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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Welded and Hydrogen Charged Zircaloy-4 and Welded Stainless Steel 347 Property Data and Microstructures for the Target Solution Vessel and Support Lines of SHINE

SHINE Medical Technologies is developing a facility that will produce molybdenum-99 for medical isotope procedures. As part of this facility, a metal tank, called the target solution vessel (TSV), will hold a uranyl sulfite solution. The TSV must withstand neutron irradiation near room temperature and potential corrosion from the aqueous solution. The large cylindrical TSV will be fabricated by welding and will have numerous pipes and connections also attached via welding. The vessel and those pipes will have varying thicknesses so both gas tungsten arc welding (GTAW) and flux-cored arc welding (FCAW) will be necessary to fabricate different sections. The original material under consideration for the TSV was Zircaloy-4 (Zry-4), but it is known to form a hydride which can degrade its mechanical properties. Thus, some investigation here is focused on the effects of hydrogen uptake in the Zry-4. Additionally, the alternative material being considered is AISI 347, a stainless steel. Both materials have little existing data for their neutron irradiation behavior below 100°C, and both have open questions on the weld behavior under neutron irradiation. Testing has focused on characterizing their weld properties with tensile tests and performing neutron irradiation of samples in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Second Annual Progress Report on Correlation Between Microstructure and Mechanical Properties of Neutron-Irradiated Ferritic-Martensitic and Austenitic Steels

Ferritic-martensitic steels G92-2b (an optimized Grade 92 heat), NF616 and T91, and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) and the Advanced Test Reactor (ATR) of Idaho National Laboratory (INL). Selected G92-2b samples were irradiated up to 14.66 dpa in the HFIR at two temperature ranges: 400–496.7°C and 683.3– 720°C. NF616 and T91 were irradiated in the ATR up to 8.16 dpa with the irradiation temperatures ranged from 241°C to 447.5°C. Alloy 800H and 800H-TMP samples were irradiated in both the HFIR and the ATR. Selected 800H and 800H-TMP samples had HFIR irradiation to 1.28 dpa at 580°C and ATR irradiation up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurements, fractography, and microstructural characterization were performed on the selected samples in the Low Activation Materials Design and Analysis (LAMDA) laboratory. Radiation-hardening of G92-2b was observed at the lower doses and lower irradiation temperatures (400- 496.7°C), with GB03 (0.52 dpa at 400°C) and GB04 (7.44 dpa at ~490°C) showing ~12% and ~8% hardening, respectively. Softening by ~14% was observed for GB05 (14.66 dpa at 496.7°C). Radiationsoftening of G92-2b was more prevalent at the higher irradiation temperatures (683.3-~720°C), with GB10 (0.46 dpa at 683.3°C), GB11 (7.44 dpa at ~720°C), and GB12 (14.63 dpa at ~720°C) showing ~8%, ~8%, and ~40% softening, respectively. Radiation-hardening of NF616 and T91 was observed with the hardness increased by ~37% to ~65% depending on the irradiation doses and irradiation temperatures. Within the studied irradiation conditions of NF616 and T91, samples with a higher dose had a larger hardness after irradiation. All the tested alloy 800H and 800H-TMP samples in this work showed radiation-hardening by ~96±7% to ~152±10%. Alloy 800H-TMP tended to have slightly smaller radiation-hardening than alloy 800H. The fractography results of G92-2b sample GB03, GB10, and GB11, together with the previously characterized fractography of GB04, GB05, and GB12, indicated that the ductility of G92-2b was maintained up to 14.66 dpa at the lower irradiation temperatures of 400-496.7°C, while some loss of ductility (less necking) was observed for higher doses at the higher irradiation temperatures of 683.3-720°C. This agrees with the previously reported tensile test results of G92-2b, where the elongation of G92-2b was reduced at higher doses at the higher irradiation temperatures. Dimple sizes increased at higher doses, which are more evident at the higher irradiation temperatures of 683.3-~720°C. The fractography of NF616 sample D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for sample D2, while ductile failure for samples D4 and D6. Fractography of alloy 800H and 800H-TMP samples in various irradiation conditions showed ductile failure with obvious necking. Dimples were observed, with some of them containing large Ti-rich particles, in all the 800H/800H-TMP samples. Electron backscatter diffraction characterization of GB12 indicated the recovery of the lath structure, which was generally replaced by an equiaxed grain structure. Transmission electron microscopy (TEM) characterization showed the presence of frequent M 23 C 6 (M = primarily Cr), MX (M = primarily V), spherical Nb(C,N) precipitates, and occasional Laves phase precipitates in the G92-2b samples. MX precipitates with sizes of 20-30 nm were observed at boundaries of smaller grains, indicating the pinning effect of the V-rich precipitates. The lath structure recovery was more evident at the higher irradiation temperatures (683.3-~720°C), with decreased densities of line dislocations and M23C6 precipitates. The irradiated T91 (TA04) showed the growth of M23C6 precipitates to 101 ± 40 nm from the initial 68 ± 22 nm in the unirradiated condition. Dislocation loops of both {100} and {111} types were present in TA04. TEM characterization was also performed on the irradiated 800H (N4, N5, N6, and AR2) and 800H-TMP (P4, P5, P6, and HG1). Accumulation of M 23 C 6 precipitates at grain boundaries was observed in all the xii 800H/800H-TMP samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in the irradiated 800H-TMP. Some Ti(C,N) precipitates are embedded in the M 23 C 6 precipitates, maintaining specific orientation relationships between the precipitates and between the precipitate and matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all the 800H/800H-TMP samples, with EDS Si maps tending to have a lower contrast in 800H-TMP samples. Atom probe tomography was conducted on the same samples, supported by a Rapid Turnaround Examination project under Nuclear Science User Facilities. The results are being analyzed to be integrated with the TEM results for a confident description of the γ’ precipitates. Dislocation loops also formed in all the 800H/800H-TMP samples. The density and the average size of dislocation loops were quantified to be in the order of 10 22 – 10 23 m -3 and 11.7 – 15.9 nm, respectively, in the ATR-irradiated 800H/800H-TMP samples. The loop densities in irradiated 800H were higher than that in irradiated 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterization of paste microstructure for durability properties of concrete

