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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Crack Growth Rate and Fracture Toughness Tests on Irradiated Ex-Plant Materials

The performance of structural materials is critical for the safe and economic operation of light water reactors. Exposed to neutron irradiation during service, the reactor core internal materials can undergo significant microstructural and microchemical changes, leading to irradiation hardening and embrittlement. To ensure the structural integrity and functionality of nuclear reactor components during long-term operation, material degradation and damage mechanisms must be understood and managed adequately. In this work, irradiated materials harvested from the decommissioned Zorita reactor were studied for their cracking susceptibility and fracture resistance as a function of irradiation dose up to 47 displacement per atom (dpa). The material is a Type 304 stainless steel sectioned from the baffle plates of this pressurized water reactor with 38 years of service. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed in low-corrosion-potential environments at ~315°C. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant stress intensity factors, most samples did not show elevated crack growth rates, suggesting an adequate stress corrosion cracking resistance in the test environments. However, an unstable cracking behavior was observed in a 47-dpa sample, resulting in significantly higher crack growth rates than expected at high stress intensity factors. Crack instability was also observed in a 0.06-dpa sample but did not lead to a sustained high crack growth rate. The impact of neutron irradiation was more evident in the fracture toughness J-R curve tests. As the dose increased, the J-R curve declined considerably and became very shallow at high doses. A fully intergranular fracture morphology was also observed among the high-dose samples ruptured in an air atmosphere at room temperature. This brittle fracture mode in the absence of high temperature water environment confirmed a high degree of embrittlement of this material resulting from its service exposure to neutron irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Biomembrane Structure and Material Properties Studied With Neutron Scattering

Cell membranes and their associated structures are dynamical supramolecular structures where different physiological processes take place. Detailed knowledge of their static and dynamic structures is therefore needed, to better understand membrane biology. The structure–function relationship is a basic tenet in biology and has been pursued using a range of different experimental approaches. In this review, we will discuss one approach, namely the use of neutron scattering techniques as applied, primarily, to model membrane systems composed of lipid bilayers. An advantage of neutron scattering, compared to other scattering techniques, is the differential sensitivity of neutrons to isotopes of hydrogen and, as a result, the relative ease of altering sample contrast by substituting protium for deuterium. This property makes neutrons an ideal probe for the study of hydrogen-rich materials, such as biomembranes. In this review article, we describe isotopic labeling studies of model and viable membranes, and discuss novel applications of neutron contrast variation in order to gain unique insights into the structure, dynamics, and molecular interactions of biological membranes. We specifically focus on how small-angle neutron scattering data is modeled using different contrast data and molecular dynamics simulations. We also briefly discuss neutron reflectometry and present a few recent advances that have taken place in neutron spin echo spectroscopy studies and the unique membrane mechanical data that can be derived from them, primarily due to new models used to fit the data.

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

Experimental Examination of Additively Manufactured Patterns on Structural Nuclear Materials for Digital Image Correlation Strain Measurements

Abstract Background There are a limited number of commercially available sensors for monitoring the deformation of materials in-situ during harsh environment applications, such as those found in the nuclear and aerospace industries. Such sensing devices, including weldable strain gauges, extensometers, and linear variable differential transformers, can be destructive to material surfaces being investigated and typically require relatively large surface areas to attach (> 10 mm in length). Digital image correlation (DIC) is a viable, non-contact alternative to in-situ strain deformation. However, it often requires implementing artificial patterns using splattering techniques, which are difficult to reproduce. Objective Additive manufacturing capabilities offer consistent patterns using programmable fabrication methods. Methods In this work, a variety of small-scale periodic patterns with different geometries were printed directly on structural nuclear materials (i.e., stainless steel and aluminum tensile specimens) using an aerosol jet printer (AJP). Unlike other additive manufacturing techniques, AJP offers the advantage of materials selection. DIC was used to track and correlate strain to alternative measurement methods during cyclic loading, and tensile tests (up to 1100 µɛ) at room temperature. Results The results confirmed AJP has better control of pattern parameters for small fields of view and facilitate the ability of DIC algorithms to adequately process patterns with periodicity. More specifically, the printed 100 μm spaced dot and 150 μm spaced line patterns provided accurate measurements with a maximum error of less than 2% and 4% on aluminum samples when compared to an extensometer and commercially available strain gauges. Conclusion Our results highlight a new pattern fabrication technique that is form factor friendly for digital image correlation in nuclear applications.

