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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 19 records

Rapid thermal processing of Czochralski silicon substrates: Defects, denuded zones, and minority carrier lifetime

Rapid thermal processing (RTP) of Czochralski (Cz) silicon substrates is discussed with its attendant effects on defects, denuded zones, and minority carrier lifetime. Preferential chemical etching and X-ray topography was used to delineate defects which were subsequently correlated with minority carrier lifetime; determined by a pulse metallo-organic decompositon (MOD) test device. The X-ray delineation of grown-in defects was enhanced by a lithium decoration procedure. Results, thus far, show excellent correlation between process-induced defects.

Rozgonyi, G. S.↗

Microstructure and strengthening of Al-6Ce-3Ni-0.7Fe (wt%) alloy manufactured by laser powder-bed fusion

In this study, an Al-6Ce-3Ni-0.7Fe (wt%) alloy was fabricated via laser powder-bed fusion and its microstructure, thermal stability, tensile properties, and creep properties are investigated for two rapid-solidification rates with different eutectic spacings. For both faster- and slower-cooled states, the as-fabricated alloy mostly shows elongated grains with very fine eutectic networks (~40 nm lamellar width) comprising a high-volume fraction (~18 %) of intermetallic phases. Faster-cooled samples, however, display a less continuous eutectic network with finer spacing (~100 nm), leading to higher Orowan strengthening and therefore superior microhardness and yield stress up to ~350 °C. Upon aging (300 - 450 °C), micron-size Al 9 (Ni,Fe) 2 needle-like precipitates form within grains, and the Al 11 Ce 3 eutectic network spheroidizes, while retaining a submicron width and spacing, resulting in 33–37 % drop in microhardness after 144 h aging at 400 °C. Limited tensile ductility (~6 %) and creep ductility (~1-2 %) are measured for both faster- and slower-cooled alloys due to an inhomogeneous microstructure, where cavitation preferentially initiates at melt-pool boundaries, precipitate-free zones, and/or denuded zones, eventually leading to local fracture. Denuded zones, induced by stress, form due to diffusional flow with stress-dependent orientations, and are identified as microstructurally weak regions leading to strain localization. The present alloy does not show a significant difference in 300 °C creep resistance between: (i) tensile and compressive loading, (ii) faster- and slower-cooled samples, and (iii) continuous and spheroidized eutectics. Creep resistance at 300 °C is comparable to that of a Ce-richer Al-10.5Ce-3.1Ni-1.2Mn (wt%) alloy, despite the formation of denuded zones and needle-like Al 9 (Ni,Fe) 2 precipitates.

36 MATERIALS SCIENCE↗

Analysis of position-dependent cavity parameters in irradiated metals to obtain insight on fundamental defect migration phenomena

Motion of point defects is a fundamental process that governs microstructure and properties of materials. Here, we examine the near-surface and grain boundary cavity swelling depth profiles in neutron- and ion-irradiated simple metals (Cu, Ni, and Fe-Cr) and investigate diffusional broadening of implanted Ni ions in Fe-Cr alloys. Vacancy migration energies and radiation-enhanced diffusion were experimentally estimated. Cavity denuded zone widths near planar sinks are shown to be dependent on temperature, damage rate, balance of point defects and sinks, and vacancy migration energies. An enhanced cavity swelling zone adjacent to the void-denuded zone was observed in specimens with low to moderate sink strength and interpreted as evidence of 1D gliding interstitial clusters. Radiation-enhanced diffusional broadening of implanted Ni ions (at 400–550 °C) in Fe and Fe-Cr was up to 250 nm toward the surface, which leads to a much broader near-peak-damage region where cavity swelling is suppressed. Diffusional broadening of implanted ions is calculated to be similarly pronounced for self-ion irradiations, particularly near the peak and higher swelling temperature regimes. Adequately high ion energies (~8–15 MeV) are recommended for ion-irradiation studies to provide a sufficiently broad midrange safe analysis region with minimized surface and injected ion effects.

