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At least 1,459 records · Page 81
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Development of Low-Cost, Crack-Tolerant Metallization Using Screen Printing

One of the ways to reduce the cost of solar electricity to 3/kWh, thus reaching parity with fossil-fuel-based generation, is to reduce the degradation rate of solar modules and extend their lifetime well beyond 30 years. The extended module lifetime in turn can positively influence the financial model and the bankability of utility-scale PV projects. Today, the highest-riskpriority solar module degradation mechanism is what is known as hot spots, often induced by cell cracks. In order to address this degradation mechanism, we make use of low-cost, multi-walled carbon nanotubes embedded in commercial screen-printable silver pastes. When the carbon nanotubes are properly functionalized and appropriately incorporated into commercial silver pastes, the resulting metal contacts on solar cells, after screen-printing and firing, show exceptional fracture toughness. These composite metal contacts possess increased ductility, electrical gap-bridging capability up to 50 um, and 'self-healing' to regain electrical continuity even after cycles of complete electrical failure under extreme strain.

14 SOLAR ENERGY↗
DOE OSTI2016

Hydride effects on discharged fuel clad related to accident conditions during dry storage and handling - 2016-0043

Safe handling of nuclear reactor discharged fuel requires an understanding of how the fuel (cladding) will perform under the various encountered conditions, including off-normal events, such as drop events in handling the storage canister. Clad failures under two distinct loading modes are considered: a diametrical pinch mode and an axial-bend mode. The mechanical response to the pinch mode and the bend mode loading are evaluated with ring compression testing (RCT) and three-point bend (TPB) testing, respectively. ZIRLO{sup R} clad samples were hydrogen charged to levels from 100 ppm to 800 ppm and then subjected to a hydride reorientation treatment involving heating to 400 deg. C and cooling while under hoop stresses ranging from 90 MPa to 170 MPa. Radial hydride reorientations were characterized and RCT and TPB tests were performed over a temperature range of ambient to 200 deg. C. Ductile-to-brittle transition temperatures (DBTT) were determined. Using finite element analysis (FEA), true stresses and strains are calculated and comparisons made between the RCT and TPB test modes. Because of the anisotropic properties of the clad and the relative orientation and generic effects of the hydrides, the DBTT was observed to be strongly dependent on the mode of strain application such that a DBTT associated with an RCT test is different from a DBTT obtained using an axial-bend test. Although the degree of radial hydriding strongly affected the pinch-mode RCT test, the same radial hydride impact was not observed for the axial-bend TPB test, which was affected strongly by the bulk hydride levels and not by the fraction of radial hydrides. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Mechanical behavior at high temperatures of highly oxygen- or hydrogen-enriched α and prior-β phases of Zirconium alloys - 2016-0063

During a hypothetical loss-of-coolant accident, zirconium alloy fuel claddings can be loaded by internal pressure and exposed to steam at high temperatures (HTs) up to 1,200 deg. C and then cooled and quenched in water. A significant fraction of the oxygen that reacts with the cladding during HT oxidation diffuses beneath the oxide through the metallic substrate. This induces a progressive transformation of the metallic β{sub Zr}-phase layer into an intermediate layer of α{sub Zr}(O) phase containing up to 7 weight % (wt.%) oxygen. Furthermore, in some specific conditions, the cladding may rapidly absorb a significant amount of hydrogen during steam exposition at HTs. As a β{sub Zr} stabilizer, hydrogen would mainly diffuse and concentrate up to several thousands of parts per million by weight (wppm) into the inner β{sub Zr}-phase layer. This study intends to provide new and more comprehensive data on the HT mechanical behavior of the α{sub Zr}(O) and (prior-)β{sub Zr} phases containing high contents of oxygen and hydrogen, respectively. Model samples produced from M5{sup R} and Zircaloy-4 cladding tubes homogeneously charged in oxygen (less than or equal to 6 wt.%) and hydrogen (less than or equal to 3,000 wppm), respectively, were prepared. Their mechanical behavior was determined under vacuum between 800 and 1,100 deg. C for the oxygen-enriched α{sub Zr} phase and in air between 700 and 20 deg. C after cooling from the β{sub Zr} temperature domain for the hydrogen-enriched (prior-)β{sub Zr} phase. The α{sub Zr} phase was substantially strengthened and embrittled by oxygen. The contribution of the α{sub Zr}(O) layer to the HT creep behavior of an oxidized fuel cladding tube subjected to internal pressure is evaluated by finite element analysis. Mechanical strength and ductility of the model (prior-)β{sub Zr} phase appear to be affected by hydrogen contents of 2000-3000 wppm in ways that depend on temperature. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Equivalent radiation damage in zirconium irradiated in various reactors - 2016-0101