Novich, K. A. (ORCID:0000000204466022)↗
DOE OSTI2025

Greenhouse gas emissions of global construction material production

Abstract Global production of building materials is a primary contributor to greenhouse gas (GHG) emissions, but the production of these materials is necessary for modern infrastructure and society. Understanding the GHG emissions from building materials production in the context of their function is critical to decarbonizing this important sector. In this work, we present estimates of global production, approximate ranges of GHG emissions, and ranges of material properties of 12 critical building material classes to provide a unified dataset across material types. This dataset drew from industry analyses of production and emissions, ranges of emission factors within a material type, and broad reporting of thermal and mechanical properties to compare both within and between material types. Globally, in 2019, we estimate 42.8 Gt of these 12 materials were produced, with 38.6 Gt used in the building and construction industry. As a result of this production, 9.3 Gt of CO 2 was emitted, or 25% of global fossil GHG emissions, with 5.8 Gt CO 2 (16% of global GHG emissions) due to materials used in construction applications. Both construction material production and emissions are primarily driven by structural materials, such as concrete and steel. Material selection can play a key role in reducing emissions in the context of the function, with variation in emissions of structural materials per unit strength between 0.001–0.1 kg CO 2 /kg/MPa and in insulation materials per R -value/thickness of 0.018–0.14 kg CO 2 /kg/(K⋅m 2 W −1 ))). The developed dataset can play a key role in supporting decision-making in materials by providing a unified source for examining emissions, material properties, and quantity of material produced.

Kane, Seth (ORCID:0000000269401369)↗
DOE OSTI2025

Prioritizing Nuclear Materials for SAM-3 Neutron Irradiation Campaign: Structural and Cladding Materials Candidates

This report outlines a framework for selecting structural and cladding materials for the Nuclear Science User Facilities (NSUF) SAM-3 neutron irradiation campaign to support the advancement of nuclear energy technologies. The document begins with an introduction that provides background context, highlights the motivations for launching a new irradiation campaign, and defines the overall objectives. The core of the report describes the design considerations for the irradiation campaign, including capsule configurations, irradiation temperature ranges, and target dose levels (defined by displacements per atom, or dpa). The material recommendation was guided by the Specimen Identification and Prioritization (SIP) Working Group, a multidisciplinary team of experts representing national laboratories, academia, industry, federal government and agency. This group played a central role in identifying candidate materials, evaluating technical justifications, and ensuring alignment with boarder programmatic goals. A detailed set of criteria for material prioritization is then presented, taking into account reactor relevance, performance gaps, advanced manufacturing methods, and emerging material classes. Based on the input of SIP working group, specific materials were selected and justified for inclusion in the irradiation campaign by the NSUF leadership and its U.S. Department of Energy (DOE)-Office of Nuclear Energy (NE) management. The final section provides recommended capsule designs, summarizing critical parameters such as material type, fabrication method, sample geometry, irradiation conditions, and specimen quantities. This report serves as a foundation for executing a focused and high-impact neutron irradiation campaign aimed at addressing key materials challenges for both existing and advanced nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Enhancing fatigue life by ductile-transformable multicomponent B2 precipitates in a high-entropy alloy

Catastrophic accidents caused by fatigue failures often occur in engineering structures. Thus, a fundamental understanding of cyclic-deformation and fatigue-failure mechanisms is critical for the development of fatigue-resistant structural materials. Here we report a high-entropy alloy with enhanced fatigue life by ductile-transformable multicomponent B2 precipitates. Its cyclic-deformation mechanisms are revealed by real-time in-situ neutron diffraction, transmission-electron microscopy, crystal-plasticity modeling, and Monte-Carlo simulations. Multiple cyclic-deformation mechanisms, including dislocation slips, precipitation strengthening, deformation twinning, and reversible martensitic phase transformation, are observed in the studied high-entropy alloy. Its improved fatigue performance at low strain amplitudes, i.e., the high fatigue-crack-initiation resistance, is attributed to the high elasticity, plastic deformability, and martensitic transformation of the B2-strengthening phase. This study shows that fatigue-resistant alloys can be developed by incorporating strengthening ductile-transformable multicomponent intermetallic phases.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Adsorption of methylene blue by bismuth impregnated biochar of different biomass materials: pore structure and surface chemistry characteristics

Abstract In this study, three different types of biochar (biomass materials were beech wood, corncob, and bamboo) were loaded with bismuth to form bismuth impregnated biochars (named respectively as Bi@WBC, Bi@CBC, and Bi@BBC) to investigate their properties (pore structure and surface chemistry characteristics) on methylene blue (MB) removal capacities. Three bismuth impregnated biochars made from different biomass materials were ranked for their adsorption capacity of MB as follows: Bi@WBC > Bi@CBC > Bi@BBC, in which Bi@WBC showed the largest Langmuir adsorption capacity of 178.59 mg g −1 . While the specific surface area (SSA) and pore volume of Bi@WBC were the largest (accounting for 477.74 m 2 g −1 and 0.15 cm 3 g −1 , respectively), according to the FTIR and XPS results, Bi@WBC (72.3%) exhibited a higher proportion of BiO bonds than the Bi@CBC (52.3%) and Bi@BBC (48.4%) did. Conclusively, the rich porous structure and surface functional groups of Bi@WBC affected the adsorption capacity of MB, and the chemical adsorption provided the main contribution. © 2022 Society of Chemical Industry (SCI).