36 MATERIALS SCIENCE↗

Void-interface wetting to crossing transition owing to bubble to void transformation

Understanding cavity-interface interaction is crucial in designing high-strength, radiation-tolerant nanocomposites. In this work, bubbles near the Cu–Nb interface in Nb could be absorbed by Cu voids wetting the interface due to the high system energy difference produced by the huge pressure difference between the bubbles and voids and high mobility of bubbles, and no bubble-denuded zone forms owing to fast Brownian motion of bubbles, which keeps bubbles distributed homogeneously. However, owing to the low system energy difference generated by the low internal pressure difference and low mobility, voids near the Cu–Nb interface in Nb would simply coalesce with Cu voids wetting the interface, leading to the void-interface wetting to crossing transition and the formation of void-denuded zones due to the negligible migration of internal Nb voids.

36 MATERIALS SCIENCE↗

Grain boundary strain as a determinant of localized sink efficiency

The opportunity to achieve radiation tolerance in crystalline materials hinges on understanding the structure and response of grain boundary sinks to irradiation. A common descriptor of grain boundary efficiency as a defect sink is the denuded zone, which is a defect free zone adjacent to the grain boundary dictated by its ability to absorb radiation induced defects. This descriptor is often used at the mesoscale, which requires an averaging of absorption events. In this paper, we resolve the defect sink efficiency as a function of interfacial strain with respect to grain boundary character, and correlate high levels of grain boundary strain to an enhanced absorption efficiency. Here, we also introduce a key relationship between localized strain in proximity with the grain boundary sink and the variation absorption efficiency associated with these regions, revealing the pitfalls of averaging absorption events along a grain boundary, and presenting a path forward toward improved models for denuded zones and localized grain boundary absorption phenomena.

36 MATERIALS SCIENCE↗

Monte Carlo simulations of Cu/Ni–Si–Mn co-precipitation in duplex stainless steels

First-Passage Kinetic Monte Carlo (FPKMC) simulations of species migration in duplex stainless steels were performed in order to establish relationships between alloy content, segregation behavior, and the formation of Ni-Si-Mn rich particles in cast duplex stainless steels during thermal aging. The Ni-Si-Mn-rich second phase forms after extended aging at reactor operating temperatures and degrades alloy properties. Simulations of Ni-Si-Mn cluster formation were validated through comparison with experimental results obtained through Atom Probe Tomography (APT) on similar alloy compositions, identifying several trends. First, Cu promotes the formation of Ni-Si-Mn clusters, but only when Ni or Mn prefer segregation to the surface of Cu particles; without the segregation of these species, the critical composition for the clusters to form was not achieved. Second, the width of precipitate-denuded zones near ?/d interfaces increases with decreasing Cu content. This finding was in strong agreement with APT results, further validating our model. Finally, our model predicts that Ni-Si-Mn cluster formation will be the most extensive when the Si:Mn ratio is approximately 1:1 and the least extensive when one of these key elements is less concentrated (Mn = 0.5 at.% or Si = 0.5 at.%). The implications of these results on how to improve alloy properties are discussed.

Cu co-precipitation, Atom probe tomography (APT), ↗

A quantitative method to determine the region not influenced by injected interstitial and surface effects during void swelling in ion-irradiated metals

