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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Perspectives on Quenching and Tempering 4340 Steel

Steels are ubiquitous due to their affordability and the landscape of useful properties that can be generated for engineering applications. But to further expand the performance envelope, one must be able to understand and control microstructure development by alloying and processing. In this paper we use multiscale, advanced characterization to better understand the structural and chemical evolution of AISI 4340 steel after quenching and tempering (Q&T), including the role of quench rate and short-time, isothermal tempering below 573 K (300 °C), with an emphasis on carbide formation. We compare the microstructure and/or property changes produced by conventional tempering to those produced by higher temperature, short-time “rapid” tempering. We underscore that no single characterization technique can fully capture the subtle microstructure changes like carbon redistribution, transition carbide and/or cementite formation, and retained austenite decomposition that occur during Q&T. Only the use of multiple techniques begins to unravel these complexities. After controlled fast or slow quenching, η transition carbides clearly exist in the microstructure, likely associated with autotempering of this high martensite start temperature (Ms) steel. Isothermal tempering below 598 K (325 °C) results in the relief of carbon supersaturation in the martensite, primarily by the formation of η transition carbides that exhibit a range of carbon levels, seemingly without substitutional element partitioning between the carbide and matrix phases. Hägg transition carbide is present between 300 °C and 325 °C. After conventional tempering at or above 598 K (325 °C) for 2 h, cementite is predominant, but small amounts of cementite are also present in other conditions, even after quenching. Previous work has indicated that silicon (Si) and substitutional elements partition between the cementite, which initially forms under paraequilibrium conditions, and the matrix. Phosphorous (P) may also be preferentially located at cementite/matrix interfaces after high temperature tempering. Slower quench rates result in greater amounts of retained austenite compared to those after fast quenching, which we attribute to increased austenite stability resulting from “autopartitioning”. Rapid, high temperature tempering is also found to diminish tempered martensite embrittlement (TME) believed to be associated with the extent of austenite decomposition, resulting in mechanical properties not attainable by conventional tempering, which may have important implications with respect to industrial heat treatment processes like induction tempering. Controlling the amount and stability of retained austenite is not only relevant to the properties of Q&T steels, but also next-generation advanced high strength steels (AHSS) with austenite/martensite mixtures.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Temperature-Dependent Thermal and Mechanical Properties of a Wire Arc Additively Manufactured Low Transformation Temperature Steel

Recent research has studied the use of low transformation temperature (LTT) martensite steel as feedstock for wire arc additive manufacturing (WAAM) and low tensile residual stresses or compressive residual stresses were detected in the printed walls. These residual stress states help to improve printed product properties such as fatigue strength and corrosion resistance. However, the thermal and mechanical properties of WAAM printed LTT martensite steel walls are largely unknown. In this work, a printed LTT martensite steel was characterized for its thermal, metallurgical, and mechanical behavior at room and elevated temperatures. The temperature-dependent specific heat capacity, thermal expansion, atomic lattice spacing, and tensile properties were measured during both heating and cooling and related to observed microstructural features and computational thermodynamics predictions. These results revealed a large hysteresis in the martensitic transformation, with a martensite start temperature of 240 °C and austenite start temperature of 680 °C. Additional thermal cycles and specimen orientation did not affect the printed specimen austenite and martensite transformations. However, it was observed that the printed metal may exhibit tempering embrittlement at about 350 °C but further studies are needed to confirm that. Further, these results suggest that a temperature control of 250 °C to 350 °C during WAAM is needed to maximize the stress reduction potential of the LTT250 martensite steel. Opportunities for future implementation of LTT martensite steels and optimization of additive manufacturing process conditions are identified.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High-Temperature Oxidation of Steels in Direct-Fired CO 2 Power Cycle Environments

