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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 1,027 records · Page 57
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Cavity Swelling in Irradiated Materials

Cavity swelling, due to three-dimensional clustering of vacancies, is an undesirable isotropic volume expansion of materials under irradiation. This phenomenon occurs above stage-III recovery temperature where vacancies are mobile, between ~0.3–0.6 Tm where Tm is the absolute melting temperature. The primary mechanism for cavity swelling is “dislocation bias”, i.e., preferential absorption of interstitials by network dislocations, causing a vacancy supersaturation. Cavity swelling is highly temperature dependent and typically follows a bell-shaped curve with peak swelling occurring at an intermediate temperature. With increasing temperature, cavity size increases, and number density decreases logarithmically. With increasing dose, cavity swelling increases, but follows a hockey-stick type pattern, characterized by a low-swelling transient incubation period and then a steady-state regime. The incubation period, dominant in cavity nucleation, is dependent upon material’s chemistry and irradiation conditions such as dose rate/temperature. Cavity swelling increases linearly with dose in the steady-state regime, that is independent of chemistry and dose rate. This regime is dominated by cavity growth. Transmutation gases like He enhance cavity nucleation such that an early onset of cavity formation occurs. However, swelling magnitudes can be either increased or suppressed when He is present. Results in austenitic steels and Cu indicate that cavity swelling peaks at intermediate He/dpa ratios of ~10 appm He/dpa. The most effective method to mitigate cavity swelling is by increasing the fixed point-defect sink density, by promoting precipitation or dispersion strengthening.

Bhattacharya, Arunodaya↗
DOE OSTI2014

The role of nickel in radiation damage of ferritic alloys

According to modern theory, damage evolution under neutron irradiation depends on the fraction of self-interstitial atoms (SIAs) produced in the form of one-dimensional glissile clusters. These clusters, having a low interaction cross-section with other defects, are absorbed mainly by grain boundaries and dislocations, creating the so-called production bias. It is known empirically that the addition of certain alloying elements influences many radiation effects, including swelling; however, the mechanisms are unknown in many cases. In this study, we report the results of an extensive multi-technique atomistic level modeling study of SIA clusters mobility in body-centered cubic Fe–Ni alloys. We have found that Ni interacts strongly with the periphery of clusters, affecting their mobility. The total effect is defined by the number of Ni atoms interacting with the cluster at the same time and can be significant, even in low-Ni alloys. Thus a 1 nm (37SIAs) cluster is practically immobile at T < 500 K in the Fe–0.8 at.% Ni alloy. Increasing cluster size and Ni content enhances cluster immobilization. Finally, this effect should have quite broad consequences in void swelling, matrix damage accumulation and radiation induced hardening and the results obtained help to better understand and predict the effects of radiation in Fe–Ni ferritic alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Ultralow thermal conductivity of β -Cu 2 Se by atomic fluidity and structure distortion

We demonstrate a prototype thermal evolution path for liquid thermal conductivity in solids. Thermal evolution of β -Cu 2 Se shows large interstitial displacement of constituent atoms marked by glass-like transitions and an asymptotic liquid thermal transport. Using ab initio molecular dynamics (AIMD), we identify these transitions, and confirm them with in situ transmission electron microscopy and electron energy loss spectroscopy. The thermal disorder of the Cu+ ions forms homopolar Cu–Cu bonds under a rigid Se framework, and at yet higher temperatures the Se framework undergoes thermal distortion. In conclusion, the non-equilibrium AIMD prediction of lattice thermal conductivity shows significant suppression of the phonon transport, in agreement with experiments and molecular behavior.

36 MATERIALS SCIENCE↗
DOE OSTI2015

High temperature annealing of ion irradiated tungsten

In this study, transmission electron microscopy of high temperature annealing of pure tungsten irradiated by self-ions was conducted to elucidate microstructural and defect evolution in temperature ranges relevant to fusion reactor applications (500–1200°C). Bulk isochronal and isothermal annealing of ion irradiated pure tungsten (2 MeV W + ions, 500°C, 1014 W + /cm 2 ) with temperatures of 800, 950, 1100 and 1400°C, from 0.5 to 8 h, was followed by ex situ characterization of defect size, number density, Burgers vector and nature. Loops with diameters larger than 2–3 nm were considered for detailed analysis, among which all loops had View the MathML source and were predominantly of interstitial nature. In situ annealing experiments from 300 up to 1200°C were also carried out, including dynamic temperature ramp-ups. These confirmed an acceleration of loop loss above 900°C. At different temperatures within this range, dislocations exhibited behaviour such as initial isolated loop hopping followed by large-scale rearrangements into loop chains, coalescence and finally line–loop interactions and widespread absorption by free-surfaces at increasing temperatures. An activation energy for the annealing of dislocation length was obtained, finding E a =1.34±0.2 eV for the 700–1100°C range.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Characterisation of radiation damage in W and W-based alloys from 2MeV self-ion near-bulk implantations