We propose and demonstrate a microstructurally-based experimental method to quantitively determine the depth regions in self-ion-irradiated metals that are affected by the injected interstitial effect and various surface effects, focusing on the choice of safe analysis zones to minimize the impact of these phenomena. The goal is to define the depth ranges where extracted data can be confidently applied to ion-neutron correlations for reactor application. Since ion energies in the range of 1–5 MeV are most frequently employed by the radiation effects community, irradiations were conducted at four energies in this range, all proceeding at 475 °C. The experiment was conducted on relatively pure single crystal iron to focus only on physical phenomena, avoiding the influence of possibly confounding chemical or segregation processes. Care was also taken to minimize the influence of other physical factors such as crystalline orientation. It was shown that, at 475 °C, ion energies of ≤1 MeV were too shallow in penetration and did not yield a safe depth range, but irradiations at 2.5 MeV and above yielded useful safe zones with predicted swelling behavior becoming independent of ion energy. The surface-affected zone width was found to be roughly twice that of the void-denuded zone width and to be independent of accumulated displacement dose. The largest injected-interstitial effect arises from the injected interstitial depression of void nucleation and growth. The interstitial-affected region starts at about one half of the projected range and does not show any “spreading” of its influence in depth as the peak damage level increases from 50 to 100 dpa. This study provides some confidence that enhances the credibility of ion simulation when applied to prediction of void swelling in neutron environments.

36 MATERIALS SCIENCE↗

Creep ductility limiting mechanisms in an additively manufactured Al-Ce-Ni-Mn-Zr alloy

Tensile creep response and cavitation damage evolution in an additively manufactured Al-7.5Ce-4.5Ni-0.4Mn-0.7Zr (wt%) alloy with peak-aging and overaging treatments were investigated in the 300–400 ºC range. Microstructural heterogeneity and its response to heat treatment and subsequent creep deformation were studied to understand the interplay between cavity formation, creep lifetime and ductility. Increasing the applied stress activated the nucleation of more cavities, an experimental observation that is well described using the vacancy accumulation model. Cavities nucleated prematurely due to localized plasticity in the denuded zones that formed at/near melt-pool or grain boundaries. Microstructure/deformation heterogeneity with consequent evolution of stress triaxiality, especially at lower stresses, causes accelerated cavitation, thus producing low creep ductility (∼ 0.2–2.4 %), compared to (∼12–21 %) ductility of the alloy measured by regular tensile tests at equivalent temperatures. A constrained diffusional cavity growth mechanism with continuous cavity nucleation during creep is established as the dominant mechanism, implying that cavitation involves vacancy diffusion, yet its growth rate is dictated by the minimum creep rate. In conclusion, the ductility-limiting creep and cavitation mechanisms discussed here provide new insight into the creep behavior of 3D-printed metallic alloys.

Additive manufacturing↗

Revealing the synergistic effects of sequential and simultaneous dual beam irradiations in tungsten via in-situ TEM

Elucidating the synergistic effects of different energetic beams on the radiation response of nuclear materials is critical for developing an improved methodology for their evaluation when exposed to extreme environments. This article describes in-situ sequential (He implantation followed by Kr irradiation and vice-versa) and simultaneous (heavy ion Kr irradiation and He implantation) dual beam irradiations performed on tungsten at 1223 K. Dislocation loop density, average area, and total loop damage as a function of irradiation history and dose/fluence are quantified. The loop Burgers vectors and cavity damage (cavity density, size and total change in volume) are also determined at the final maximum dose for each condition. The loop damage evolution was different in all cases, with the smallest loop damage observed in the simultaneous experiment. Annihilation of Kr generated dislocation loops during He implantation in the Kr + He experiment was an unanticipated observation that may be explained by the dynamic evolution of dislocation loop sink strengths and time-dependent defect fluxes. Dislocation loop raft formation, denuded zones near extended defects, and cavity damage are compared across the different conditions. The phenomena observed and discussed in this work will stimulate further experimental and computational modeling activities leading to improved fundamental understanding of the irradiation response of nuclear materials under reactor-similar environments.

36 MATERIALS SCIENCE↗

Atom probe tomography of segregation at grain boundaries and gas bubbles in neutron irradiated U-10 wt% Mo fuel

During neutron irradiation to fission densities > 5.2 × 10 21 fiss/cm 3 , Xe agglomerates forming gas bubbles of varying size within the U-Mo fuel matrix. Herein, segregation of fission products to Xe bubbles and grain boundaries (GB) were studied using atom probe tomography (APT). Segregation behavior was found to vary among GBs, small bubbles (<10 nm), and larger bubbles (>10 nm). Solid fission products were enriched at GBs and larger bubbles, but not at small bubbles. Finally, a denuded zone was identified adjacent to a > 10 nm Xe gas bubble and a GB.