We report future direct-fired supercritical CO 2 power cycles require steels resistant to oxidation/corrosion in high-temperature CO 2 environments containing various impurities. Herein we studied the oxidation behavior of 14 candidate steels in a simulated direct-fired CO 2 power cycle environment consisting of 95% CO 2 , 4% H 2 O, 1% O 2 with/without 0.1% SO 2 at 1 atm and 550 °C, 600 °C, 650 °C for up to 2500 hours. Steels with ≥11.5 wt% Cr exhibited at least partial coverage by Cr-rich oxide scales leading to a significant decrease in the oxidation rates in both gases. While SO 2 had little effect on low-Cr steels that formed Fe-rich oxides, it generally worsened performance of high-Cr (> 11.5 wt%) steels by hindering the establishment of a protective Cr-rich oxide. This effect was most pronounced at the lowest temperature of 550 °C, which was attributed to strong preferential adsorption of sulfur-containing species within the oxide at relatively low temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Technological aspects in blanket design: Effects of micro-alloying and thermo-mechanical treatments of EUROFER97 type steels after neutron irradiation

Presently available data on neutron irradiation damage raise doubts on the feasibility of using EUROFER97 steel for a water-cooled starter blanket in a DEMO reactor, since the ductile-to-brittle transition temperature (DBTT) increases significantly for irradiation temperatures below 350°C. The additional DBTT shift caused by H and He transmutation can only be estimated based on very few results with isotopically tailored EUROFER97 steel. Conservative calculations show that the DBTT of EUROFER97 steel could exceed the operating temperature in water-cooled starter blankets within a relatively short time period. This paper presents results from a EUROfusion funded irradiation campaign that was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The paper compares ten newly developed reduced activation ferritic-martensitic (RAFM) steels irradiated to a nominal dose of 2.5 dpa at 300°C. The post-irradiation experiments using Small Specimen Test Technology included hardness, tensile, and fracture mechanics tests combined with fractography and microstructure analysis are presented. Results show that micro-alloying EUROFER97-type steels influenced the mechanical properties but a dominating impact on irradiation damage resistance could not be identified. In contrast, specific thermo-mechanical treatments lead to better DBTT behavior. Discussion about irradiation response to heat treatment conditions is also given. Here, despite requiring data also at high dpa values, the results indicate that with these modified materials an increased lifetime and potentially also an increased operating temperature window can be achieved compared to EUROFER97.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Deuterium permeation and retention in 316L Stainless Steel Manufactured by Laser Powder Bed Fusion

Deployment of additively manufactured materials into nuclear energy systems requires investigation of the full range of the unique environmental effects on these materials. Hydrogen isotopes are common gaseous species in nuclear reactors and exhibit ample high mobility in most materials. While hydrogen isotopes mobility in wrought stainless steel is well understood, this property is not thoroughly studied for its additively manufactured variants. In this study, we investigated the deuterium permeation and retention in 316L stainless steel manufactured by laser powder bed fusion. The results showed that the deuterium permeability in the as-built additively manufactured 316L stainless steel (AM SS316L) is greater than that of the reference wrought 316L stainless steel by a factor of 2.8. However, the stress-relieved and solution-annealed AM SS316L samples exhibit lower deuterium permeability in comparison with the as-built condition. Following the solution annealing, the deuterium permeability of AM SS316L is comparable with that of the reference wrought SS316L. Deuterium retention in the as-built AM SS316L is 45% higher than that of the wrought 316L stainless steel and slightly higher than that of the two thermally annealed AM SS316L materials. Finally, transmission electron microscopy and positron annihilation lifetime spectroscopy were used to obtain microstructural information used to determine deuterium permeation and retention behavior in the studied materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

The microstructure effects on irradiation response of ferritic – martensitic steels

