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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

The multiple roles of small-angle tilt grain boundaries in annihilating radiation damage in SiC

Lattice defects generated by radiation damage can diffuse to grain boundaries (GBs) and be annihilated at GBs. However, the precise role of GBs in annihilating the segregated defects remains unclear. Here, we employed multi-scale models to determine how interstitials are annihilated at small-angle tilt GBs (STGBs) in SiC. First of all, we found the pipe diffusion of interstitials in STGBs is slower than bulk diffusion. This is because the increased interatomic distance at dislocation cores raises the migration barrier of interstitial dumbbells. Furthermore, we found both the annihilation of interstitials at jogs and jog nucleation from clusters are diffusion-controlled and can occur under off-stoichiometric interstitial fluxes. Finally, a dislocation line model is developed to predict the role of STGBs in annihilating radiation damage. This model includes defect flux to GBs, pipe diffusion in STGBs, and the interaction of defects with jogs. The model predicts the role of STGBs in annihilating defects depends on the rate of defects segregation to and diffusion along STGBs. STGBs mainly serve as diffusion channel for defects to reach other sinks when defect diffusivity is high at boundaries. As a result, when defect diffusivity is low, most of the defects segregated to STGBs are annihilated by dislocation climb.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effects of Fe atoms on hardening of a nickel matrix: Nanoindentation experiments and atom-scale numerical modeling

Significant hardening effect due to Fe concentrations in Ni-based alloys with face-centered-cubic structure has been studied by using a combined experimental and atomistic-based computational approach via nanoindentation tests. The obtained experimental load–displacement data for the [0 0 1] crystal orientation reached a qualitative good agreement with molecular dynamics simulations results, leading to strong evidence that the main strengthening factors are associated to sluggish dislocation diffusion, reduced defect sizes and the nucleation of tetrahedral stacking faults. Here, interstitial type prismatic dislocation loops mainly formed by $\frac{1}{6}$$\langle$1 1 2$\rangle$ Shockley dislocations are nucleated during the loading process, where their interaction leads to the formation of pyramidal shaped stacking fault which are mainly created by $\frac{1}{3}$$\langle$1 0 0$\rangle$ Hirth dislocations lines. Observing both types of defects coexisting in the same plastic deformation zone by both approaches. Reported mechanical data, measured experimentally and interpreted numerically, are also in accordance with microstructural SEM and TEM investigations.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Advanced time integration algorithms for dislocation dynamics simulations of work hardening

Efficient time integration is a necessity for dislocation dynamics simulations of work hardening to achieve experimentally relevant strains. In this work, an efficient time integration scheme using a high order explicit method with time step subcycling and a newly-developed collision detection algorithm are evaluated. First, time integrator performance is examined for an annihilating Frank–Read source, showing the effects of dislocation line collision. The integrator with subcycling is found to significantly out-perform other integration schemes. The performance of the time integration and collision detection algorithms is then tested in a work hardening simulation. The new algorithms show a 100-fold speed-up relative to traditional schemes. As a result, subcycling is shown to improve efficiency significantly while maintaining an accurate solution, and the new collision algorithm allows an arbitrarily large time step size without missing collisions.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

In Situ TEM Microcompression Pillar Size Effects in Fe-9Cr ODS

The objective of this paper is to ascertain the existence of a specimen size threshold, below which the material no longer exhibits bulk mechanical behavior when tested using in situ transmission electron microscopic (TEM) micro-compression pillars. In this work, sub-micron sized pillars of a model Fe-9Cr oxide dispersion strengthened (ODS) alloy were compressed in TEM in situ at room temperature, and attempts have been made to quantify the resulting data. Results are compared to mechanical property values in the archival literature, and a discussion on the validity of quantitative in situ TEM micro-compression data for ODS steel is included. In situ TEM mechanical testing is a relatively new technique enabling concurrent TEM imaging and mechanical testing of sub-micron-sized electron-transparent specimens. Due to limited sample volume of many nuclear materials, in situ TEM mechanical testing holds great promise for advancing our understanding of deformation phenomena. But it is well known that the dramatically reduced specimen size required for in situ testing can affect quantitative and qualitative deformation behaviors. However, Kiener, et al. used in situ TEM micro-compression pillars on irradiated and unirradiated single-crystal copper. They observed size independence in the irradiated material, while the unnirradiated material displayed a size threshold below which the mechanical properties deviated from those of bulk Cu. They assert that the more densely-spaced obstacles in irradiated Cu allow it to display bulk properties even in specimens having ∼ 100 nm dimensions. Similar to Kiener's findings, we hypothesize that unirradiated ODS will demonstrate a size threshold, but the in situ TEM micro-compression technique can be extended from single crystal materials to engineering alloys. An Fe-9Cr model ODS alloy was chosen for this study because of the broad interest in ODS for advanced nuclear reactors. ODS alloys exhibit high irradiation tolerance and high temperature creep resistance. New insight into deformation processes gained through in situ TEM mechanical testing could be valuable for the validation of ODS alloys for next generation nuclear reactors. The fact that the compressive yield stress and elastic modulus do not follow distinct trends as a function of specimen size is likely due to the high density of obstacles to dislocation motion. The ODS microstructure contains such obstacles as oxide nanoparticles, a fine grain structure, a high density of dislocation lines, and carbide precipitates. Since the deformation behavior is governed by the minimum separation distance between obstacles [1], it is plausible that the high obstacle density enables ODS alloy micropillars to exhibit bulk-like mechanical properties even prior to irradiation. Future work will focus on: (1) developing finite element models to help explain the difference in elastic moduli between bulk and micropillar tests, (2) understanding the spread in yield stress measurements at fixed micropillar volumes, and (3) extending this study to irradiated ODS. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2018