The displacement damage induced in bulk W and W-5 wt.% Re and W-5 wt.% Ta alloys by 2 MeV W + irradiation to doses 3.3×10 17 - 2.5×10 19 W + /m 2 at temperatures ranging from 300 to750°C has been characterized by transmission electron microscopy. An automated sizing and counting approach based on Image J has been proposed and performed for all irradiation data. In all cases the damage comprised dislocation loops, mostly of interstitial type, with Burgers vectors b = ½<111> (> 60%) and b = <100>. The diameters of loops did not exceed 20 nm, with the majority being ≤ 6 nm. The loop number density varied between 10 22 and 10 23 loops/m 3 . With increasing irradiation temperature, the loop size distributions shifted towards larger sizes, and there was a substantial decrease in loop number densities. The damage microstructure was less sensitive to dose than to temperature. Under the same irradiation conditions, loop number densities in the alloys were higher than in pure W but loops were smaller. In grains with normals close to z = <001>, loop strings developed in W at temperatures ≥ 500°C and doses ≥ 1.2 dpa, but such strings were not observed in the W-Re or W-Ta alloys. However, in other grain orientations complex structures appeared in all materials and dense dislocation networks formed at higher doses.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Temperature dependence of the radiation tolerance of nanocrystalline pyrochlores A 2 Ti 2 O 7 (A = Gd, Ho and Lu)

A potentially enhanced radiation resistance of nanocrystalline materials, as a consequence of the high density of interfaces and surfaces, has attracted much attention both to understand the fundamental role of these defect sinks and to develop them for high-radiation environments. Here, irradiation response of nanocrystalline A 2 Ti 2 O 7 (A = Gd, Ho and Lu) pyrochlore powders with grain sizes of 20–30 nm was investigated by 1-MeV Kr 2+ ion bombardment. In situ transmission electron microscopy (TEM) revealed that the critical amorphization fluence for each nanocrystalline compound at room temperature was greater than that for their coarse-grained counterparts, indicating an enhanced amorphization resistance. The effect of temperature on the irradiation response of one of these compounds, nanocrystalline Lu 2 Ti 2 O 7 , was further examined by performing ion irradiation at an elevated temperature range of 480–600 K. The critical amorphization temperature (T c ) was found to be noticeably higher in nanocrystalline Lu 2 Ti 2 O 7 (610 K) than its coarse-grained counterpart (480 K), revealing that nanocrystalline Lu 2 Ti 2 O 7 is less resistant to amorphization compared to its coarse-grained phase under high temperatures. We interpret these results with the aid of atomistic simulations. Molecular statics calculations find that cation antisite defects are less energetically costly to form near surfaces than in the bulk, suggesting that the nanocrystalline form of these materials is generally less susceptible to amorphization than coarse-grained counterparts at low temperatures where defect kinetics are negligible. In contrast, at high temperatures, the annealing efficiency of antisite defects by cation interstitials is significantly reduced due to the sink properties of the surfaces in the nanocrystalline pyrochlore, which contributes to the observed higher amorphization temperature in the nano-grained phase than in coarse-grained counterpart. Altogether, these results provide new insight into the behavior of nanocrystalline materials under irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

A computational study of yttria-stabilized zirconia: II. Cation diffusion

Cubic yttria-stabilized zirconia is widely used in industrial electrochemical devices. While its fast oxygen ion diffusion is well understood, why cation diffusion is much slower—its activation energy (~5 eV) is 10 times that of anion diffusion—remains a mystery. Indeed, all previous computational studies predicted more than 5 eV is needed for forming a cation defect, and another 5 eV for moving one. In contrast, our ab initio calculations have correctly predicted the experimentally observed cation diffusivity. We found Schottky pairs are the dominant defects that provide cation vacancies, and their local environments and migrating path are dictated by packing preferences. As a cation exchanges position with a neighboring vacancy, it passes by an empty interstitial site and severely displaces two oxygen neighbors with shortened Zr-O distances. This causes a short-range repulsion against the migrating cation and a long-range disturbance of the surrounding, which explains why cation diffusion is relatively difficult. In comparison, cubic zirconia's migrating oxygen only minimally disturbs neighboring Zr, which explains why it is a fast oxygen conductor.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Preferential diffusion in concentrated solid solution alloys: NiFe, NiCo and NiCoCr