Microstructural optimization to achieve greater mechanical strength has been one of the focuses in ferritic–martensitic steels development. However, these optimized microstructures’ effects on the radiation response are not well known. In this work, two ferritic–martensitic steels (9Cr-NbMo and 9Cr-Ta) underwent neutron irradiation in the High Flux Isotope Reactor, and their room-temperature post-irradiation tensile properties and microstructure evolutions were investigated and compared. These two steels exhibit similar pre-irradiation tensile behavior, and their yield strengths are higher than that of other ferritic–martensitic steels by about 200–250 MPa. Microstructural characterization on pre-irradiated materials reveals a smaller grain size in 9Cr-Ta (2.8 ± 0.3 μm in 9Cr-Ta versus 4.3 ± 0.5 μm in 9Cr-NbMo) but higher dislocation density and precipitate density in 9Cr-NbMo. As is common for ferritic–martensitic steels at low irradiation temperatures (less than about 0.45T m ), irradiation-induced hardening at 400 °C was observed for both alloys. Irradiation at 490 °C causes the two alloys to exhibit different tensile behavior: 9Cr-Ta softens by 208 MPa in yield stress, whereas 9Cr-NbMo maintains strength. Additionally, microstructural characterizations were performed, including precipitate growth, dislocation, and defect formation. Using the barrier hardening model for microstructure–property correlation, the softening in irradiated 9Cr-Ta is primarily attributed to the significant dislocation recovery, while the strength lost from the slight dislocation recovery in 9Cr-NbMo was compensated by the additional strength from the irradiation-induced cavities. The microstructure effect (primarily precipitate, dislocation and boundary) on the radiation response is discussed herein.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Tribocorrosion of stainless steel sliding against graphite in FLiNaK molten salt

The molten salt reactor (MSR) concept recently gained renewed interest in developing Generation IV nuclear reactors. One MSR design uses fluoride molten salts to cool tri-structural isotropic particle fuel encapsulated by a carbon matrix into spherical pebbles, which would inevitably contact the reactor's stainless steel container wall during salt circulation. Characterizing this interaction is crucial for reactor safety. Here, in this work, we report the tribocorrosion behavior of graphite sliding against Type 316H stainless steel lubricated by a FLiNaK (LiF:NaF:KF; 46.5:11.5:42 mol %) molten salt in an argon environment. Accelerated wear loss was observed at a higher temperature because of a lower molten salt viscosity and a higher corrosion rate. The graphite had a more rapid material loss at a higher sliding speed than stainless steel because of its higher vulnerability to vibration-induced microfracture. The salt-starved condition caused more material loss than either the no-salt or the salt-flooded condition because neither a graphite transfer film nor stable boundary lubrication could be established at salt starvation. An interesting dual-layer surface film was discovered on the stainless steel worn surface: a chromium-rich top film and a nickel-accumulated but chromium-depleted interlayer. The film composition and structure provide fundamental insights to the mechanochemical interactions between stainless steel and graphite in a molten salt environment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Tritium Aging Effects on Fracture Toughness of Stainless Steel Weldments

The long-term embrittlement effects of tritium and decay helium on the structural properties of stainless steels have been studied for years at Savannah River to provide required data for establishing safe operating conditions and lifetimes of pressure vessels used to contain tritium gas. In this study, the fracture toughness properties of the longest-aged tritium-precharged stainless-steel base metals and weldments tested at Savannah River were measured and compared to earlier results. The fracture toughness values were the lowest recorded here for tritium-exposed stainless steel. As-forged and as-welded specimens were thermally precharging with tritium gas at 34.5 MPa and 623 K, then aged for up to 17 years to buildin decay helium prior to testing. ASTM J-Integral fracture mechanics analyses, transmission electron microscopy (TEM), and small angle neutron scattering (SANS) examinations were conducted to characterize the effects of tritium and its radioactive decay product, helium-3. Results show that the fracture toughness values were reduced to less than 2-4% of the as-forged values for specimens with more than 1300 appm helium from tritium decay. The trend of decreasing fracture toughness values with increasing helium content was consistent with earlier observations, and the data show that Type 304L stainless steel is more resistant to tritium-induced cracking than Type 21-6-9 stainless at similar decay helium levels. The fracture toughness properties of long-aged weldments were also affected, but the reductions were not as severe over time because weldments did not retain as much tritium as did the base metals. TEM observations were used to characterize the effects of decay helium bubbles on the deformation substructures, but nanometer-sized helium bubbles were not easily resolved because of high dislocation densities within the forged microstructures. SANS results are presented that suggest the technique can provide information on decay helium bubble size, spacing, and distribution in these steels.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Impurities on the Compatibility of Steels in Supercritical CO 2 at 450°–650°C

