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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Structural Studies of Alloyed and Nanoparticle Transition Metal Dichalcogenides by Selenium-77 Solid-State Nuclear Magnetic Resonance Spectroscopy

Layered transition metal dichalcogenides (TMDCs) such as MoS 2 , MoSe 2 , and WSe 2 are under intense investigation because they are atomically thin semiconductors with photophysical properties that can be tuned by changing their composition or morphology. Mechanochemical processing has been proposed as a method to obtain alloyed TMDCs in the series Mo 1–x W x S y Se 2–y (x = 0–1; y = 0–2). However, elucidating the chemical transformations occurring at the atomic scale following mechanochemical processing can be challenging because the products are often amorphous or microcrystalline. To address this challenge, we probe TMDC mixing and alloying by using a combination of powder X-ray diffraction, Raman spectroscopy, diffuse reflectance spectroscopy, 77 Se solid-state nuclear magnetic resonance (SSNMR) spectroscopy, and planewave density functional theory (DFT) calculations. The nature of the milling material and reaction atmosphere are shown to be essential factors in limiting the formation of undesired oxide byproducts. We demonstrate acquisition of 77 Se SSNMR spectra using different combinations of Carr-Purcell Meiboom-Gill acquisition (CPMG) pulse sequences, magic angle spinning (MAS), and MAS dynamic nuclear polarization. Further, the combination of SSNMR with the other characterization methods clearly demonstrates that high energy impact ball milling induces molecular level alloying of Mo, W and chalcogen atoms in the family Mo 1–x W x S y Se 2–y . Gauge including projector augmented wave DFT calculations yield accurate 77 Se chemical shift (CS) tensor components. 77 Se SSNMR spectroscopy was also applied to study the structure of WSe 2 nanocrystals intercalated with ethylenediamine. The intercalated WSe 2 nanocrystals exhibit a more positive isotropic 77 Se CS as compared to bulk WSe 2 , however, the 77 Se CS anisotropy is the same, confirming the WSe 2 layers have a similar structure as in their bulk counterparts.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Assessment of UTP A 2133 Mn as a Matching Filler Metal for Alloy 800H in Section III, Division 5 Applications

Six alloys are qualified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear components. Due to reactor design alloy restrictions and motivation to improve operational efficiency and plant lifetimes, it is necessary to maximize the usefulness of qualified alloys. One of these alloys is Alloy 800H, which is qualified for a maximum temperature of 760°C and a maximum 300,000-hour service life. However, the welding filler metals qualified to join Alloy 800H have stress rupture factors that reduce the allowable weldment strength by factors as low as 0.59 times the Alloy 800H base metal. In an effort to improve the weldment creep-rupture performance, or to increase the stress reduction factor, non-code qualified filler metals are under investigation by the Department of Energy?s Advanced Reactor Technologies program. For this investigation, UTP A 2133 Mn filler metal (Fe-Cr-Ni-Mn-Nb) was used to join 0.5 in. thick Alloy 800H plate to demonstrate passing an ASME BPVC Section IX weld qualification. The multi-pass, pulsed GTAW process showed successful weld qualification results, and additional property measurements were conducted to compare previous analyses of Alloy 82 and Alloy 617 filler metals. Future work will evaluate the cross-weld creep-rupture performance of the UTP A 2133 Mn in comparison to Alloy 82 and Alloy 617 filler metals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Report on the Assessment of Ion Irradiations on High Entropy Alloys

Idaho National Laboratory (INL) initiated a joint international effort with the Czech Republic to explore the feasibility of manufacturing High Entropy Alloys for high-temperature nuclear applications using Additive Manufacturing. This effort was funded at INL by the United States Department of Energy’s Office of Nuclear Energy under the Advanced Reactor Technologies and Advanced Materials and Manufacturing Technologies (AMMT) Program. The High Entropy Alloys were specifically designed for the corrosive and irradiation environments as experienced in gas fast reactors, molten salt reactors, and fusion power. These alloys were manufactured by multiple processes to determine the impact of manufacturing processes on the performance of the alloys in corrosive and irradiation environments. Preliminary results show that MoNbTiV and MoNbTi alloys exhibit no degradation after multiple corrosion tests. These MoNbTiV and MoNbTi alloys were also evaluated after irradiation experiments were conducted at the Michigan Ion Beam Laboratory at the University of Michigan, which is the focus of this milestone report. For these irradiation experiments, the MoNbTiV and MoNbTi alloys manufactured by arc melting and Electric Field Assisted Sintering (EFAS) were tested at 500 C at a dose of 30 dpa to evaluate the microstructural stability and decomposition.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