Li, Peiyi↗
DOE OSTI2025

Skyrmion-like Spin Textures Emerging in the Material Derived from Structural Frustration

Magnetic materials with complex spin textures present both fundamental and practical appeal. The complex patterns of magnetic moments emerging on underlying crystal lattices hold potential for robust information storage and processing, including the promise of topological quantum computing. The scope of materials that host such patterns, however, remains rather limited. Here, in this study, we report a discovery of a complex spin texture in a noncentrosymmetric material that emerges from the structural frustration at the boundary between centrosymmetric parent structures MnCoGe (the hexagonal Ni 2 In or the orthorhombic TiNiSi structure type) and MnCoAs (the TiNiSi structure type). Our findings demonstrate that such structural frustration provides a powerful handle for identifying compositional spaces where complex magnetic behavior and associated nontrivial magnetic structures are likely to emerge. Thus, the new phase MnCoGe 1/3 As 2/3 exhibits a modulated cycloidal antiferromagnetic arrangement of electron spins on a noncentrosymmetric lattice (of the hexagonal ZrNiAl type) that materializes in the space between centrosymmetric collinear ferromagnets. This work provides a pathway for discovering novel materials with exotic spin textures for next-generation spintronics and quantum technologies.

Wang, YiXu [Florida State University, Tallahassee,↗
DOE OSTI2022

[Presentation Slides] Code-to-Code Benchmark Study for Thermal Stress Modeling and Preliminary Analysis of the High-temperature Single Heat-Pipe Experiment

In the heat-pipe-cooled microreactor, heat pipes remove heat from the reactor core as a passive heat-transfer device, so the fluid circulation is not required for cooling, which can substantially simplify the overall reactor design. However, given the extremely high temperatures in the core region and potentially large temperature gradients across the structure materials, thermal stresses need to be well-analyzed to ensure structural integrity during normal operations and accident scenarios. This presentation slides discuss finite element method-based thermal-stress analysis for the high-temperature single heat-pipe test article in the Single Primary Heat Extraction and Removal Emulator (SPHERE) facility at Idaho National Laboratory (INL), using two commercial software packages, Abaqus and Star-CCM+. A code-to-code benchmark study was performed to crosscheck the model setup and capability of each code and to gain preliminary insights into the potential thermal stress concerns from the current experimental setup. It is observed the significant thermal stresses happen at the inner surface of the heat pipe hole surrounded by electric heaters where the largest temperature gradients appear. The temperature fields have good agreements between Abaqus and Star-CCM+, while the induced thermal stresses show modest deviations probably due to differences of meshing engines used in these two codes. It is found that the local maximum thermal stresses may reach close to the ultimate tensile strength and yield strength of structural material depending on the heater power. Ultimately, the coupled thermal-structural analysis will help guide the current experimental plan and ensure the facility safety for future experimental study.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Fracture toughness evaluations for spent nuclear fuel dry storage canister welds and spent nuclear fuel clad-pellet structures