The physical properties and in-reactor performance of core components made from zirconium alloys are a function of the radiation-induced atomic displacements that occur during service. The vacancy and interstitial point defects and their clusters created by the atomic displacements modify the microstructure, and therefore physical properties such as strength and ductility, by forming dislocation loops. The rate of production of point defects also affects time-dependent properties such as creep and diffusion (mass transport). To simulate the effect of irradiation on core components, researchers have used materials test reactors for many years to irradiate and test materials. Such reactor tests can target an accumulated displacement damage dose to simulate the state of a material at a given stage in reactor life but at an accelerated rate. Each nuclear reactor has a specific neutron energy spectrum that also depends on the in-reactor location. Therefore, to use test reactor data to determine the effect of radiation damage on power reactor materials, one must convert the accumulated dose into a unit that is common in its effect on the material properties. For many property changes in nuclear reactor cores, this unit is displacements per atom. This paper compares the damage production for zirconium irradiated under differing neutron energy spectra in various power and test reactors. The damage rates as a function of different measures of neutron flux were used to relate atomic displacements in various materials test reactors to CANDU and light-water reactor conditions. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2011

FCRD Milestone Report: M21AF050901

The objective of this study was to perform mechanical testing on large scale heats of the advanced ODS 14YWT alloy to investigate the effects of processing parameters on mechanical properties. Mechanical properties tests were conducted on two heats of the advanced ODS 14YWT ferritic alloy: the 14YWT-SM11 was produced by extrusion at ORNL and OW4 was produced by HIP at UCSB. The 14YWT-SM11 showed very high tensile strength compared to OW4, but showed less ductility as a result. The fracture toughness transition temperature of 14YWT-SM11 was determined in two orientations and showed T{sub 0} = 48 C in the favorably strong L-T direction while shifting by 63 C to T{sub 0} = 111 C in the weaker T-L direction. The fracture toughness transition temperature for OW4 was not determined but appeared to be within the range observed for 14YWT-SM11. The fracture toughness of 14YWT-SM11 at room temperature was 86.8 MPa{radical}m and 93.1 MPa{radical}m, which was much higher than that of OW4 (27.4 MPa{radical}m). The strain rate jump tests conducted on OW4 indicated that the creep properties were similar to MA957 at 750 C.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Creep Rupture Testing of Alloy 617 and A508/533 Base Metals and Weldments

