (MSRs) can potentially revolutionize the nuclear industry by providing a path to a near-zero nuclear waste fuel cycle, contributing to more sustainable energy sources. As a plethora of MSR developers in the United States work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have limited operational testing with nuclear material. Licensing and operating these reactors require an understanding of corrosion effects on reactor materials of construction under operational conditions. The TerraPower Molten Chloride Fast Reactor (MCFR) is a liquid-fueled chloride-salt fast reactor which has received notable interest from the utility sector based on its desirable economic characteristics. The reactor operates at low pressure but does not require the use of highly reactive chemicals, leading to a reduced use of concrete and steel during construction. Additionally, liquid fuel allows for inherently stable behavior and natural circulation during a loss-of-site-power scenario. MCFR can be refueled while operating which makes it compatible with variable generation sources such as wind and solar. MCFR is a breed-and-burn in-situ reactor that does not implement any chemical processing or separations in the fuel cycle. Only mechanical filtration of noble metals and off-gassing of noble gases are utilized while the actinides stay mixed with the fuel at all times. The MCFR will require technology development to reach commercialization. With a breed-and-burn in-situ reactor like MCFR, the transmutation of fertile U-238 to fissile Pu-239 allows for much greater fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Assessing the Impact of Molten Halide Salts on Creep of Structural Alloys at 650°-750°C

There is a critical lack of data on the mechanical behavior of candidate structural materials for advanced nuclear reactors under molten halide salt environments. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy in molten fluoride salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. The present work aims to assess the role of molten salt corrosion on the creep behavior of three alloys 316H, 617 and 282 at 650-816 °C. Creep tests were conducted in fluoride (FLiNaK) and chloride (NaCl-MgCl2) salts. Initial results from the ongoing testing will be presented which suggest that the molten salt environment caused a 25-50% reduction in creep rupture lifetime compared to air exposures. Physics-based corrosion and creep models were employed to gain some insights into the potential degradation mechanisms.

Pillai, Rishi↗
DOE OSTI2024

IASCC of 304 SS in BWR environments: Effects of post-irradiation annealing and surface condition

To investigate the impact of low-temperature post-irradiation annealing (PIA) on the stress corrosion cracking of neutron-irradiated 304 stainless steel, constant extension rate tests were conducted in simulated boiling water reactor normal water chemistry (BWR-NWC) and hydrogenated water chemistry (HWC) environments. Ten tensile samples, comprising five as-irradiated and five PIA specimens, were prepared by electropolishing the gauge section of electric discharge machined (EDM) samples. Here, the annealing treatment reduced the yield strength from approximately 550 MPa to around 425 MPa and significantly restored the ductility and the strain hardening capability of the alloy. Consequently, the susceptibility of this material to irradiation-assisted stress corrosion cracking (IASCC) was effectively mitigated, which is more prominent in HWC, as evident from fractography, which indicated a decreased propensity for intergranular (IG) fracture. Furthermore, it was observed that the polished surface facilitated crack initiation more readily than the EDM surface, suggesting that the EDM process suppressed crack initiation to some extent.

Crack initiation↗
DOE OSTI2021

A time-dependent chloride diffusion model for predicting initial corrosion time of reinforced concrete with slag addition

The effect of granulated blast furnace slag (GBFS) addition on the threshold chloride concentration (TCC) for rebar corrosion was investigated. A modified diffusion model, coupled with a time-dependent effective diffusion coefficient and surface chloride concentration, was proposed to predict the chloride profile. The corrosion initiation time was estimated based on the model predictions and the measured TCC. The results indicate that adding GBFS decreases the TCC by lowering the pH value of the pore solution. The evolution of corrosion potential and current density is found to obey a 3-parameter Weibull distribution. MIP tests show that adding GBFS contributes to refinement of pore structure by decreasing the fraction of large capillary pores. The time-dependent model exhibits good predictive strength and helps understand how GBFS addition delays the corrosion initiation by retarding the chloride diffusion, though a lower TCC is obtained.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Augmented Monitoring and Condition Assessment Program for SNF Wet Storage Life Extension - 20489