Integrity of spent nuclear fuel (SNF) interim storage canisters is very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method was used to estimate the canister weldment fracture toughness. SNTT was developed to measure the intrinsic fracture toughness of structural materials using small specimens. The SNTT method has been applied to a wide variety of structural materials, such as low-alloy steels, stainless steel, aluminum alloy, ceramics, concrete, and composites. The SNTT system operates by applying pure torsion to cylindrical specimens with a notch line that spirals around the specimen at a 45° pitch. In order to carry out pure torsion load mode, biaxial tension/torsion tester was developed accordingly to perform SNTT protocol. Moreover, applying fracture mechanics approach to SNF system reliability investigation is warranted due to the inherent flaws and hydride structures existed in a SNF system after nuclear reactor operation. However, none of the existing fracture toughness data deal with fuel cladding specific geometry or spent fuel material conditions, such as cladding structure with the pellet-inserts and the associated pellet clad mechanical interactions induced mixed-mode damage mechanisms. Thus, the development of an intrinsic fracture mechanics approach that is suitable for SNF materials is needed. Furthermore, due to thin wall and small dimension of clad tubing structure, the SNTT method was used to estimate the clad tubing structure fracture toughness. Fracture testing were performed on the received stainless steel canister weldment, most SNTT weld samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness J Q ’ for the baseline SS304 steel is at 283 kJ/m². The estimated JQ’ for the SS304/308 weld and baseline metals are 148 kJ/m 2 and 459 kJ/m 2 , respectively. Out of cell fracture testing for spent fuel structure were carried out on the surrogate rods made of Zr-4 clad and alumina inserts, the estimated fracture toughness values for baseline Zr-4 cladding with alumina-pellet inserts are: (1) For SNTT samples with a short or medium crack length, between 5.4-mm and 8-mm, the estimate J IQ upon fracture initiation for the baseline Zr-4 cladding is at 50 kJ/m 2 with 2-sigma uncertainty of 3.26 kJ/m 2 , and the associated K IQ is at 67.46 MPa$\sqrt{m}$; and (2) For SNTT samples with a long crack length, around 13-mm, the crack initialization is deviated from that of the Mode-I tensile fracture and appears to be a mixed-mode fracture of Mode I - tensile stress and Mode III - out of plane shear stress; the estimated J MQ is at 18.9 kJ/m 2 , the associated K MQ is at 41.4 MPa$\sqrt{m}$.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Molten Salt Corrosion Tests of Additively Manufactured Stainless Steel 316H

Molten salt reactors (MSRs) have drawn considerable interest due to their favorable safety features, high thermal efficiency, and compatibility with different fuel cycles. Yet, the success of MSRs hinges critically on the performance of structural materials to be used in these aggressive molten salt environments, where corrosion and material compatibility remain primary challenges to long-term reliability. Additively manufactured (AM) nuclear structural materials prompt the use of novel geometries and compositions to enhance material performance and reduce costs of constructing MSRs. The rapid solidification conditions inherent to AM processing impart distinctive microstructural features, including cellular sub-structures, dislocation densities, residual stress, and oxide inclusions, which can influence material performance in MSR components. While the mechanical properties of AM stainless steels have been widely studied, their corrosion behavior, particularly in molten salt environments, has received far less attention. Addressing these needs, the Advanced Materials and Manufacturing Technologies (AMMT) program provides a framework for systematically evaluating how unique microstructures produced by AM processes influence the performance of these materials in these demanding environments and for developing reproducible testing workflows that can support future code qualification efforts and standards development. Bridging this knowledge gap is essential for assessing the viability of AM alloys in MSRs and informing qualification strategies. A further challenge is the absence of standardized protocols for molten salt corrosion testing. Accordingly, this report provides an account of the corrosion evaluation of AM 316H stainless steel in NaCl 2 -MgCl 2 molten salt at 550 °C, with exposure times of 100 and 500 hours. It documents the experimental procedures implemented under the AMMT program, including salt preparation, exposure protocols, and post-test characterization methods, to establish reproducibility and transparency. Importantly, the study examines AM 316H samples in the as-fabricated condition, directly reflecting the surface state most relevant to engineering applications, and compares their behavior to machine-cut surfaces. Overall, preliminary evaluations have noted that surface conditions (e.g. morphology, contamination, etc.) have a noticeable impact on the corrosion resiliency. The impact of the corrosion is difficult to detect at 100h, unless, in the case of AM 316H, the specimen surface is decontaminated. After 500 h, as-fabricated surfaces of AM and wrought 316H display evidence of general versus preferential corrosion attack, respectively. Both AM and wrought 316H machine-cut surfaces exhibit a continuous Cr depletion zone, evident of general corrosion. While the estimated extent of corrosion appears within the same order of magnitude regardless of the surface condition, it is apparent that more predictable behavior is observed on machine-cut surfaces. Nonetheless, further investigation is necessary to fully elucidate the corrosion mechanism under these conditions.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

History, present status, and future directions of vanadium alloys for fusion reactors

Since the need for low-activation fusion reactor materials was recognized in the mid-1980s, vanadium alloys have been researched and developed as promising candidates for blanket structural materials. Vanadium alloys are non-ferromagnetic and ductile materials and thus have advantages different from other candidates. However, since vanadium alloys are still at a premature industrial stage, research and development have been carried out not only on issues specific to fusion reactors but also on those related to industrial materials. Blankets using vanadium alloys as structural materials and liquid Li as a tritium breeder and coolant (self-cooled V/Li blankets) have unique characteristics, and research and development of issues specific to V/Li blankets have been conducted in conjunction with the development of vanadium alloys. In this paper, the progress of this research and development is reviewed, and the remaining issues are pointed out. As a result, the future direction of vanadium alloy research and development is discussed, considering recent changes in the environment surrounding fusion energy.