Fundamental Studies of Irradiation-Induced Modifications in Microstructural Evolution and Mechanical Properties of Advanced Alloys

This final technical report summarizes the research performed during October 2014 and December 2017, with a focus on investigating the radiation-induced microstructural and mechanical property modifications in optimized advanced alloys for sodium-cooled fast reactor (SFR) structural applications. To accomplish these objectives, the radiation responses of several different advanced alloys, including austenitic steel Alloy 709 (A709) and 316H, and ferritic/ martensitic Fe–9Cr steels T91 and G92, were investigated using a combination of microstructure characterizations and nanoindentation measurements. Different types of irradiation, including ex situ bulk ion irradiation and in situ transmission electron microscopy (TEM) ion irradiation, were employed in this study. Radiation-induced dislocations, precipitates, and voids were characterized by TEM. Scanning transmission electron microscopy with energy dispersive X-ray spectroscopy (STEM-EDS) and/or atom probe tomography (APT) were used to study radiation-induced segregation and precipitation. Nanoindentation was used for hardness measurements to study irradiation hardening. Austenitic A709 and 316H was bulk-irradiated by 3.5 MeV Fe ++ ions to up to 150 peak dpa at 400, 500, and 600°. Compared to neutron-irradiated stainless steel (SS) 316, the Frank loop density of ion-irradiated A709 shows similar dose dependence at 400°, but very different temperature dependence. Due to the noticeable difference in the initial microstructure of A709 and 316H, no systematic comparison on the Frank loops in A709 vs 316H was made. It would be helpful that future ion irradiation study on 316 stainless steel could be conducted to directly compare the temperature dependence of Frank loop density in ion-irradiated 316 SS with that in neutron-irradiated 316 SS. In addition, future neutron irradiation on A709 at 400–600° at relative high dose (≥10 dpa) can be carried out to compare with ion-irradiated A709. The radiation-induced segregation (RIS) of Ni and Si was observed in both A709 and 316H in all irradiated conditions and was found at various sinks: line dislocations, dislocation loops, void surfaces, carbide-matrix interfaces, etc. Radiation also induced the formation of Ni,Si-rich precipitates. As suggested in a previous study on neutron-irradiated 316 stainless steel, one possible consequence of the significant RIS of Si is that the enrichment at defect sinks depletes the silicon in the matrix, which can lead to enhanced void nucleation rate. The enrichment of Ni and Si is accompanied by the depletion of Cr at defect sinks, which could also affect the corrosion resistance. Radiation-induced change in the orientation relationship of pre-existing MX precipitates was observed at 600°. It is believed that this change is associated with the network dislocations formed under irradiation. The underlying mechanism is still not well understood. This change could be a positive indication that the MX precipitates can survive high density network dislocations. It would be helpful if neutron irradiation at similar dose conditions could be carried out to verify that this effect is not unique for ion irradiation. Intragranular Cr-rich carbides with a core-shell structure, i.e. Cr-rich carbide core and Ni,Si-rich shell was found at 500° and 600° in the highest dose (150 peak dpa) specimens. Coarse voids (30 nm in diameter) were only commonly found at 500° in the 50 and 150 peak dpa specimens in regions less than 750 nm in depth. The highest swelling for A709 irradiated to 50 and 150 peak dpa at 500° is about 0.44% and 0.37%, respectively. Due to the choice of 100 degree temperature intervals, this study did not attempt to precisely identify peak void swelling conditions, merely the range of irradiation temperatures where this could be a concern. It is known high-dose ion irradiation can significantly suppress void nucleation. Future neutron irradiation in the 500–600° range (without considering the temperature shift) is needed to determine the onset of accelerated void swelling (possibly at lower dose).