The NGNP, which is an advanced HTGR concept with emphasis on both electricity and hydrogen production, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 750-1000°C. Alloy 617 is a prime candidate for VHTR structural components such as reactor internals, piping, and heat exchangers in view of its resistance to oxidation and elevated temperature strength. However, lack of adequate data on the performance of the alloy in welded condition prompted to initiate a creep test program at Argonne National Laboratory. In addition, Testing has been initiated to evaluate the creep rupture properties of the pressure vessel steel A508/533 in air and in helium environments. The program, which began in December 2009, was certified for quality assurance NQA-1 requirements during January and February 2010. Specimens were designed and fabricated during March and the tests were initiated in April 2010. During the past year, several creep tests were conducted in air on Alloy 617 base metal and weldment specimens at temperatures of 750, 850, and 950°C. Idaho National Laboratory, using gas tungsten arc welding method with Alloy 617 weld wire, fabricated the weldment specimens. Eight tests were conducted on Alloy 617 base metal specimens and nine were on Alloy 617 weldments. The creep rupture times for the base alloy and weldment tests were up to ~3900 and ~4500 h, respectively. The results showed that the creep rupture lives of weld specimens are much longer than those for the base alloy, when tested under identical test conditions. The test results also showed that the creep strain at fracture is in the range of 7-18% for weldment samples and were much lower than those for the base alloy, under similar test conditions. In general, the weldment specimens showed more of a flat or constant creep rate region than the base metal specimens. The base alloy and the weldment exhibited tertiary creep after 50-60% of the rupture life, irrespective of test temperature in the range of 750-950°C. The results showed that the stress dependence of the creep rate followed a power law for both base alloy and weldments. The data also showed that the stress exponent for creep is the same and one can infer that the same mechanism is operative in both base metal and weldments in the temperature range of the current study. SEM fractography analysis indicated that both base metal and weldment showed combined fracture modes consisting of dimple rupture and intergranular cracking. Intergranular cracking was more evident in the weldment specimens, which is consistent with the observation of lower creep ductility in the weldment than in the base metal.

08 HYDROGEN↗
DOE OSTI2011

Microstructures and oxidation behavior of some Molybdenum based alloys

The advent of Ni based superalloys revolutionized the high temperature alloy industry. These materials are capable of operating in extremely harsh environments, comprising of temperatures around 1050 C, under oxidative conditions. Demands for increased fuel efficiency, however, has highlighted the need for materials that can be used under oxidative conditions at temperatures in excess of 1200 C. The Ni based superalloys are restricted to lower temperatures due to the presence of a number of low melting phases that melt in the 1250 - 1450 C, resulting in softening of the alloys above 1000 C. Therefore, recent research directions have been skewed towards exploring and developing newer alloy systems. This thesis comprises a part of such an effort. Techniques for rapid thermodynamic assessments were developed and applied to two different systems - Mo-Si alloys with transition metal substitutions (and this forms the first part of the thesis) and Ni-Al alloys with added components for providing high temperature strength and ductility. A hierarchical approach towards alloy design indicated the Mo-Ni-Al system as a prospective candidate for high temperature applications. Investigations on microstructures and oxidation behavior, under both isothermal and cyclic conditions, of these alloys constitute the second part of this thesis. It was seen that refractory metal systems show a marked microstructure dependence of oxidation.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to its superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Letter Report Documenting Progress of Second Generation ATF FeCrAl Alloy Fabrication

Development of the 2 nd generation ATF FeCrAl alloy has been initiated, and a candidate alloy was selected for trial tube fabrication through hot-extrusion and gun-drilling processes. Four alloys based on Fe-13Cr-4.5Al-0.15Y in weight percent were newly cast with minor alloying additions of Mo, Si, Nb, and C to promote solid-solution and second-phase precipitate strengthening. The alloy compositions were selected with guidance from computational thermodynamic tools. The lab-scale heats of ~ 600g were arc-melted and drop-cast, homogenized, hot-forged and -rolled, and then annealed producing plate shape samples. An alloy with Mo and Nb additions (C35MN) processed at 800°C exhibits very fine sub-grain structure with the sub-grain size of 1-3μm which exhibited more than 25% better yield and tensile strengths together with decent ductility compared to the other FeCrAl alloys at room temperature. It was found that the Nb addition was key to improving thermal stability of the fine sub-grain structure. Optimally, grains of less than 30 microns are desired, with grains up to and order of magnitude in desired produced through Nb addition. Scale-up effort of the C35MN alloy was made in collaboration with a commercial cast company who has a capability of vacuum induction melting. A 39lb columnar ingot with ~81mm diameter and ~305mm height (with hot-top) was commercially cast, homogenized, hot-extruded, and annealed providing 10mm-diameter bar-shape samples with the fine sub-grain structure. This commercial heat proved consistent with materials produced at ORNL at the lab-scale. Tubes and end caps were machined from the bar sample and provided to another work package for the ATF-1 irradiation campaign in the milestone M3FT-14OR0202251.