Liquid Li blanket↗
DOE OSTI2022

IER 517: Molybdenum Critical Experiment Design [Slides]

In nuclear fuels, U-Mo metallic fuels are used in new space reactor designs and new research reactor fuels. MITR, MURR, and NBSR reactors plan on converting from different HEU fuels to a U-Mo HALEU fuel. In spend nuclear fuel, 95 Mo is one of the 15 main absorbing fission products in irradiated LWR fuel. This makes it important for criticality safety studies in transportation and reprocessing. In structural materials. Molybdenum is found commonly in alloys that make up the structural materials of nuclear reactors such as type 316 SS. Molybdenum helps to improve high temperature performance and corrosion resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Revisiting Multi-Material Composite Structures with Homogenized Composite Properties

Composite structures inherently develop residual stresses during their curing process. Driven predominately by mismatched thermal strains between differing materials or ply orientations, but also affected by curing process phenomena like polymer shrinkage, these residual stresses can lead to failure within composite structures. There are several methods varying in complexity that can be used to model the development of residual stresses, all of which are capable of capturing sufficient detail to understand the residual stress state at the ply level. However, explicitly modeling all plies of a layup in a composite structure can be prohibitively expensive based on the number of plies, structure size, and required element size. The computational cost can be reduced through the homogenization of the composite layup without losing much fidelity of the overall response of the structure. The homogenization process reduces the many plies of a laminate to a single lamina that reduces complexity and increases the mesh size where a single element can span multiple plies. This report focuses on verification and validation efforts for a homogenization process using a suite of finite element simulations rather than an analytic solution derived from classical laminate theory. Initial verification using representative element volumes indicated there was minimal error in the homogenization process; however, this compounded to a small, but acceptable error in strip and split ring experimental composite structures. The error does under predict the residual stress state in the strip and split ring and should be accounted for when simulating composite structures with homogenized properties.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Grizzly Development for Light Water Reactor and Advanced Reactor Applications in Fiscal Year 2021

This report summarizes work performed during Fiscal Year (FY) 2021 at Idaho National Laboratory (INL) for the U.S. Department of Energy’s Nuclear Engineering Advanced Modeling and Simulation (NEAMS) program for the Structural Materials and Chemistry Technical Area in the work package entitled “MS-21IN050106 - Structural Materials - INL.” This effort mainly focused on development and application of capabilities for engineering-scale analysis of nuclear reactor structural components in the Grizzly and BlackBear codes. These efforts include performance improvements and application of models for high-temperature component response, cluster dynamics modeling of precipitation in light water reactor pressure vessel steel, engineering-scale capability development in the areas of concrete simulation and reactor pressure vessel analysis, and preparation for and conducting an independent assessment of the adherence of these codes to software quality standards.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Strain-Induced asymmetry and on-site dynamics of silicon defects in graphene

In the last decade, the atomically-focused electron beams utilized in scanning transmission electron microscopes (STEMs) have been shown to induce a broad set of local structural transformations in materials, opening pathways for directing material synthesis and modification atom-by-atom. The mechanisms underlying these transformations remain largely unknown, due to the intractability of modeling the myriad of reaction pathways that can be accessed through high-energy electron scattering. The information on materials’ structure and dynamics that can be extracted from STEM images is similarly left underexplored. Here, we report the observation of anomalous on-site dynamics of individual silicon impurity atoms in graphene during STEM imaging. Density functional theory-based structural optimizations of anisotropically-strained molecular nanographenes reveal two distinct (but nearly degenerate) stable structures for four-fold coordinated silicon impurities, where interconversion between the two structures manifests slight changes of the silicon position within the lattice site. Implications for defect-based strain engineering in graphene are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

In Situ Transmission Electron Microscopy: Signal processing challenges and examples

Transmission electron microscopy (TEM) is a powerful tool for imaging material structure and characterizing material chemistry. Recent advances in data collection technology for TEM have enabled high-volume and high-resolution data collection at a microsecond frame rate. Here, taking advantage of these advances in data collection rates requires the development and application of data processing tools, including image analysis, feature extraction, and streaming data processing techniques. In this article, we highlight a few areas in materials science that have benefited from combining signal processing and statistical analysis with data collection capabilities in TEM and present a future outlook on opportunities of integrating signal processing with automated TEM data analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) 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. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