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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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116 records · Page 7
All sourcesNASA NTRSDOE OSTI
NASA NTRS1985

The effect of tantalum on the structure/properties of two polycrystalline nickel-base superalloys: B-1900 + Hf MAR-M247

The microstructure, phase compositions, and phase fractions were studied in conventionally cast B-1900 + Hf and both conventionally cast and directionally solidified MAR-M247 as a function of tantalum concentration. The hot tensile and creep rupture properties of the solutionized and aged MAR-M247-type alloys were also determined as a function of tantalum level. The effects of tantalum on the microstructure and phase compositions of B-1900 + Hf and MAR-M247 (conventionally cast and directionally solidified) were found to be very similar. The addition of tantalum to the as cast and heat treated alloys was shown to cause the partial replacement of the Hf in the MC carbides by Ta, although the degree of replacement was decreased by the solutionizing and aging heat treatment. The gamma prime and minor phase fractions (primarily MC type carbides) both increased approximately linearly with tantalum concentration. The gamma prime phase compositions were relatively insensitive to tantalum variations with the exception of the tantalum and/or hafnium levels. Bulk tantalum additions increased the tantalum, chromium, and cobalt levels of the gamma phase in both alloy series. The increase in the concentrations of the latter two elements in the gamma phase was a result of the decrease in the gamma phase fraction with increasing bulk tantalum concentration and constant gamma/gamma prime partitioning ratio. Tantalum additions increased the yield stress and ultimate tensile strength of the directionally solidified MAR-M247 type alloys and had no significant effect on ductility.

Janowski, G. M.↗
DOE OSTI2022

Study of interface reaction in a B 4 C/Cr mirror at elevated temperature using soft X-ray reflectivity

Boron carbide is a prominent material for high-brilliance synchrotron optics as it remains stable up to very high temperatures. The present study shows a significant change taking place at 550°C in the buried interface region formed between the Cr adhesive layer and the native oxide layer present on the silicon substrate. An in situ annealing study is carried out at the Indus-1 Reflectivity beamline from room temperature to 550°C (100°C steps). The studied sample is a mirror-like boron carbide thin film of 400 Å thickness deposited with an adhesive layer of 20 Å Cr on a silicon substrate. The corresponding changes in the film structure are recorded using angle-dependent soft X-ray reflectivity measurements carried out in the region of the boron K-edge after each annealing temperature. Analyses performed using the Parratt recursive formalism reveal that the top boron carbide layer remains intact but interface reactions take place in the buried Cr–SiO2 region. After 300°C the Cr layer diffuses towards the substrate. At higher temperatures of 500°C and 550°C the Cr reacts with the native oxide layer and tends to form a low-density compound of chromium oxysilicide (CrSiOx). Depth profiling of Si and Cr distributions obtained from secondary ion mass spectroscopy measurements corroborate the layer model obtained from the soft X-ray reflectivity analyses. Details of the interface reaction taking place near the substrate region of boron carbide/Cr sample are discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents mechanical testing capability demonstration activities performed in fiscal year (FY)2025 at Oak Ridge National Laboratory(ORNL)on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~13GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide(SiC)thermometry (TM) was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target. Mechanical testing at ambient temperature demonstrated the expected irradiation-induced hardening, with yield strength(YS)and ultimate tensile strength(UTS)values increasing substantially relative to unirradiated counterparts. However, this strengthening was accompanied by a reduction in ductility, as indicated by lower uniform and total elongations(UE and TE). Both coated and uncoated claddings exhibited similar mechanical response, though Cr-coated specimens showed surface cracking perpendicular to the loading direction, attributable to the hardness mismatch between the coating and substrate. Fracture in all cases remained ductile, and no coating spallation was observed following HFIR irradiation. Complementary efforts were directed toward the fabrication of test specimens from commercially irradiated cladding (rod 47I, ~31.1 GWd/t average burnup). Axial sectioning and computer numerical control(CNC)machining successfully produced ATT geometries suitable for benchmarking against HFIR-irradiated specimens. This capability enables direct comparison of cladding behavior between test reactor and commercial reactor environments, thereby supporting the validation of HFIR as a surrogate irradiation platform for accident tolerant fuel (ATF) development. Once HFIR irradiations are completed in FY26,the relevant comparison tests will be completed. Collectively, the FY 2025 PIE campaign has provided mechanical performance data for irradiated advanced claddings The demonstrated capabilities support the framework for mechanical testing and further evaluations in subsequent years. These efforts will represent an important contribution toward the licensing and deployment of Cr-coated zirconium alloy cladding as a near-term ATF solution.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Crack-Free 30% Chromium-Nickel Alloy Welding Products for Nuclear Service