36 MATERIALS SCIENCE↗

Directed energy deposition of Haynes 282 nickel superalloy

Laser powder directed energy deposition was employed to fabricate Haynes® 282 Ni-based superalloy for high temperature creep testing. Material was heat treated using a single-step aging heat treatment of 800 °C for 4 h. Microstructural analysis was conducted in the as-deposited, solution-annealed, and age-hardened conditions. γ′ precipitates were absent in the as-deposited and annealed condition but formed during aging with an average size of 48.5 ± 12 nm. γ’ denuded zones ranging in thickness from 73 to 469 nm were observed adjacent to an effectively continuous grain boundary carbide film that covered 72.5% of the measured boundary length. The heat-treated material contained a high-volume fraction of TiC carbides ranging in size from 38 to 457 nm. Electron backscatter diffraction showed partial recrystallization, with 77-83% recrystallized area and a heterogeneous grain-size distribution. Creep testing at 750 °C at 300, 320, and 350 MPa revealed significantly inferior performance compared to wrought material, with higher strain rates, earlier tertiary onset, and shorter rupture lives. Fractography shows that failure was dominated by intergranular rupture. Post rupture microstructural analysis results suggest that the creep behavior was primarily governed by the non-equilibrium interface network rather than the bulk γ/γ’ matrix. In conclusion, the findings indicate that achieving wrought-like creep performance necessitates substantial improvements in the additive process, specifically focusing on enhanced homogenization, full recrystallization, and precise control over the morphology and distribution of carbides.

36 - MATERIALS SCIENCE↗

Development of an Open-source Alloy Selection and Lifetime Assessment Tool for Structural Components in CSP

Lack of sufficient data on high temperature mechanical and corrosion behavior of structural materials is a huge barrier in the technological maturity of current and future Concentrating Solar Power (CSP) technologies. Rapid development and selection of materials cannot be achieved by expensive and time-consuming acquisition of experimental data. The goal of the proposed work is development of an open-source alloy selection and lifetime prediction tool that will integrate validated physics-based models to describe influence of temperature, alloy composition, environment and component geometry (thickness) on mechanical and corrosion behavior of Ni and Fe-based alloys employed in molten salts/sCO 2 heat exchangers. This one-year project leveraged the extensive dataset on the creep\corrosion behavior of candidate materials generated at ORNL through past projects and input from current collaborations with industrial partners. Based on previous experience and the feedback provided by industry (Brayton Energy and Echogen), three candidate materials of interest, Ni-based alloys 740H, 282 and 625 and application-specific operating conditions (max. temperature of 730 °C and stress of 150 MPa) were identified for the heat exchanger. An extensive corrosion and creep dataset was assimilated for the relevant operating conditions and was supported by detailed characterization of about 100 metallographic cross-sections. The corrosion dataset consisted of scanning electron microscopy images (secondary electron and backscatter electron), measured concentration profiles of alloying elements using energy dispersive X-ray spectroscopy (EDS), widths of denuded zones (dissolution of strengthening phases) and depths of attack in molten KCl-MgCl 2 mixtures using image analyses. The creep dataset comprised of creep rupture data and creep strain curves (for 740H and 282). Coupled thermodynamic-kinetic microstructure-based models were employed to predict the stress-corrosion induced compositional and phase evolutions in the alloy during operation under the identified operating conditions. Reduced order models were developed from advanced physics-based models and were integrated in a user-friendly alloy selection tool. The corrosion model was able to predict the time to a critical Cr concentration at the oxide/alloy interface (chemical lifetime) within ±10% (1 standard deviation) of typical statistical variation in corrosion tests and EDS measurement errors (±0.5 wt%). The initial scope of the project was limited to predict creep rupture times (Larson-Miller parameter). Based on the input provided by industry, the mechanical lifetime of the heat exchanger is governed by accumulated creep strains (2%) rather than creep rupture. To be able to predict the times to specific creep strains, a more extensive creep model development was undertaken largely beyond the initial scope of the project. The continuum damage mechanics creep model was able to predict times to 2% creep strain, t 2% with an accuracy of ±500h. Ultimately, a screening protocol for SiC was generated to demonstrate the pathway for integration of one of the currently immature materials from a commercial adoption standpoint in the current material evaluation tool. The modeling tool developed here is accessible to the science community and stakeholders and lays the foundation for methods that will enable a rapid evaluation of optimum materials for CSP applications and reliable prediction of material degradation thereby considerably reducing operational costs, improving reliability and increasing overhaul intervals. However, the complete potential of such a tool to include a wider range of materials and test conditions can only be realized with a more concentrated combined experimental-characterization-computation effort.