Direct-fired supercritical CO 2 (sCO 2 ) power cycles are a pathway to low-CO 2 fossil energy but contain O 2 and H 2 O in the sCO 2 from combustion. The effect of impurities on structural steels was investigated at 450°-650 °C in 30 MPa sCO 2 . The test matrix included 9 and 12%Cr ferritic-martensitic (FM) steels and conventional and advanced austenitic steels exposed for 1000-2000 h with and without additions of 1%O 2 and 0.1%H 2 O to simulate the cycle after water removal. For FM steels, the mass gains and scale thicknesses were similar with and without impurities with the formation of thick, duplex Fe-rich scales in all cases including the observation that Fe 2 O 3 only formed with 1%O 2 . For the austenitic steels, higher mass gains were observed at all temperatures with increased formation of Fe-rich oxides when impurities were added. Carbon ingress was assessed by bulk combustion analysis, glow discharge optical emission spectroscopy (GDOES) and measuring postexposure room temperature tensile properties. In conclusion, bulk C content was strongly increased at 650 °C but not at 450° or 550 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Outgassing measurements of bare and magnetite-coated low-carbon steel vacuum chambers

The outgassing properties of bare and magnetite-coated AISI 1020 low-carbon steel vacuum chambers were evaluated to establish material selection criteria for extreme high vacuum applications, namely, to explore the possibility of using these materials to build next-generation spin-polarized photoelectron guns. Water outgassing measurements using the throughput method revealed that the magnetite-coated chamber exhibited five times lower outgassing at room temperature prior to baking, but this advantage disappears after 80 °C baking. Hydrogen outgassing measurements demonstrated significant differences after intensive heat treatment: the bare low-carbon steel vacuum chamber achieved a specific outgassing rate of 9.6 × 10 −16 Torr L s −1 cm −2 after 400 °C/50 h bake plus additional heat treatment at lower temperatures, 25 times lower than the magnetite-coated low-carbon steel chamber. Residual gas analysis showed >99% hydrogen composition after heat treatment for both materials, with carbon species below detection limits for bare low-carbon steel versus 0.8% for magnetite-coated surfaces. These measurements indicate that properly conditioned bare low-carbon steel can achieve the <10 −12 Torr pressures required for next-generation spin-polarized photogun applications. In conclusion, the paper includes various analysis techniques intended to explain observed behaviors: isotherm analysis of pump-down plots, Arrhenius analysis of hydrogen outgassing rate data, and residual gas composition tracking.

Adsorption isotherm↗
DOE OSTI2022

Stainless and carbon steel corrosion in aqueous piperazine at absorber and water wash conditions

Corrosion of carbon steel (C1010) and stainless steels (304, 316L, 430) was measured at absorber and water wash conditions at the bench scale. Corrosion rate decreases with increasing concentration of piperazine (PZ). With more than 0.003 m PZ in solution, carbon steel has acceptable corrosion performance. Corrosion of carbon steel increases with increasing partial pressure of CO2, suggesting loading is another dominant parameter. The effect is more significant for dilute PZ; therefore, untreated flue gas with high CO2 content should be avoided in the water wash. Temperature has a less significant effect than PZ concentration and loading. Carbon steel corrosion increases with increasing flow velocity at both absorber and water wash conditions

Liu, Ching-Ting↗
DOE OSTI2020

Evaluation of the Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC N690-18, for Application to Nuclear Power Plants

