Engineering topics

Gietl, Hanns

Publications and source records attributed to Gietl, Hanns.

10 records

DOE OSTI2024

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

The effect of mixed spectra and thermal neutron shielded irradiation on tungsten was evaluated with plasma exposure in the tritium plasma experiment followed by thermal desorption spectroscopy, X-ray photoelectron spectroscopy, and transmission electron microscopy. The two different irradiation campaigns were performed at the High Flux Isotope Reactor to 0.39–0.74 displacement per atom (dpa) in the 894–1379 K temperature range. A neutron spectrum influence on the void size and void number density was not observed. However, a strong correlation was found between void size and void number density with temperature, but not with dpa in the limited dpa range of this study. Thermal neutron shielding significantly reduced the transmutation to Re+Os. Higher irradiation temperature will lead to larger voids with lower number density, which reduces deuterium retention. In conclusion, grain growth was also observed for high-temperature irradiation of over ~1300 K within the limited grains visible in the transmission electron microscopy specimens.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Hydrogen permeation in iron-chromium-aluminum (FeCrAl) alloys and the effects of microstructure and surface oxide

Iron–chromium–aluminum (FeCrAl) class alloys are candidates for use as cladding for accident-tolerant fuels and moderators. In this context, hydrogen isotope permeation in FeCrAl alloys is an important material property. Here, in the present work, the apparent permeability, effective diffusivity, and apparent solubility of hydrogen in the FeCrAl alloys C26M and Kanthal D (KD) were measured with gas-driven hydrogen permeation. Permeation measurements were conducted at temperatures of 400 to 700 °C and at gas-driven pressures from 1 to 100 kPa. In particular, the effect of grain size on hydrogen transport was studied with KD samples with three different microstructures: nanocrystalline (NC), ultra-fine grained (UFG), and coarse-grained (CG). The UFG and NC specimens had higher apparent activation energies (73.4 kJ mol -1 and 65.2 kJ mol -1 , respectively) for hydrogen permeability than the CG sample (46.9 kJ mol -1 ). An aluminum oxide layer formed on the primary- and secondary-side surfaces of all samples subjected to permeation experiments which demonstrated the propensity of FeCrAl alloys to form these innate oxide permeation barriers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Emerging continuous SiC fibers for high-temperature applications

Silicon carbide (SiC) fibers are responsible for the ultimate strength and toughness of SiC-fiber reinforced composites in harsh environments. The development of a new generation of continuous SiC fibers continues to advance the mechanical properties of composite materials. Tyranno™ SA4 fiber was recently released as a successor of Tyranno™ SA3 fiber. Laser-driven chemical vapor deposition (LCVD) has been adopted as an alternative fiber processing route to synthesizing high-strength SiC fiber with tailorable small diameters and chemical compositions. Both Tyranno™ SA4 and laser-driven CVD fibers show very high tensile strength, about 4 GPa in the as-fabricated condition. The degradation of thermal stability and strength due to annealing in an inert environment were similar for Tyranno™ SA3 and SA4 fibers because of their similar carbon-rich, crystalline microstructure. Silicon-rich fibers produced by LCVD possessed heterogeneous crystallinity, which was attributed to laser power distribution and showed microstructural instability at 1500 °C and above. Finally, the new SiC fibers demonstrated an increase in as-fabricated strength but faced the same challenges in environmental resistance as the traditional SiC fibers do.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Review of Recent Progress in Plasma-Facing Material Joints and Composites in the FRONTIER U.S.-Japan Collaboration

The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, in this work, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material’s fabrication and benefits are described, and some discussion of remaining questions is given.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Corrosion characteristics of monolithic SiC materials in beryllium-bearing molten fluoride salt

In this work, chemical vapor deposited SiC single crystal and polycrystals were exposed to a static beryllium-bearing molten fluoride salt, 2LiF–BeF 2 for 500 and 1000 h at 750 °C. Overall, the single crystal and polycrystalline SiC materials experienced slight weight loss and nominal surface etching regardless of the material type. Localized pitting and nonuniform Si removal from surface regions were associated with trace metal impurities and oxygen. Cross-sectional observation revealed minimal subsurface changes in the microstructure. The corrosion damage processes were similar to those reported for molten LiF–KF–NaF salt with trace impurities, indicating that the impurities significantly affect SiC corrosion.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Heterogeneous microstructure of yttrium hydride and its relation to mechanical properties