In this study, paste samples from different cementitious systems were characterized by using X-Ray Micro Computed Tomography (µ-CT) and Fluorescence Microscopy techniques. Quantitative and qualitative aspects of defects from paste samples were investigated to develop indicators of durability related properties of corresponding concrete samples. In the first phase of this study, hour glass shaped 3D printed molds were used to investigate five cementitious combinations consisting of three cement types (Coarse Ground portland cement, Type IL Portland Limestone Cement and Type I/II ordinary portland cement) and fly ash as a supplementary cementitious material. In the second phase, metal molds were used to compare the results of Coarse Ground portland cement with Type I ordinary portland cement, 5% of which was replaced with colloidal nanosilica. Potential correlations between paste and concrete have been presented. This work was supported by US Department of Energy and North Dakota Department of Commerce.

36 MATERIALS SCIENCE↗
NASA NTRS2016

Microstructural and Creep Properties of Boron- and Zirconium-Containing Cobalt-Based Superalloys

The effects of micro-additions of boron and zirconium on grain-boundary (GB) structure and strength inpolycrystalline γ(f.c.c.) plus γ’(L12) strengthened Co-9.5Al-7.5W-X at. % alloys (X=0-Ternary, 0.05B, 0.01B,0.05Zr, and 0.005B-0.05Zr at. %) are studied. Creep tests performed at 850 °C demonstrate that GB strength and cohesion limit the creep resistance and ductility of the ternary B- and Zr-free alloy due to intergranular fracture. Alloys with 0.05B and 0.005B-0.05Zr both exhibit improved creep strength due to enhanced GB cohesion,compared to the baseline ternary Co-9.5Al-7.5W alloy, but alloys containing 0.01B or 0.05Zr additions displayed no benefit. Atom-probe tomography (APT) is utilized to measure GB segregation, where B and Zr are demonstrated to segregate at GBs. A Gibbsian interfacial excess of 5.57 ± 1.04 atoms nm(exp) -2 was found for B at aGB in the 0.01B alloy and 2.88 ± 0.81 and 2.40 ± 0.84 atoms nm2 for B and Zr, respectively, for the 0.005B-0.05Zr alloy. The GBs in the highest B-containing (0.05B) alloy exhibit micrometer-sized boride precipitates with adjacent precipitate denuded-zones (PDZs), whereas secondary precipitation at the GBs is absent in theother four alloys. The 0.05B alloy has the smallest room temperature yield strength, by 6%, which is attributedto the PDZs, but it exhibits the largest increase in creep strength (with an ~2.5 order of magnitude decrease inthe minimum strain rate for a given stress at 850 °C) over the baseline Co-9.5Al-7.5W alloy.

Creep↗
DOE OSTI2023

The hydration, microstructure, and mechanical properties of vaterite calcined clay cement (VC 3 )

Limestone (calcite) calcined clay cement (LC 3 ) is a promising low-CO 2 binder, but the low activity of calcite cannot compensate the reduction in clinker factor, resulting in low one-day strength and limiting its broad applications. As recent carbon capture and utilization technologies allow scalable production of vaterite, a more reactive CaCO 3 polymorph, we overcome the challenge by introducing vaterite calcined clay cement (VC 3 ), inspired by the vaterite-calcite phase change. In the present study, VC 3 exhibits higher compressive strengths and faster hydration than LC 3 . Compared to hydrated LC 3 , hydrated VC 3 exhibits increased amount of hemi- and mono-carboaluminate formation and decreased amount of strätlingite formation. With gypsum adjustment, the 1-day strength of VC 3 is higher than that of pure cement reference. Finally, VC 3 , a low-CO 2 binder, presents great potential as a host of the metastable CaCO 3 for carbon storage and utilization and as an enabler of carbon capture at gigaton scales.

36 MATERIALS SCIENCE↗