Approximately 27 MTHM of spent nuclear fuel (SNF) owned and managed by the U.S. Department of Energy, Office of Environmental Management is stored in the L Basin at the Savannah River Site (SRS). This 'DOE SNF' is comprised of approximately 12,000 aluminum-clad, aluminum-based fuel assemblies (∼7 MTHM), and approximately 2000 non-aluminum fuel assemblies (∼20 MTHM). A program is in progress to perform Non-Destructive Examination (NDE) of the fuel and their storage containers to characterize the materials' condition, and to evaluate the effects of service to enable continued safe wet storage of the SNF. The Augmented Monitoring and Condition Assessment Program (AMCAP), a two-part program to develop and implement remote underwater Non-Destructive Examination of the aluminum SNF, and of the containers of the non-aluminum SNF, respectively, is aimed at the characterization and evaluation of corrosion degradation of aluminum fuel and container materials. The predominant design of the aluminum SNF (ASNF) stored in L Basin is the Materials Test Reactor (MTR) equivalent design, a plate fuel design. These SNF are stored in 5'' diameter tubes or 5'' x 5'' squares (called bundles) that are nominally 12' or up to 14' long. The SNF is de-bundled and inspected using a custom-designed MTR Fuel Inspection Table. The inspection table provides for indexed fuel positioning for a video camera examination with controlled lighting. A total of 10 of the SNF originating from foreign research and test reactors were selected for examination based on burnup, enrichment, and prior damage caused by service/storage history. A special inspection campaign of these 10 assemblies is in progress. The observed corrosion damage included minor to moderate attack from general corrosion, pitting, crevice, end grain and galvanic corrosion. Example results from the completed inspections are shown and discussed. The focus is a comparison of the as-received condition versus the as-found current storage condition that will serve to validate Water Chemistry and Corrosion Monitoring Programs. The non-aluminum SNF (NASNF) stored in L Basin are of diverse design that includes various geometries with claddings of stainless steel, Zircaloy, and Hastelloy. The fuel core materials include uranium alloys, oxides/mixed oxides, and carbides. These SNF materials, originating from early experimental and test reactors, are in various physical forms including single fuel elements and cut pieces. The fuel is stored in L Basin in various configurations including in direct bundled storage in aluminum tubes and in isolation cans that are in the bundles or in a larger over-size storage container (OSC). Concern with inside-out corrosion and the potential loss of configuration control and ability to handle the storage containers prompted the development of remote NDE methods that include visual and UT technologies to assess the condition of the containers. Candidate stored materials were selected for a special inspection campaign; the development of the NDE methods for the inspection for galvanic, crevice and sediment-induced corrosion (inside-out) are discussed. A summary of the SNF storage in L Basin at the SRS, and an overview of the AMCAP to enable continued safe storage of DOE SNF are described. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Large Eddy Simulation and Analysis of Truck Generated Droplet Spray Transport at Bridges

Weathering steel is a particular type of steel that develops a higher level of atmospheric corrosion protection than non-weathering grades of steel. While uncoated weathering steels have the appearance of being “rusted”, there are many forms of ferric oxyhydroxides (aka. rust) that can form on steel. The predominant ferric oxyhydroxide that gives uncoated weathering steel its tightly adherent patina is goethite. The atmospheric corrosion protection comes from alloying, primarily with a copper composition greater than 0.20 percent and also small amounts of chrome, nickel, and silicon. Bare atmospheric corrosion resistant steels have been recognized as far back as 1905 with very early editions of the American Society for Testing and Materials (ASTM) A 7 “Standard Specification for Steel for Bridges and Buildings” [1]. Uncoated weathering steel (UWS) is the lowest cost corrosion protection system possible for a steel bridge based on the first cost. Provided the UWS is used in an optimal environment that allows the protective patina to form, UWS also provides the lowest life cycle cost. The protective patina on UWS does not form or degrades in continuously damp conditions, forming a higher proportion of lepidocrocite, or in the presence of a concentration of chlorine ions that is too high, forming a higher proportion of akageneite.