In single-phase concentrated solid-solution alloys (CSAs), including high entropy alloys (HEAs), remarkable mechanical properties are exhibited, as well as extraordinary corrosion and radiation resistance compared to pure metals and dilute alloys. But, the mechanisms responsible for these properties are unknown in many cases. In this work, we employ ab initio molecular dynamics based on density functional theory to study the diffusion of interstitial atoms in Ni and Ni-based face-centered cubic CSAs including NiFe, NiCo and NiCoCr. We model the defect trajectories over >100 ps and estimate tracer diffusion coefficients, correlation factors and activation energies. Furthermore, we found that the diffusion mass transport in CSAs is not only slower than that in pure components, i.e. sluggish diffusion, but also chemically non-homogeneous. The results obtained here can be used in understanding and predicting the atomic segregation and phase separation in CSAs under irradiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Radiation instability of equal channel angular extruded T91 at ultra-high damage levels

Although several studies have shown that nano-grained alloys have higher radiation resistance than coarse-grained variants, with dense grain boundary structures acting as defect sinks to remove damage, the present study shows the failure of severe deformation to suppress void swelling in T91 alloy at ultra-high damage levels. Fe self-ion irradiation up to 1000 peak dpa produced significantly more swelling in extruded T91 than conventional T91 at 475 °C. Dramatic radiation-enhanced grain growth was observed in the extruded T91, manifested in two rapid grain growth stages. One stage occurs at a relatively low damage level, and the other at a higher damage level beyond 750 peak dpa. The second rapid growth stage is correlated with void swelling and starts at the end of the swelling incubation period. The study suggests that the defect sink properties of the growing grain boundaries under a coarsening process are different from that of stable boundary configurations. When void swelling becomes significant and voids act as a biased sink for vacancies, the growing grain boundaries act as a biased sink for interstitials, leading to rapid growth at higher damage levels.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½< 111 > and <100 > Burgers vectors. The proportion of ½< 111 > dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signs of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. As a result, the largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modelling recrystallization textures driven by intragranular fluctuations implemented in the viscoplastic self-consistent formulation

This work presents a recrystallization model driven by intragranular orientation gradients and strain energy fields calculated by means of the viscoplastic self-consistent (VPSC) formulation. The VPSC model is extended for calculation of the coupling between intragranular stress fluctuations with corresponding second moments of lattice spin and misorientation fields in the grains. Access to these quantities allows modelling of transition bands and nucleation kinetics. In the proposed recrystallization model, grain growth is assumed to be proportional to the difference between the stored energy of each grain and that of the effective medium. Recrystallization textures for several cubic metals are simulated, showing good agreement with corresponding experiments. The model reveals the importance of considering appropriate, microstructurally-based and orientation-dependent recrystallization nucleation mechanisms. The recrystallization texture of heavily rolled copper with a strong cube texture component is found to be a consequence of nucleation at transition bands, which is also the cause of the recrystallization textures in compressed iron and drawn copper wire. Finally, in contrast, the recrystallization texture of rolled interstitial-free steel is found to be caused by grain boundary nucleation occurring in grains with the highest strain energy.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Unprecedented irradiation resistance of nanocrystalline tungsten with equiaxed nanocrystalline grains to dislocation loop accumulation

Nanocrystalline metals are often postulated as irradiation tolerant materials due to higher grain boundary densities. The efficiency of these materials in mitigating irradiation damage is still under investigation. Here, we present an in-situ transmission electron microscopy with ion irradiation study on equiaxed 35 nm grained tungsten (NCW-35 nm) and compare its radiation tolerance, in terms of dislocation loop damage, to several other grades of tungsten with different grain sizes at two temperatures (RT and 1073 K). The NCW-35 nm was shown to possess significant higher radiation tolerance in terms of loop damage. As demonstrated by Kinetic Monte Carlo simulations, at least part of the higher radiation tolerance of the small grains is due to higher interstitial storage (at the grain boundaries) and defect recombination (in the grain interiors) in the small grain material. In addition, experimental observations reveal rapid and efficient dislocation loop absorption by the grain boundaries and this is considered the dominant factor for mass transport to the boundaries during irradiation, enabling the remarkable radiation tolerance of 35 nm grained tungsten. Furthermore, this study demonstrates the possibility of attaining high radiation tolerant materials, in terms of dislocation loop damage, by minimizing grain sizes in the nanocrystalline regime.

36 MATERIALS SCIENCE↗