GCR: Development of Improved Alloy 800H Weldment

In Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Alloy 800H is qualified for elevated-temperature nuclear construction for temperatures up to 760°C (1400°F) and a maximum service life of 300,000 hours. There are two permissible filler metals for Alloy 800H weldments specified in Division 5: ENiCrFe-2 (Alloy A) and ERNiCr-3 (Alloy 82). Low creep-rupture strengths of these weldments at the upper limits of the qualified temperatures and service lives may restrict the design envelope for elevated-temperature nuclear construction with Alloy 800H. As a result, an alternative filler metal is desired to improve the creep-rupture strengths of Alloy 800H weldments for the qualified temperatures and service lives. This work investigates an overmatched filler metal. Specifically, a weldment with Alloy 800H base metal and Alloy 617 filler metal fabricated by semiautomated gas tungsten arc welding is investigated. A scoping creep-rupture test program was conducted of cross-weld specimens at temperatures ranging from 750 to 1000°C (1292 to 1832°F). Preliminary results on the creep-rupture strengths of the Alloy 800H weldment with an Alloy 617 filler metal do not show significant improvement compared to the filler metals currently qualified in Division 5 for Alloy 800H weldments. Consequently, work is in progress to investigate a matching filler metal, UTP A 2133 Mn.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural Characterization of Ion Irradiated ODS MA956 Alloy

Oxide Dispersion Strengthened (ODS) alloys are considered as promising nuclear materials for the next generation advanced nuclear reactors due to their improved elevated temperature mechanical properties and radiation damage resistance. MA956 is a commercial ODS alloy with superior oxidation resistance and creep strength at elevated temperatures. The improved properties for ODS alloys can be attributed to the presence of dispersed nanosized oxide particles. These oxide particles can act as barriers to dislocation motion and provide additional sinks and sites for point defect recombination. Hence, their stability can greatly affect the properties of the irradiated ODS alloys. Besides, irradiation will induce defects into the ODS alloys, and these defects can develop further as irradiation dose increases. Furthermore, the defects propagation can significantly undermine the mechanical properties of the bulk materials, resulting in irradiation hardening and lower ductility. Therefore, it is necessary to have a good understanding of the defect’s evolution with irradiation. Hence, efforts are ongoing to understand the effect of ion irradiation on ODS MA956 with emphasis on oxide particles and evolution of defects as a function of dose (2.5, 50 and 100 dpa).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Isothermal oxidation of zirconium using steam

Oxidation of zirconium and its alloys under any atmosphere is generally characterised by a protective and non-protective stage. During the protective stage, the kinetics of oxidation are quasi-parabolic, while the oxide scale formed on the surface of the metal is black, compact, and protective in nature. For the non-protective stage, the oxidation rate becomes linear, while the protective oxide film turns into a deteriorated white scale. Breakaway is known as the transition point from protective to non-protective stage. The actual mechanism leading to breakaway of zirconium is still a matter of controversy, however some of the most common factors affecting breakaway are the following: - Impurities on the metal (Carbon, Oxygen, Nitrogen); - Oxidizing atmosphere (Presence of H{sub 2}O (g) or N{sub 2} (g)); - Sample preparation (Chemically etched or polished); - Temperature (LWR, LOCA). This research investigates the effect of temperature on 'breakaway' when zirconium is oxidized isothermally in steam for 60 min at 800 and 900 deg. C using thermogravimetric and optical microscopic imaging techniques. 1. Steam isothermal oxidations were performed on zirconium at 800 and 900 deg. C using a Thermogravimetric Analyzer (TGA). 2. Log-log plots of mass gain vs time were obtained from thermogravimetric results to calculate the reaction orders (n) and reaction rate constants (K) during oxidation. 3. Results of the kinetics of reaction were correlated with optical microscopic images form Light Microscope (LM) and Scanning Electron Microscope (SEM). Breakaway depends very strongly on temperature. At 800 deg. C, the weight gain followed a cubic-parabolic behaviour with evidence of a protective oxide film. On the other hand, oxidation at 900 deg. C showed signs of breakaway since the oxide film was cracked and oxidation rate was almost linear throughout the reaction.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

NMR studies of the ground states of Ni 50-x Co x Mn 35 In 15 (x=1, 2.5) and Ni 45 Co 5 Mn 37 In 13 Heusler alloys

Three temperature-induced phase transitions at T=T 1 , T M /T A , and T C , related to the ferromagnetic order of the martensitic phase (FMMP), martensitic (structural) transitions (MT), and the ferromagnetic order of the austenitic phase (FMAP), respectively, have been observed in the off-stoichiometric Heusler alloys, Ni 50-x Co x Mn 35 In 15 (x=1, 2.5) and Ni 45 Co 5 Mn 37 In 13 . In this work, the phase transitions temperatures are found to be depended on alloy composition. A kinetic arrest of the AP was observed for Ni 47.5 Co 2.5 Mn 35 In 15 in the magnetization measurements during field-cooling cycle (FCC) at 50 kOe. Depending upon the cooling protocols, ZFC and FCC (at H = 50 kOe), two different ground states of the alloys can be found in Ni 47.5 Co 2.5 Mn 35 In 15 and Ni 45 Co 5 Mn 37 In 13 alloys. The ground states (T=4.2 K and external field H=0) of the alloys was found to be characterized by three main line: two, partially overlapping, at higher frequencies (300-450 MHz), most likely corresponding of manganese resonance lines and one at lower frequency at about 200 MHz. A significant shift in the spectrum of Ni 45 Co 5 Mn 37 In 13 by about 100 MHz to higher frequencies was observed. The correlation of magnetizations obtained from magnetic moment and NMR studies is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Creep and Creep Fracture Modeling with Surrogate Creep Models and the Extended Finite Element Method