2nd generation ATF FeCrAl alloy↗
DOE OSTI2011

Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

The project on irradiation embrittlement of reactor pressure vessel (RPV) steels irradiated at low temperatures (<93°C) was carried out with three major tasks; (1) data collection to match that used in the HFIR steel embrittlement trend curve study published in 1994 by Remec et. al, (2) new embrittlement data of A212B steel that are not included in the earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with consideration of the irradiation temperature effect. A low temperature RPV steel irradiation embrittlement predictive curve was developed, as described below. ΔNDTT(°C) = 23.85 log(x) 2 + 203.3 log (x) + 434.7, with 2-σ uncertainty of 34.6°C, where parameter x is total dpa. The low temperature RPV steel irradiation embrittlement curve has a higher embrittlement rate compared to that of the HFIR trend curve developed in 1994.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Radiation behavior of high-entropy alloys for advanced reactors. Final report

In the first task, we have demonstrated the radiation damage and the recrystallization behaviors in multicomponent alloys through molecular-dynamics simulations. It is found that by alloying with atoms of different sizes, the atomic-level strain increases, and the propensity of the radiation-induced crystalline to amorphous transition increases as the defects cluster in the cascade body. Recrystallization of the radiation induced supercooled or glass regions show that by tuning the composition and the equilibrium temperature, the multicomponent alloys can be healed. The crystalline-amorphous-crystalline transitions predict the potential high radiation resistance in multicomponent alloys. In the second task, three types of high-entropy alloys (HEAs) were fabricated from AlCoCrFeNi and AlCuCrFeNi quinary alloys. Hardness and reduced contact modulus were measured using nanoindentation tests. Heavy ion irradiation were performed using 10 MeV gold and 5 MeV nickel to study radiation effects. Al 0.5 CrCuFeNi 2 shows phase separation upon the presence of copper. Both hardness and contact modulus exhibit the same trend as increasing the applied load, and it indicates that excessive free volume may alter the growth rate of the plastic zone. The as-cast Al 0.1 CoCrFeNi specimen undergone the hot isostatic pressing (HIP) process and steady cooling rate which mitigate the quenching effect. The swelling behavior was characterized by the atomic force microscopy (AFM), and the swelling rate is approximately 0.02% dpa. Selected area diffraction (SAD) patters show irradiation-induced amorphization throughout the ion projected range. Within the peak damage region, an amorpous ring is observed, and a mixture of amorphous/ crystalline structure at deeper depth is found. The Al 0.3 CoCrFeNi HEAs shows good radiation resistance up to 60 peak dpa. No voids or dislocations are observed. The crystal structures remain face-centered-cubic (FCC) before and after 5 MeV Ni irradiation. Higher dpa might be required to study defects formation mechanisms. In the third task, all the constituent binary and ternary systems of the Al-Co-Cr-Fe-Ni system were thermodynamically modeled within the whole composition range. Comparisons between the calculated phase diagrams and literature data are in good agreement. The multi-component thermodynamic database of the Al-Co-Cr-Fe-Ni system was then obtained via extrapolation. The current Al-Co-Cr-Fe-Ni thermodynamic database enables us to carry out the calculations of phase diagrams, which can be used as useful guidelines to identify the Al-Co-Cr-Fe-Ni HEAs with desirable microstructures. In the fourth task, we discuss how as-cast and homogenized phases can be identified, what phases are usually found in the as-cast and homogenized conditions, and what the thermodynamics and kinetics of phase transformations are in the AlCoCrFeNi HEA. The microstructure and phase composition were studied in as-cast and homogenized conditions. It showed the dendritrical structure in the as-cast condition consisting primarily of a nano-lamellar mixture of A2 [disorder body-centered-cubic (BCC)] and B2 (ordered BCC) phases, in addition to a very small amount of A1 [disorder face-centered-cubic (FCC)] phases. The homogenization heat treatment resulted in an increase in the volume fraction of the A1 phase and formation of a Sigma phase. Tensile properties in as-cast and homogenized conditions are reported at 700 °C. Thermodynamic modeling of non-equilibrium and equilibrium phase diagrams for the AlCoCrFeNi HEA gave good agreement with the experimental observations of the phase contents. The reasons for the improvement of ductility after the heat treatment are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Uncertainty Quantification and Management for Multi-scale Nuclear Materials Modeling