14 SOLAR ENERGY↗

Microstructural Characterization of the Second High Fluence Baffle-Former Bolt Retrieved from a Westinghouse Two-loop Downflow Type PWR

As one of the pressurized water reactor (PWR) internal components, baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts. In support of evaluating long-term operational performance of materials used in core internal components, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, Materials Research Pathway (MRP) has harvested two high fluence BFBs from a commercial Westinghouse two-loop downflow type PWR. The two bolts of interest, i.e. bolts # 4412 and 4416, were withdrawn from service in 2011 as part of a preventative replacement plan. No identification of cracking or potential damage was found for these bolts during their removal in 2011. However, the bolts required a lower torque for removal from the baffle structure than the original torque specified during installation. Irradiation displacement damage levels in the bolts range from 15 to 41 displacements per atom. The goal of this project is to perform detailed microstructural and mechanical property characterization of BFBs following in-service exposures. The information from these bolts will be integral to the LWRS program initiatives in evaluating end of life microstructure and properties. Furthermore, valuable data will be obtained that can be incorporated into model predictions of long-term irradiation behavior and compared to results obtained in high flux experimental reactor conditions. In this report, we present our latest study in FY22 on microstructural characterizations of the second high fluence baffle-former bolt, i.e., bolt # 4412. Analytical electron microscopy and atom probe tomography characterization were performed. The radiation-induced defects in the material add to the large wealth of knowledge for neutron-induced defects in 304/316 grades of stainless steels, specifically for radiation-induced precipitation after high fluence commercial PWR irradiation. The main findings are summarized as follows: 1) The cavity size was considerably larger in the bolt thread section than in the bolt head, with the bolt thread section having a bimodal distribution of cavities greater than ~6 nm in diameter and less than ~3 nm in diameter. The bolt head only had the small-sized cavities. In addition, there was a denuded zone of large cavities near grain boundaries in the thread section of the bolt. 2) Radiation-induced precipitation in the BFB #4412 was highly complex, with the volume fraction, size, and number density of Ni/Si and Cu-rich precipitates depending strongly on the radiation temperature/dose. In many cases, co-precipitates of adjoined clusters were found with Ni/Si-rich precipitates sandwiched between Cu-rich clusters and Mo/Cr/P-rich clusters. 3) Solute segregation out of solution was highest for most solutes in the thread section of the bolt #4412 with the exception of Cu, which experienced more separation out of solution into Curich clusters in the bolt head section. This highlights the difference in the mechanisms for precipitation of Ni/Si clusters, which have the Ni 3 Si phase composition, and precipitation of Cu-rich clusters. 4) There appear to be multiple simultaneous influences that affect the microstructural variation along the length of the bolt that overcomes the ~2X difference in irradiation dose between the bolt head and the bolt thread. The irradiation temperature, thermal/fast neutron ratio variation, potential strain gradient, and exposure to PWR coolant water that each section of the bolt sees may have more influence on the microstructural evolution than the total irradiation dose. The bolt thread and shank, with higher temperature, higher relative fast neutron flux, higher strain, and exposure to coolant but lower dose, underwent more enhanced cavity formation, precipitation, and solute segregation than the bolt head section.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of Corrosion Resistant Coatings for Structural Materials for Liquid Fueled Molten Salts Reactors Applications (Final Report)