36 MATERIALS SCIENCE↗
DOE OSTI2020

Structural damage and phase stability of Al 0.3 CoCrFeNi high entropy alloy under high temperature ion irradiation

In this study, an initially single phase high entropy alloy (HEA) Al 0.3 CoCrFeNi was irradiated by 3 MeV Au ions to a fluence of 6 × 10 15 cm –2 (~31 dpa at damage peak) at four different temperatures ranging from 250 °C to 650 °C. Transmission electron microscopy (TEM) and Atom probe tomography (APT) were employed to study the evolution of structural damage and phase stability with irradiation temperature. Al 0.3 CoCrFeNi exhibited a similar evolution of irradiation-induced defects with temperature as compared with conventional FCC alloys. At 250 °C and 350 °C, most of the visible irradiation-induced defects were faulted 1/3$\langle{111}\rangle$ dislocation loops. As the irradiation temperature increased to 500 °C, perfect 1/2$\langle{110}\rangle$ dislocation loops were observed along with the faulted loops. At the highest irradiation temperature 650 °C, only dislocation lines and networks could be observed. Regarding phase stability, the 3 MeV Au irradiation was observed to suppress the precipitation of (Ni, Al)-enriched nano clusters and the L12 ordered structure at irradiation temperatures 250 °C to 500 °C whereas precipitation of the B2 ordered structure was accelerated at 650 °C. This resulted in qualitatively opposite precipitation behavior between the ion irradiated damage region and unirradiated region at 500 °C and 650 °C. The opposite phase stability of the ion-irradiated damage region and unirradiated region at different temperatures is attributed to the competing effects of ballistic dissolution vs irradiation enhanced diffusion on precipitation.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Temperature and dose rate effects on dislocation loop formation and decoration in self-ion irradiated high-purity Fe-Cr model alloys

Body-centered cubic (BCC) Fe-based alloys are promising candidate materials for advanced nuclear reactors. A widespread perception is that the addition of Cr favors the formation of ½ <111> dislocation loops. However, the Cr effects on dislocation loop formation/evolution remain unclear due to the limited experimental studies. In this study, we irradiate ultra-high purity bulk BCC Fe-xCr (x = 3, 5, and 8 wt%) with 6.7-8 MeV Fe ions over a wide range of temperatures (Tirr = 250-500 °C) and dose rates (10⁻⁵ to 10⁻³ dpa/s) to two midrange doses (0.35 and 3.5 dpa). The major irradiation-induced microstructures are loop decoration along non-screw dislocation lines at low temperatures and high dose rates as well as self-assembled loops at elevated temperatures. Post-irradiation Burgers vector analysis via g•b method on loops and electron tomography reveals that <001> preponderance is observable in all irradiated Fe-Cr samples at elevated temperature, resulting from stronger elastic energy originating from <001> edge.

Li, Yao [University of Tennessee (UT)]↗
DOE OSTI2019

Combined effect of injected interstitials and He implantation, and cavities inside dislocation loops in high purity Fe-Cr alloys

High purity Fe-Cr alloys with Cr content ranging from 3 to 14 wt% were irradiated by self-ions at 500 °C in dual-beam mode up to 157 displacements per atom (dpa), 17 appm He/dpa. Transmission electron microscopy (TEM) on focused ion beam foils revealed the depth distribution of irradiation induced cavities and dislocation loops. A detailed quantitative analysis of cavity microstructure was performed on Fe14%Cr, while Fe(3, 5, 10, 12)%Cr were analyzed qualitatively. A homogeneous distribution of small cavities were observed up to the damage peak. From surface to the damage peak, cavity number density increased almost linearly. The cavity microstructure changed drastically at and after the damage peak, where the size increased and number density decreased. Most notably, the microstructure consisted of a striking mixture of heterogeneously nucleated cavities inside dislocation loops and freely nucleated cavities in the matrix. Further, the depth-dependent void swelling increased continuously along the damage depth. This contrasts with most ion irradiation results where suppression of void swelling occurs adjacent to the damage peak due the injected interstitial effect. We hypothesize a plausible mechanism of the observed swelling variation based on a combined effect of the injected interstitials and helium implantation near the damage peak by comparing the results with those in pure iron irradiated under same conditions. This synergistic effect may develop cavity microstructures which do not necessarily reflect microstructures expected due to injected interstitial effect and can be easily wrongly interpreted. The depth dependent dislocation loop microstructure was studied qualitatively in Fe14%Cr sample. Up to the damage peak, a complex dense network of dislocations formed due to the interaction/impingement of the dislocation loops. Individually resolvable loops were only observed after the damage peak, where their association with cavities was noted. Furthermore helium also induced heterogeneous cavity nucleation on dislocations constituting a grain boundary and on pre-existing dislocation lines.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Microstructural changes of proton irradiated Hastelloy-N and in situ micropillar compression testing of one single grain at different local damage levels