This report describes the assessment of the recently published ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities. This work was performed for the U.S. Nuclear Regulatory Commission for potential endorsement of N690-18 for use in the design of nuclear power plants. Currently the NRC endorses the use of an earlier version: N690-1994 (R2004), and thus, a technical review of the updated N690 Specification, which contains significant changes, is warranted. The N690-18 Specification reflects advances and improvements in steel design gained from experience, analytical studies, and experimental data over the years. The N690-18 specification also contains a new Appendix N9 that addresses the design of safety-related steel-plate composite (SC) walls which are a relatively new type of structural member for which previously there has not been any U.S. code or standard that governs its design. The technical review of N690-18 was performed in order to assess the adequacy of the provisions for use by the NRC in updating its regulatory guidance for nuclear power plants. The research developed the technical basis for acceptance of the new specification and also identified areas where additional staff guidance is needed for the design of safety-related steel structures at nuclear power plants. The results of the research have determined that the N690-18 specification is a significant enhancement from the N690-1994 (R2004) edition. The new specification contains substantial upgrades in the design of structural steel and includes new provisions for design of SC wall sections, which can benefit the nuclear power industry as well as the NRC review process. All of the improvements, if endorsed by the NRC with certain regulatory positions, will allow clearer and more consistent regulatory guidance, and will avoid the need for justification by licensees and related reviews each time certain methods or approaches are used.

42 ENGINEERING↗
DOE OSTI2021

Fatigue Performance of High-Strength Pipeline Steels and Their Welds in Hydrogen Gas Service

Objectives of the project include: Enable the use of high strength steel hydrogen pipelines, as significant cost savings can result by implementing high strength steels as compared to lower strength pipes. Demonstrate that girth welds in high-strength steel pipe exhibit fatigue performance similar to lower-strength steels in high-pressure hydrogen gas. Identify pathways for developing high-strength pipeline steels by establishing the relationship between microstructure constituents and hydrogen-accelerated fatigue crack growth (HA-FCG)