Current structural alloys, code certified for the MSRs temperature ranges, contain high levels of chromium, making them highly susceptible to molten halide salt corrosion. One potential solution to circumvent the need for code certification of novel alloys, which is expensive and time consuming, is to design claddings that will protect the underlying code certified materials from corrosion damage during operation. In this work, we examined the corrosion of Ni and Cu electroplated, Ni and CuNi weld overlay, Mo-laser clad, and carburized claddings on SS316H for use in molten salt reactor environments. After characterization of the cladded materials (Task 1), static corrosion tests were used to assess corrosion resistance of the claddings in typical molten fluoride salts (Task 2). The corrosion tests were performed in molten FLiNaK at 700°C up to 1000 hours. Pre- and post-corrosion Scanning Electron Microscopy (SEM), energy-dispersive-spectroscopy (EDS), Transmission Electron Microscopy (TEM), X-ray Diffraction (XRD) and glow-discharge-optical-emission-spectroscopy (GDOES) were performed on cladding cross-sections and surfaces to evaluate degradation. The Cu and Ni electroplated samples, as well as the carburized samples, showed excellent corrosion resistance relative to the bare SS316H. To assess high temperature cladding stability, ageing experiments were performed at temperatures up to 900 C in inert atmosphere for the electroplated and carburized samples and a diffusion model was developed to predict long term cladding behavior (Task 3). It was found that the Cu cladding was basically insensitive to the high temperature ageing, except for small secondary phases forming at the interface. On the other hand, the Ni electroplated cladding experienced significant interdiffusion. Nevertheless, the gain in corrosion resistance for a 100μm Ni electroplated cladding is phenomenal, with more than 50% reduction in chromium dissolution from the substrate material for the first 15 years of salt exposure at 700°C. To assess radiation resistance and phase stability, high-temperature 4MeV Ni heavy ion irradiation was performed across the cladding/substrate interface up to 50 displacement-per-atom (DPA) at 500°C and 700°C to assess the phase stability and irradiation behaviors of the electroplated systems (Task 4). The Ni and Cu electroplated systems did not experience void swelling at the contrary to the SS316H substrate due to their nanocrystalline nature. Some level of recrystallization was observed in the cladding, as well as radiation induced segregation and enhanced diffusion. The interface acted as a potent sink for point defects with the presence of a void denuded zone. The mechanical properties of the claddings were assessed using thermal shock, micro-indentation, nano-indentation, and four-point bend testing experiments (Task 5). These experiments were performed on the claddings in as-received, corroded, irradiated, and thermally aged states to determine the effects of typical molten salt reactor environments on cladding integrity. No significant mass loss was observed after repeated thermal shocks. While the electroplated samples softened after high-temperature ageing, irradiation hardening compensate this effect, such that there is little different with the as-received materials. The results of these experiments suggest that the electroplated (copper/nickel) show the greatest promise for application in molten salt reactor development. These claddings prevented any chromium dissolution from occurring during the static corrosion experiments. Additionally, they demonstrated favorable properties for high temperature diffusion, phase stability, interfacial mechanical properties, and irradiation resistance. Weld-overlay cladding are also of great interest and could reach properties similar to the electroplated samples upon optimization (especially with multiple weld passes). The carburized SS316H showed excellent behavior as well, but more work needs to be done to assess their long-term stability. Finally, the Mo-laser clad system was not pursued further due to the manufacturing challenges. It is also worth noting that the Ni cladding systems should behave relatively well in terms of weldability since the Ni-weld overlay microstructure and associated corrosion rate are sound. While the NiCu weld-overlay has even lower corrosion rates than the Ni-weld overlay, more studies on welding of Cu-electroplated systems are necessary since Cu clusters are known to embrittle steels.