Here, the goal of this study is to investigate the properties of yttrium hydride materials in relation to the microstructure, especially its homogeneity. High-throughput nanoindentation mapping was used to evaluate hardness distribution. Raman spectral imaging demonstrated its sensitivity to the presence of YH2 and impurities. Raman peak position maps were correlated with residual stress in the specimens. Electron backscatter diffraction mapping provided phase distributions with correlation to high-energy X-ray diffraction analysis. The experimental mapping data were combined and analyzed using unsupervised machine learning cluster procedures. The machine learning analysis revealed that yttrium hydride specimens contained a major δ-YH2 – x phase component and minor α-Y and δ-YH2 – x components with significant residual stress. The minor phase fraction decreased with increasing nominal H/Y ratio, which affected the nanoindentation and Vickers hardness. The multimodal mapping procedures described herein affect developing important microstructure–property relationships, as well as correlations in heterogeneity and mechanical properties.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Neutron irradiation-enhanced grain growth in tungsten and tungsten alloys

Herein, to understand the microstructural stability of candidate plasma-facing materials under fusion-relevant environments, neutron irradiation of W and W-3%Re alloys with and without K and La dopants was performed in the mixed-spectrum High Flux Isotope Reactor at nominal temperatures of ~850 °C and ~1100 °C to calculated doses between 0.42 and 0.47 dpa. To the best of our knowledge, this study presents the first experimental evidence of radiation-enhanced recrystallization in W and undoped W–Re alloys at ~850 °C, conditions where thermal annealing does not cause any grain growth in a similar timescale. Potassium- or lanthanum-doped tungsten alloys showed more resistance to radiation-enhanced grain growth. We explain the acceleration of grain growth by analyzing the self-diffusion constant under atomic displacement environments. The microstructural observations of the studied W variants suggest that La doping is more effective than K doping for mitigating recrystallization. This study also found that radiation-enhanced recrystallization is an important consideration when designing and applying W to plasma-facing components in future nuclear fusion reactors.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Charpy impact tests of tungsten fiber–reinforced composite from -150 °C to 1000 °C

Tungsten features a unique combination of properties that makes it a favorable plasma-facing material in fusion reactors. The main drawbacks of tungsten are its brittleness at room temperature and the potential embrittlement during operation. Tungsten fiber–reinforced composites have the potential to overcome these limitations. Charpy impact tests ranging from -150 °C to 1000 °C have been performed. To identify the ductile-to-brittle transition temperature and understand the behavior at high temperature a detailed microscopic investigation was used to identify the transition temperature of fiber and matrix, which was between -100 °C to -50 °C and 1000 °C, respectively.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Kinetics of Hydrogen Desorption from Yttrium Hydride: The Development of Material Characterization Methods

Yttrium hydride possesses attractive neutronic and thermal properties for moderator applications as a high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. For safe operation of the nuclear reactors, it is critical to understand the kinetics of H desorption from yttrium hydride at elevated temperatures. In previous studies, the H desorption flux as a function of temperature was measured by thermal desorption spectroscopy. To fully understand the H desorption kinetics, the amount of H within the specimen must be locally evaluated, especially if this uniformity strongly affects the thermomechanical properties and irradiation resistance of the bulk material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Improving the W Coating Uniformity by a COMSOL Model-Based CVD Parameter Study for Denser W f /W Composites

Tungsten (W) has the unique combination of excellent thermal properties, low sputter yield, low hydrogen retention, and acceptable activation. Therefore, W is presently the main candidate for the first wall and armor material for future fusion devices. However, its intrinsic brittleness and its embrittlement during operation bears the risk of a sudden and catastrophic component failure. As a countermeasure, tungsten fiber-reinforced tungsten (W f /W) composites exhibiting extrinsic toughening are being developed. A possible W f /W production route is chemical vapor deposition (CVD) by reducing WF 6 with H 2 on heated W fabrics. The challenge here is that the growing CVD-W can seal gaseous domains leading to strength reducing pores. In previous work, CVD models for W f /W synthesis were developed with COMSOL Multiphysics and validated experimentally. In the present article, these models were applied to conduct a parameter study to optimize the coating uniformity, the relative density, the WF 6 demand, and the process time. A low temperature and a low total pressure increase the process time, but in return lead to very uniform W layers at the micro and macro scales and thus to an optimized relative density of the W f /W composite. High H 2 and low WF 6 gas flow rates lead to a slightly shorter process time and an improved coating uniformity as long as WF 6 is not depleted, which can be avoided by applying the presented reactor model.

36 MATERIALS SCIENCE↗