Understanding and improving microstructural mechanical stability in metals and alloys is central to the development of high strength and high ductility materials for cladding and cores structures in advanced fast reactors. Design and enhancement of radiation-induced damage tolerant alloys are facilitated by better understanding the connection of various unit processes to collective responses in a multiscale model chain, including: dislocation nucleation, absorption and desorption at interfaces; vacancy production, radiation-induced segregation of Cr and Ni at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels; investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S); time evolution of swelling (cluster growth) phenomena of irradiated materials; and energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Optimized Gen-II FeCrAl cladding production in large quantity for campaign testing

There are two major objectives in this report; (1) to optimize microstructure control of ATF FeCrAl alloys during tube drawing processes, and (2) to provide an update on the progress of ATF FeCrAl tube production via commercial manufacturers. Experimental efforts have been made to optimize the process parameters balancing the tube fabricability, especially for tube drawing processes, and microstructure control of the final tube products. Lab-scale sheet materials of Gen II FeCrAl alloys (Mo-containing and Nb-containing FeCrAl alloys) were used in the study, combined with a stepwise warm-rolling process and intermediate annealing, aiming to simulate the tube drawing process in a commercial tube manufacturer. The intermediate annealing at 650ºC for 1h was suggested for the tube-drawing process of Mo-containing FeCrAl alloys because it successfully softened the material by recovering the work hardening introduced through the rolling step, without inducing grain coarsening due to recrystallization. The final tube product is expected to have stabilized deformed microstructure providing the improved tensile properties with sufficient ductility. Optimization efforts on Nb-containing FeCrAl alloys focused on the effect of alloying additions and annealing conditions on the stability of deformed microstructure. Relationships between the second-phase precipitates (Fe 2 Nb-Laves phase) and microstructure stability are discussed. FeCrAl tube production through commercial tube manufacturers is currently in progress. Three different manufacturers, Century Tubes, Inc. (CTI), Rhenium Alloys, Inc. (RAI), and Superior Tube Company, Inc. (STC), are providing capabilities for cold-drawing, warm-drawing, and HPTR cold-pilgering, respectively. The first two companies are currently working on large quantity tube production (expected 250 ft length) of Gen I model FeCrAl alloy (B136Y3, at CTI) and Gen II (C35M4, at RAI), with the process parameters obtained from the experimental efforts. The expected delivery dates are at the end of July, 2016, and the middle of June, 2016, respectively. Tube production at STC would be the first attempt to apply cold-pilgering to the FeCrAl alloys. Communication has been initiated, and the materials have been machined for the cold-pilgering process.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Final Report: Self-Consolidating Concrete Construction for Modular Units