36 MATERIALS SCIENCE↗

The effect of heat treatment on the defect evolution in LPBF 316H stainless steel

This work investigates the effect of processing and heat treatment on defect evolution and irradiation response of laser powder bed fusion (LPBF) 316H stainless steel, with comparisons to LPBF 316L and wrought 316L/316H. Using in-situ and ex-situ ion irradiations across a wide parameter space—temperature (300–675 °C), dose (0.2–25 dpa), dose rate (10 -3 –10 -5 dpa/s), and helium co-implantation (20–2500 appm)—we correlated void swelling, dislocation loop evolution, and segregation behavior with pre-irradiation microstructures (as-built, stress-relieved, solution-annealed, and cold-worked). Results show that swelling is strongly controlled by dislocation density: intermediate densities maximize swelling, while solution annealing or cold working reduce susceptibility to levels comparable to wrought alloys. Low dose rates and helium both promote cavity nucleation and lower the incubation barrier, with helium suppressing the role of dislocation density and driving swelling behavior toward wrought-like response. Loop evolution in SA LPBF 316H resembles wrought 316L but shows localized denuded zones near low-angle grain boundaries. STEM-EDS mapping further revealed Ni segregation at voids and sparse Al-rich oxides, without evidence of Ni–Si precipitates. Collectively, these findings identify dislocation density, helium content, and dose rate as the factors governing swelling in LPBF 316H and provide mechanistic datasets for model validation, supporting LAIN and the qualification of AM austenitic steels for nuclear service.

316 stainless steel↗

Precipitation of silicon from splat-cooled Al-Si alloys

Splat cooled Al-Si solid solutions with 1 to 11 at.% Si were prepared and their precipitation kinetics were studied by transmission electron microscopy. The time required for appearance of particles visible at a magnification of 35,000 times was determined at temperatures between 248 K and 573 K. The resulting Arrhenius plots yielded activation energies ranging from 55 to 40 plus or minus 2kJ/mol over the composition range. Precipitate densities were higher and denuded zones of 100 to 150 nm were narrower than in comparable solid quenched samples. The activation energies are explained in terms of excess point defect concentrations.

Matyja, H.↗

Carburization and heat treatment to cause carbide precipitation in gamma/gamma prime-delta eutectic alloys

In an attempt to improve their longitudinal shear strength, several directionally solidified eutectic alloy compositions with minor element modifications were pact, carburized, and heat treated to provide selective carbide precipitation at the cell and grain boundaries. The directionally solidified Ni-17.8 Nb-6Cr-2.5Al-3Ta (weight percent) alloy was selected for the shear strength evaluation because it showed the shallowest delta-denuded zone at the carburized surface. The carburization-carbide precipitation treatment, however, did not appear to improve the longitudinal shear strength of the alloy.

Tewari, S. N.↗

Oxygen and carbon impurities and related defects in silicon

Oxygen and carbon are the predominant impurities in Czochralski-grown silicon. The incorporation of oxygen and carbon during crystal growth is reviewed and device effects are discussed. Methods for controlling oxygen and carbon incorporation during crystal growth are discussed and results supporting a segregation coefficient of k=0.5 for oxygen are presented. The nucleation and precipitation behavior of oxygen is complex. Temperature and doping level effects which add insight into the role of point defects in the nucleation process are highlighted. In general, precipitation is found to be retarded in N+ and P+ silicon. The types and quantities of defects resulting from the oxygen precipitates is of interest as they are technologically useful in the process called intrinsic gettering. A comparison is made between the available defect sites and the quantities of metallic impurities present in a typical wafer which need to be gettered. Finally, a discussion of the denuded-zone, intrinsic-gettered (DZ-IG) structure on device properties is presented.

Pearce, C. W.↗