08 HYDROGEN↗
DOE OSTI2021

Progress Report on Performance of A709 and G91 Steels in Sodium

Specimens in six different processing and heat treatment conditions of A709 H58776 were exposed to sodium at 550, 600, and 650°C, respectively for various exposure times. G91 base metal was tested in sodium at 550, 600, and 650°C, and G91 weldment at 550 and 600°C, respectively. Parallel thermal aging experiments were conducted on G91 and A709 to obtain thermal aging data for comparison with sodium exposure results to separate the thermal and sodium effects. The corrosion data obtained on A709 steel continue to show low corrosion rates, which indicate good compatibility of A709 with sodium when oxygen content is controlled. G91 has also shown acceptable corrosion rates over the temperature range of 550-650°C in sodium. Thermal aging or sodium exposures reduced the tensile strength, uniform elongation, and total elongation for A709 ESR, AOD specimens. These specimens consistently showed higher yield stress and tensile strength and lower uniform and total elongations after sodium exposure than after thermal aging under comparable testing conditions. It implies that there was an additional effect of sodium exposure that gave rise to an increase in tensile strength of A709. Dynamic strain aginginduced flow serrations were completely removed by either thermal aging or sodium exposure at 650°C, which imply that carbon or nitrogen in solution was removed from the solution and formed precipitates during aging or sodium exposures. The tensile data ruled out the possibility of decarburization under the sodium exposure conditions. The HOMO specimens behaved somewhat differently from the ESR and AOD specimens in sodium. While thermal aging and sodium exposures at 550 and 600°C decreased the yield stress and the ultimate tensile strength of G91, there was virtually no additional effect resulting from sodium exposures at these two temperatures. In contrast, sodium exposures at 650°C had a drastic effect on the yield stress and the ultimate tensile strength of G91. It is suggested that G91 experienced decarburization in the 650°C sodium environments. Carbon concentrations in sodium in the SMT-1 and SMT-2 loops were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8-1.2 ppm in the SMT-1 loop and 0.3-0.7 ppm in the SMT-2 loop. The carbon activity in sodium in the SMT-2 loop was estimated to be 0.03-0.08 at 600°C, and 0.08-0.2 at 550°C. The carbon activity in sodium at 650°C in the SMT-1 loop was 0.04-0.07. Equilibrium simulation of the carburization – decarburization processes was conducted for A709 and G91 steels exposed in sodium environments at temperatures of 550-700°C. The carbon activity-concentration relationship for G91 was re-evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC carbides dominated the process at low carbon activities. Formation of NbC increases the decarburization resistance of G91 steel. It remains to be understood whether the decarburization resistance provided by MC carbides in G91 can be maintained during long-term operations of SFRs. The carbon concentration-activity relationship for A709 was calculated based on the equilibrium of fcc-austenite and M 23 C 6 carbide phase. The calculations showed that A709 decarburizes at 650°C in the SMT-1 loop environments, which is different from the experimental findings. The carbon concentration-activity relationship in A709 was further evaluated by considering four phases, including fcc-austenite, M 23 C 6 , NbC, and TiC carbide phases. TiC is the most stable carbides among the three carbide phases. The carburization-decarburization process in A709 is dictated by M 23 C 6 carbides at high carbon activities and by TiC at low carbon activities. Formation of TiC increases the decarburization resistance of A709 steel. Because of the complexity of the precipitation process in A709 during sodium exposures, detailed characterization of precipitates in sodium-exposed specimens and improved thermodynamic models are needed to understand the carburization-decarburization behavior of A709. Kinetic analysis of carbon transfer will be investigated for G91 and A709 in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

STEM imaging of irradiation induced defects in Eurofer97 steel variants irradiated in the EUROfusion collaboration

Ten exploratory variants of Eurofer97 reduced activation ferritic martensitic (RAFM) steel were irradiated in HFIR to ~2.94 – 32.4 dpa, 300±30 °C as a part of the EUROfusion collaboration. The irradiations were performed in rabbit capsules. ES21-22 capsule were used for irradiating SS-J3 tensile samples while M4CVN bend bar samples were irradiated in ES31-35 capsules. The mechanical properties and multi-length scale characterization of the microstructures of these ten steels, designated as H, I, P, J, K, L, M, N, O, P and reference E, in the nonirradiated form was reported in "Mechanical properties and microstructure characterization of unirradiated Eurofer-97 steel variants for the EUROfusion project, ORNL/SPR-2018/882," while their irradiated properties and some microstructure analysis was reported in "Post-irradiation examination of Eurofer97 steel variants irradiated to 2.5 dpa, ~300 °C in HFIR for the EUROfusion program, ORNL/SPR-2020/1440." Here, additional microstructures using TEM/STEM and STEM-EDX of the irradiated steels is presented. Characterization was performed at the Low Activation Materials Development and Analysis (LAMDA) laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2022