This report focuses on work completed on DE-NE0000667, Self-Consolidating Concrete for Modular Units, in connection with the Department of Energy Nuclear Energy Enabling Technologies (DOE-NEET) program. This project was completed in the School of Civil and Environmental Engineering at the Georgia Institute of Technology, with Westinghouse Corporation as the industrial partner. The primary objective of this project was to develop self-consolidating concrete (also termed “self-compacting concrete” or SCC) mixtures so that concrete placement can be made into steel plate composite (SC) modular structures without the need for continuous concrete placement. As part of the research, SCC mixtures were developed and validated to ensure sufficient shear capacity across cold-joints, while minimizing shrinkage and temperature increase during curing to enhance concrete bonding with the steel plate construction found in modular units. The SCC mixtures developed were able to carry shearing forces across the cold-joint boundaries. This “self-roughening” was achieved by adding a tailored fraction of lightweight aggregate (LWA) to the concrete mix, some of which raised to the surface during curing, forming a rough surface on which subsequent concrete placements were made. The self-roughening behavior was validated through three sets of structural tests. Shear friction on small-scale specimens with cold joints was assessed using varying fractions of LWA and with varying amounts of external steel plate reinforcement. The results show that the shear friction coefficient, to be used with the provisions of ACI 318-14, Section 22.9, can be taken as 1.35. Mid-scale beam tests were completed to assess the cold-joint capacity in both in-plane and out-of-plane bending. The results showed that the self-roughened joints performed as well as monolithic joints. The final assessment was a full-scale test using a steel composite module supplied by Westinghouse and similar in construction to the steel composite modules being assembled at the Vogtle and V.C. Summer plant expansions. The final test showed that the roughened cold-joint showed excellent shear and flexural capacity, and substantial ductility, when used in conjunction with steel composite construction.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2020

Localized Strain Analysis of Ce- and Mg-Treated Cast Iron under Uniaxial Compression

Cast iron exhibits a wide range of mechanical properties, depending on its microstructural features. The microstructure of cast iron consists of several microconstituents with different elastic-plastic behavior, making the strain non-uniform across the bulk material. To understand the individual effects of these microconstituents on the overall mechanical behavior, local strain analysis using digital image correlation analysis was carried out. Samples with two different compositions (varying cerium, magnesium and silicon) were processed at different solidification velocities in a Bridgman furnace. Sections of the directionally solidified samples were loaded under uniaxial compression to measure global and local strain behavior. Despite the variability of the microstructure, the stress–strain curves obtained by digital image correlation (DIC) were found to react in a well-controlled way to changes in solidification velocity. It was observed that high-strain failure (greater than 15%) was accompanied by local straining of the softer ferritic phase, but during low-strain failure, local straining was not prominent. Higher nodularities, due to higher solidification velocities, raised the compressive strength without affecting the toughness significantly. Higher percentages of carbides led to higher compressive strengths with corresponding losses in ductility. The continuity of the matrix was also found to play an important role in the behavior during compression.

ductile failure↗
DOE OSTI2019

High Resolution Mapping of Orientation and Strain Gradients in Metals by Synchrotron 3D X-ray Laue Microdiffraction

Synchrotron 3D X-ray Laue microdiffraction, available at beamline 34-ID-E at Advanced Photon Source in Argonne National Laboratory, is a powerful tool for 3D non-destructive mapping of local orientations and strains at sub-micron scale in the bulk. With this technique, it is possible to study local residual stresses developed during manufacturing or while in service due to interactions between, for example, different phases and/or grains with different orientations in materials containing multiple or single phase(s). Such information is essential for understanding mechanical properties and designing advanced materials, but is largely non-existent in the current generation of materials models. In the present paper, the principle and experimental set-up of the 3D microdiffraction are introduced, followed by a description of a method for quantification of the local plastic deformation based on high-angular-resolution orientation maps. The quantification of local residual stresses in two model materials, ductile cast iron (two phases) and partially recrystallized pure nickel (single phase), using 3D microdiffraction will then be presented. The results show that 3D microdiffraction is important for understanding the origin of local residual stresses and to relate them to the microstructural evolution. Finally, the limitations of the 3D microdiffraction on the current generation synchrotron source and new possibilities after the synchrotron upgrade are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Complex Nano-Scale Structures for Unprecedented Properties in Steels

Processing bulk nanoscrystalline materials for structural applications still poses a rather large challenge, particularly in achieving an industrially viable process. In this context, recent work has proved that complex nanoscale steel structures can be formed by solid reaction at low temperatures. These nanocrystalline bainitic steels present the highest strength ever recorded, unprecedented ductility, fatigue on par with commercial bearing steels and exceptional rolling-sliding wear performances. In this paper, a description of the characteristics and significance of these remarkable structures in the context of the atomic mechanism of transformation is provided.

36 MATERIALS SCIENCE↗