Alloy components in advanced nuclear reactors will be subjected to environmental conditions that could include high temperatures, irradiation, and exposure to corrosive salts. These conditions could lead to the formation of crack-like defects, which could grow over time in a mechanism known as creep crack growth (CCG). Predicting growth rates of these defects is important for assessing the safe operating life of advanced reactors. This project documents progress toward developing and testing next-generation data-driven constitutive models for deformation creep. It also documents the application of the extended finite element method in conjunction with surrogate creep models to predict CCG parameters under a variety of conditions. These important incremental developments contribute to the longer-term objective of developing microstructure-aware constitutive models that can be used for predicting creep deformation and CCG at the component scale with improved accuracy.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Formation of multicomponent alloy particles in doped ceria under I 2+ ion irradiation and thermal annealing

Multicomponent alloys resembling the e-phase particles in nuclear fuel have the potential to immobilize 99 Tc in a stable waste form. Formation of such multicomponent alloys under extreme conditions is studied in a non-activated model system: CeO 2 doped with 2 wt% Mo, 1.5 wt% Ru, 0.75 wt% Pd, 0.5 wt% Re and 0.25 wt% Rh. The doped CeO 2 films were irradiated with I 2+ ions (610°C, 10 16 and 5×10 16 I 2+ /cm 2 ). For selected samples post-irradiation heat treatment was conducted (900°C, 1100°C). Analytical TEM revealed the formation of 5-20 nm precipitates containing Pd, Mo, and Re sometimes associated with cavities. Iodine was found to segregate to grain boundaries and cracks. After heat treatment at 1100°C, the CeO 2 matrix had recrystallized. Precipitates containing Pd, Mo, Re, and Ru were observed near the interface with the polycrystalline YSZ substrate. It follows that Pd is the most mobile element, followed by Mo and Re. Finally, while irradiation promotes precipitation, high-temperature effects are more significant.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effect of dose rate on the phase stability of a CrFeNiMn alloy

We report CrFeNi-based multi-principal element alloys (MPEAs) are being considered for nuclear applications because of their potentially superior mechanical properties and irradiation resistance. While the alloys appear stable against phase decomposition during ion irradiation performed at high dose rate, previous electron irradiation and thermal annealing studies suggested that some of them may be susceptible to phase decomposition. Therefore, to clarify the role of dose rate during ion irradiation on phase decomposition, a CrFeNiMn alloy was irradiated using Fe ions at 500 °C with an intermediate dose rate of 10 -4 dpa/s. Transmission electron microscopy and atom probe tomography observations revealed a high number density of NiMn precipitates, cavities, and dislocation loops, which had not previously been observed after ion irradiation at higher dose rates.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Formation and Consolidation of UN Nanostructures in U-Mo Fuel via Mechanical Alloying and Spark Plasma Sintering

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and high fuel loading capability. However, U-Mo systems suffer from irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interaction. To address these shortcomings, this thesis investigated the formation of uranium mononitride (UN) nanoparticles within a U-10Mo matrix. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen cover gas (99.9995%) atmosphere. Variations were made in stainless steel milling media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. UN nanoparticulates were successfully formed within the U-Mo matrix and was characterized utilizing light element analysis, X-ray diffractometry, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. Presence of the UN nanoparticles was found as early as 1-hour into the mechanical alloying process, and significant iron contamination found after 10-hours of milling. The U-Mo/UN powders were compacted using spark plasma sintering techniques in order to stabilize the nanostructure. U-Mo/UN compacts containing small amounts of neodymium and helium were analyzed to observe the defect migration tendencies of fission products within the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

The chlorination and separation of aluminum using low-temperature sulfur chloride reagents

There is currently no strategy for the permanent waste disposal or recycling for used nuclear fuelsfrom research reactors. For this reason, low-temperature reactions have been developed for thechlorination of the Al alloys and subsequent separation from used nuclear fuels to reducethe volume of high-level waste in storage. Three sulfur chloride reagents – S 2 Cl 2 , SOCl 2 , and SO 2 Cl 2 —were tested, and two were found to quantitatively chlorinate Al metal and Al alloys undermild conditions. Further, these low-temperature reactions proceed between 298 and 411 K, and up to 5 gof metal is chlorinated in 1–3 h. Preliminary results indicate that the reactivity and exothermicity ofthe reaction between the Al and sulfur chloride reagents is highly dependent on the surface area-to-volume ratio of the metal and the volume of solvent. Elemental S is produced as a by-productduring the chlorination with S 2 Cl 2 but can be quantitatively rechlorinated under mild conditions toregenerate the initial chlorination reagent. Therefore, in this case, chlorine is the only elementconsumed in the reaction, thus minimizing the waste generated during the chlorination process.The AlCl 3 may then be separated from other materials present in Al 6061 or Al 8001 because of itshigh solubility in the sulfur chloride reagents. This process may also be extended to chlorinate Alfrom research reactor fuels.

aluminum↗