36 MATERIALS SCIENCE↗
DOE OSTI2025

The effect of copper additives on hybrid Zr-based chemical conversion coating morphology and chemical compositions

In the realm of corrosion protection, Zr-based conversion coatings offer an environmentally friendly, chromate-free alternative to conventional coating. This study uses advanced X-ray, electron microscopy and electrochemical testing techniques to better understand the impact of varying Cu²⁺ ion concentrations on the characteristics of Zr-based coatings on Fe substrates. Our findings demonstrate that within the tested conditions, higher Cu²⁺ additive concentrations at 40 ppm enhance surface characteristics, increasing stability towards anti-corrosion capability, particularly under NaCl treatment. Conversely, at a lower Cu²⁺ concentration of 20 ppm, coatings exhibited more significant dissolution of Cu clusters and increased vulnerability to chloride-induced degradation. X-ray photoelectron spectroscopy and synchrotron X-ray fluorescence (XRF) and X-ray absorption near edge structure (XANES) spectroscopy analyses revealed Cu²⁺ formation in Cu20, while Cu40 retained its metallic state (Cu⁰) with a slower reaction rate. Cu20 offers some protection but lacks durability, whereas in Cu40, lower I corr , and enhance corrosion resistance, making it ideal for protective coatings in salt-water environments. This study underscores the importance of balancing Cu²⁺ ions concentration in the coating solution to optimize performance, highlighting the role of Cu in enhancing both surface properties and long-term stability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Effect of V Content on Corrosion Behavior of Al-V Alloys Produced by Mechanical Alloying and Subsequent Spark Plasma Sintering

Al-V alloys produced via high-energy ball milling have been reported to show simultaneous improvement of corrosion resistance and mechanical properties compared to traditional Al alloys. In these alloys, V content plays a crucial role in increasing or decreasing the corrosion resistance. Therefore, the effect of V and microstructure on corrosion of high-energy ball milled and subsequently spark plasma sintered Al-xV alloys (x = 2, 5, 10 at%) has been studied. Cyclic potentiodynamic polarization tests and electrochemical impedance spectroscopic analysis revealed the increment of V content up to 5 at% enhanced the corrosion resistance of the alloy. However, highly heterogeneous microstructure in Al-10 at%V resulted in significant localized corrosion over the immersion time. The electrochemical impedance spectroscopy studies over 14 days of immersion revealed underlying corrosion mechanisms.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Distinct Mechanism of Anti-Corrosion and Swelling-Adhesion Modeling of Low-Dimensional Nylon-Fluoropolymer Composite Coatings

Compared to other polymers, composite coatings with fluoropolymers as the matrix have attracted considerable interest due to their mechanical performance, chemical stability, and low surface energy. Herein, we present an anticorrosion mechanism of fluoropolymer composite coatings achieved by dispersing varying amounts of polyamide 12 (PA-12) particles over a poly(vinylidene fluoride-co-hexafluoropropylene) (PVDF-HFP) matrix. The choice of PVDF-HFP as a coating matrix was driven by its superior processability and mechanical properties over Teflon, while PA-12 afforded swelling capacity, preventing electrolyte permeation, and inhibiting matrix stress crack propagation. The corrosion resistance of the coatings was assessed mainly by potentiodynamic polarization and impedance measurements while being immersed in a NaCl solution. Our electrochemical measurement findings showed that 0.75% w/w PA-12 in the PVDF-HFP matrix significantly enhanced corrosion protection even after 21 days of immersion. Microscopy, dielectric spectroscopy, surface analysis, and mechanical testing (American Society for Testing and Materials standards) corroborated the results. PVDF-HFP coatings containing a minimum percolation threshold of 0.75 wt % PA-12 with 1 mil thickness on mild steel exhibited a remarkable increase in resistance and film adhesion, and a notable corrosion rate decrease. Finite element analysis simulation with a bilinear traction-separation law confirmed the swelling mechanism and adhesion behavior. Finally, this study highlights the potential of nylon particle dispersion in PVDF-HFP matrices as a practical approach to developing advanced anticorrosion coatings with promising practical applications in various industries.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Considerations of Radiation-Hardened Electronics for Alpha Detectors in Molten Salt Reactors

Along with advanced generations of reactors quickly becoming reality, the need for instrumentation to provide reliable information for their safety and operation is also becoming relevant. Instrumentation-wise, molten salt reactors (MSRs) represent an especially challenging case in which detectors must survive extreme temperatures, high amounts of radiation, and a highly corrosive coolant. Adding to the complexity, tight material control and accounting must be maintained with any special nuclear material (SNM), a practice made more difficult by the nature of a loop of flowing fuel. Alpha spectroscopy offers a unique window into investigating the isotopic concentrations of actinides within just such a molten salt environment. The characteristic α emissions of specific actinide isotopes in the salt enable quantification of their concentrations. Semiconductor devices based on wide bandgap materials such as silicon carbide and gallium oxide can withstand the extreme temperatures within MSRs and offer good chemical stability with the corrosive salt. Collaborators at The Ohio State University are fabricating and testing a β gallium oxide α particle sensor that could identify actinide concentrations within the fuel salt to help improve the material control and accounting required for complicated system inherent to MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Predicting Rupture Pressure and Leak Rate of Alloy 600 Tubes with ODSCC and Comparing with Test Results (Technical Letter Report)