We report prior research in the development of 30% chromium-nickel alloy nuclear welding wires has resulted in the resolution of primary water stress corrosion cracking (PWSCC) and ductility dip cracking (DDC) as well as improvement in solidification cracking (SC) resistance. The resolution of DDC exhibits some Laves phase, which has a negative effect on SC resistance. In this study, the use of an alternate carbide former, tantalum (Ta), in combination with niobium (Nb) was researched. Three heats of recently designed Filler Metal 52MSS-Ta (i.e., HV1648, HV1673A, and VX131WXW) were melted, fabricated, and systematically studied. DDC and SC were evaluated with thermodynamic modeling using the Scheil solidification simulation model, two types of varestraint tests, and strain-to-fracture (STF) testing. The varestraint and STF test results showed an improved SC resistance with reduced Laves phase and concurrent excellent DDC resistance. Optimized compositions with low Laves phase also exhibited high threshold strain values (TSVs) in the STF test. VX131WXW — which contains 2.81 wt-% Ta, 0.6 wt-% Nb, and 6 wt-% iron (Fe) - exhibited a TSV of 24%. Thermo-Calc computed the Laves phase to be 0.24% for VX131WXW compared to 0.06% in HV1673A. This difference in Laves phase resulted in the lower SC resistance of VX131WXW compared to HV1673A when measured with longitudinal varestraint testing. The maximum crack distance for HV1673A was about 0.6 mm while that of heat VX131WXW was about 1.0 mm. The typical diluted weld deposit made with VX131WXW was also resistant to PWSCC due to the chromium content exceeding 24%. These simultaneous results mark progress toward crack-free welds and provide direction for further optimization of Ta-containing filler metals.

30% chromium-nickel alloy↗
DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents post-irradiation examination (PIE) activities performed in FY 2025 at Oak Ridge National Laboratory on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~40 GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide thermometry was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1994

Tribology and Microstructure of PS212 with a Cr2O3 Seal Coat

PS212 is a plasma sprayed metal bonding chrome carbide coating with solid lubricant additives which has lubricating properties at temperatures up to about 900 deg C. The coating is diamond ground to achieve an acceptable tribological surface. But, as with many plasma spray coatings, PS212 is not fully-dense. In this study, a chromium oxide base seal coating is used in an attempt to seal any porosity that is open to the surface of the PS212 coating, and to study the effect of the sealant on the tribological properties of PS212. The results indicate that the seal coating reduces friction and wear when it is applied and then diamond ground leaving a thin layer of seal coating which fills in the surface pits of the PS212 coating.

Sliney, Harold E.↗
DOE OSTI2021

Steam oxidation of chromium corrosion barrier coatings for sic-based accident tolerant fuel cladding

SiC-based materials are currently being considered for accident tolerant fuel (ATF) cladding, however, there are concerns regarding hydrothermal corrosion rates and hermeticity. These issues may be addressed with the utilization of a Cr corrosion barrier coating (CBC) but the interaction of Cr with SiC under loss of coolant accident (LOCA) conditions needs to be investigated. In the present work, Cr coatings were deposited onto chemically vapor deposited (CVD) SiC with either high power impulse magnetron sputtering (HiPIMS), cathodic arc (CA), or a combination of both. Annealing and steam thermogravimetric analysis (TGA) was used to study reaction product formation with and without oxidation. Additionally, after 1200 °C annealing, formation of a Cr x C y carbide layer at the ambient interface and a Cr x Si y C z silicide layer at the SiC interface was observed. After 1200 °C TGA exposure, similar carbide and silicide layers were observed but with a Cr 2 O 3 outer reaction layer and a Si and O rich interfacial layer forming between the carbide and silicide layers. Initial mass gain of the coatings during the 1200 °C TGA exposure was parabolic and were similar to reported rates of chromia formation. Mass gain behavior and observed transitions in chromia microstructure indicate that metallic Cr may be consumed within the first ~1 h at 1200 °C and that subsequent chromia formation occurs at the expense of the underlying Cr carbide layer. Overall, the results show no detrimental impact of Cr coatings on the steam oxidation resistance of SiC up to 4 h at 1200 °C.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Variations in Creep Performance of P92 Steel Within Its Composition Range: Influence of N Solubility

Underspecified composition ranges often lead to alloys with unpredictable mechanical performance. To better understand the changes in microstructure and mechanical performance associated with variations of key elements, three versions of P92 are formulated within, or close to, the specified allowable for N, B, and C ranges. Chromium and Si are also varied to influence N solubility. Different service conditions (i.e., temperature and stress) are explored. It is observed that >80% decrease in creep life occurs at 625 °C and 155 MPa for the highest B and N containing alloy. Multiscale characterization reveals key changes due to the trace element variation. The high B and N containing alloy forms deleterious BN precipitates with morphology that promotes crack nucleation and damage accumulation, but this alloy additionally forms higher fractions of beneficial MX precipitates. The alloy with the lowest B and N concentrations but greater C content shows the best creep performance—a consequence of the refined M 23 C 6 carbide precipitate population and the absence of large-scale inclusions or BN precipitates. Calculations of creep activation energy reveal that the high B and N containing alloy is more prone to damage accumulation which causes an early onset of accelerating creep and greater minimum creep rate.

36 MATERIALS SCIENCE↗