High Strength Steel-Aluminum Components by Vaporizing Foil Actuator Welding

This project aimed to address the challenge of effectively welding dissimilar materials—high-strength steel and high-strength aluminum for creating lightweight, multi-material automotive components. For automotive companies, reducing weight of a vehicle is critical task regulated by the government to solve the issue of greenhouse gas emissions. Production of lightweight cars and trucks can be achieved by substitution of current all-steel structures with multi-material lightweight structures that include high strength-to-weight-ration materials such as high-strength steels, aluminum alloys, magnesium alloys, titanium alloys, coupled with lightweight designs. This requires dissimilar metal welding, which is challenging for state of the art joining processes such as resistance spot welding. The cycle of melting-cooling-freezing during traditional welding that can easily ruin the designed outstanding properties of the advanced base metals, such as aluminum alloys, making the welded area much weaker than the base metals. To weld two different metals with great difference in melting points, such as aluminum and steel, it’s even more difficult or impossible because of the formation of brittle intermetallic compounds at the welded interface. In this project, a novel welding method, developed at OSU, was selected for validation and development. This novel technology enables welding by impact without melting and proves to be robust to join various dissimilar lightweight metals. Termed as vaporizing foil actuator welding or VFAW, the technology uses a thin aluminum foil that is rapidly vaporized by a high current pulse to produce an explosive-like pressure pulse to drive one metallic piece into another at the high speed required for impact welding. This project entailed development of the early-stage welding technology in terms of (a) the consumables, the welding apparatus and the power sources, (b) coupon scale screening of many material combinations including corrosion studies, (c) computational modeling and design of the welded interface as well as of the multi-material prototype component, and (d) mechanical testing for strength and durability at coupon scale and to a certain extent the prototype scale. The all-steel engine cradle of 2016 Chevrolet Cruze was chosen as the baseline prototype component. The target set for the project was to demonstrate a 20% weight reduction at a cost premium of less than $\$ $5/lb saved without compromising on baseline mechanical properties. At project completion, a 12% lighter prototype component was demonstrated with an estimated cost premium of $\$ $9.8/lb saved. Besides prototype level demonstration of the technology, this project also enabled elevation of the technology’s readiness level to where a hydraulically actuated welding head was developed and made ready for deployment at a research and development facility for Tier 1 automotive supplier.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effects of Surface Treatments on the Outgassing of 3D-Printed 316L Stainless Steel.

Fermi National Accelerator Laboratory (Fermilab) is a world leader in designing and operating particle accelerators for nuclear and particle physics research. These accelerators rely on electric fields to propel particle beams, which are then directed and focused using magnetic fields. To ensure these beams travel without interference, they must move through an ultra-high vacuum environment, typically around 10⁻⁹ torr. Maintaining such a vacuum requires materials with exceptionally low outgassing rates. 316L stainless steel is a common material choice for vacuum systems due to its low outgassing rates, excellent corrosion resistance, non-magnetic properties, and high mechanical strength. When manufactured using traditional methods like machining, extrusion, and casting, the vacuum properties of 316L stainless steel are well understood. However, the behavior of 316L stainless steel produced through additive manufacturing, such as Direct Metal Laser Sintering (DMLS), is not as well documented. This study aims to determine the outgassing rate of 316L stainless steel fabricated using DMLS. The investigation focuses on measuring the outgassing rates of 1.5 2 vacuum reducers and evaluating the effects of surface preparation techniques, such as mechanical polishing and electropolishing, on the outgassing performance. Understanding the outgassing performance of 3D-printed 316L stainless steel could enable Fermilab engineers to adopt this fabrication method for producing complex parts designed for ultra-high vacuum environments. These advancements have the potential to benefit projects, such as the Long-Baseline Neutrino Facility (LBNF) and the Proton Improvement Plan-II (PIP-II), by improving design flexibility and fostering innovation in accelerator development.

Cloud, Jaiden [Central State University]↗
NASA NTRS1974

Stress corrosion cracking susceptibility of 18 Ni maraging steel

The stress corrosion cracking (SCC) resistance of 18Ni maraging steel (grades 200, 250, 300, and 350) was determined in 3.5 percent salt (NaCl) solution, synthetic sea water, high humidity, and outside MSFC atmosphere. All grades of the maraging steel were found to be susceptible to SCC in varying degrees according to their strengths, with the lowest strength steel (grade 200) being the least susceptible and the highest strength steel (grade 350), the most susceptible to SCC. The SCC resistance of 250 grade maraging steel was also evaluated in salt and salt-chromate solutions using fracture mechanics techniques. The threshold value, K sub SCC, was found to be approximately 44 MN/sq m square root m, (40 ksi square root in.) or 40 percent of the K sub Q value.

Humphries, T. S.↗