This report summarizes structural integrity work performed for the U.S. NRC that predicts the rupture pressure and leak rate of flawed steam generator (SG) tubes. The work includes pressure testing of straight tubes with lab-induced outer diameter stress corrosion cracks (ODSCC). Existing models for predicting rupture pressure and leak rate of straight tubes with well-defined, rectangular, electro-discharge-machined (EDM) flaws were modified and used to predict the measured rupture pressures and leak rates from straight tubes with ODSCC. Models previously developed for EDM notches in straight tubes were modified for ODSCC by using the equivalent rectangular crack (ERC) method.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Estimating Internal Stress of an Alteration Layer Formed on Corroded Boroaluminosilicate Glass through Spectroscopic Ellipsometry Analysis

Aqueous corrosion of glass may result in the formation of an alteration layer in the glass surface of which chemical composition and network structure are different from those of the bulk glass. Since corrosion occurs far below the glass-transition temperature, the alteration layer cannot fully relax to the new structure with the lowest possible energy. Molecular dynamics simulations suggested that such a network will contain highly strained chemical bonds, which can be manifested as a stress in the alteration layer. Common techniques to measure stress in thin films or surface layers were found inadequate for thick monolithic glass samples corroded in water. Here, we explored the use of spectroscopic ellipsometry to test the presence of internal stress in the alteration layer formed by aqueous corrosion of glass. Furthermore, a procedure for analyses of spectroscopic ellipsometry data to determine birefringence in the alteration layer was developed. Findings with the established fitting procedure suggested that a stress builds up in the corroded surface layer of a boroaluminosilicate glass if there is a change in relative humidity, pH, or electrolyte concentration of the environment to which the glass surface is exposed. A similar process may occur in other types of glass, and it may affect the surface properties of corroded glass objects.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Hanford Double Shell Waste Tank Corrosion Studies FY2020 (Final Report)

For fiscal year (FY) 2020, the Savannah River National Laboratory (SRNL) focused on three experimental tasks related to Hanford Double Shell Tank (DST) chemistry and integrity. The first task focused on understanding risk of corrosion due to formation of either continuous layers or discrete patches of deposits on the tanks’ inner sidewalls and bottoms. The second task studied vapor space corrosion (VSC) of the secondary liner using two commercially available vapor corrosion inhibitors (VCIs) at manufacturer recommended dosages. Finally, the third task investigated the long-term open circuit potential (OCP) drift for mill-scale coupons, coupons with polished surfaces, and partial mill-scale coupons that were exposed to tank waste simulants. Additional tests were performed to evaluate the OCP drift as a function of the quantity and composition of organic compounds added to the simulant. A test that was included in the matrix that established the new chemistry limits for tank corrosion, but was previously unreported, was presented and reviewed as well.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Corrosion susceptibility and chromium loss in Austenitic steels and Nickel-based alloys in molten FLiNaK at 700 °C

A comparative study was conducted to evaluate the corrosion susceptibility of 316L, 316H, Ni 200, Inconel 625, and Hastelloy N in molten FLiNaK at 700 °C for 100 h. Top-view and cross-sectional scanning electron microscopy (SEM) imaging, combined with energy-dispersive X-ray spectroscopy (EDS) mapping, was performed to investigate microstructural and compositional changes. SEM images were further processed by introducing a contrast threshold to map cavity distribution. Using EDS mapping, intergranular and intragranular Cr loss were separately characterized. The ranking of mass loss after corrosion, from highest to lowest, is as follows: 316H > 316L > Inconel 625 > Hastelloy N > Ni 200. Cr loss was found to be correlated with Mo concentration, in agreement with many previous studies, with higher Mo content resulting in reduced Cr loss. On the other hand, there is no evidence that the high carbon content in 316H enhances corrosion resistance. The relatively low solubility of carbon in austenite at the testing temperature limits the amount of dissolved carbon. Therefore, the carbon-retarded vacancy diffusion, and consequently the reduced Cr diffusion, were not observed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