Here, in situ micropillar compression was used to study the deformation of proton-irradiated Hastelloy-N at different damage levels. Multiple pillars were prepared from a single domain along the cross-section of 2.5 MeV proton-irradiated Hastelloy-N. Depending on the location of micropillars, the critical resolved shear stress was obtained as a function of local damage levels. Such an approach eliminates the variation of yield stress due to the difference in the Schmid factor. Microstructural characterization showed that irradiation hardening is contributed by complicated defect structures, including (a) dislocation loops with many in corduroy-like alignments, (2) dislocations pile up, and (3) Cottrell atmosphere. Silicon atoms are found to segregate at dislocation lines and loops and form complicated Cottrell cloud patterns. The critical resolved shear stress, $t_{crss}$, and hardening exponents were both extracted as a function of displacements per atom values up to 2.3. There was a 60% increase in $t_{crss}$ at the highest damage level.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Microstructural evolution of compositionally complex solid-solution alloys under in-situ dual-beam irradiation

Here, this work attempts to link the microstructural evolution of single-phase compositionally complex alloy (CCA) compositions under dual-beam irradiation to their Mn-content via the stacking-fault energy (SFE) and vacancy migration energies. Two alloys, Cr 18 Fe 27 Mn 27 Ni 28 and Cr 15 Fe 35 Mn 15 Ni 35 , along with less compositionally complex pure Ni and Fe 56 Ni 44 binary were irradiated at 500 and 600 °C under dual-beam 1 MeV Kr 2+ and 16 keV He + heavy-ions up to 7 displacements per atom (dpa) with a He/dpa ratio of 0.75 %/dpa using in situ transmission electron microscopy (TEM). Due to the bubble-stabilizing effect of implanted He, bubbles were observed in all irradiations, and populations of faulted interstitial loops were characterized in Cr 18 Fe 27 Mn 27 Ni 28 and Cr 15 Fe 35 Mn 15 Ni 35 . A reduction in swelling was observed in the two CCAs compared to pure Ni and Fe 56 Ni 44 . Although swelling increased from 500 to 600 °C in Fe 56 Ni 44 , Cr 15 Fe 35 Mn 15 Ni 35 and Cr 18 Fe 27 Mn 27 Ni 28 both swelled slightly more at 500 °C. This was attributed to the difference in vacancy mobility, stronger pinning effect of vacancies on He, and the sink strength of faulted dislocation loops. Faulted interstitial loops nucleated with a higher number density and dislocation line density in Cr 15 Fe 35 Mn 15 Ni 35 at both temperatures, and at 600 °C in both materials. The differences in faulted loop population and temperature effect on swelling are correlated to the Mn-content and the measured SFE (20.2 ± 6.7 mJ/m 2 for Cr 18 Fe 27 Mn 27 Ni 28 and 9.2 ± 3.4 mJ/m 2 for Cr 15 Fe 35 Mn 15 Ni 35 ).

36 MATERIALS SCIENCE↗
DOE OSTI2026

In situ TEM annealing of neutron-irradiated Ti reveals a two-stage mechanism for elevated temperature radiation damage recovery

Understanding how irradiation-induced defects evolve at elevated temperatures is of critical importance to predicting materials' behavior under steady-state and accident scenarios. However, such mechanistic insight into microstructural evolution is limited by the nature of ex situ annealing and subsequent imaging. Here we show direct observation and quantification of defect recovery in neutron-irradiated Ti using in situ transmission electron microscopy (TEM) annealing experiments. In agreement with our prior work, and at temperatures below the irradiation temperature (T irr = 300 °C), dislocation loops are observed to glide. At elevated temperatures (>500 °C), dislocation lines become mobile and promote significant recovery of the microstructure. These mechanisms challenge the established electron irradiation-based model for radiation damage recovery, which originally suggests dissolution of static defect clusters, and demonstrates the importance of in situ characterization in understanding defect evolution in irradiated materials.

Hirst, Charles A. [Univ. of Michigan, Ann Arbor, M↗
DOE OSTI2025

The influence of V-defects, leakage, and random alloy fluctuations on the carrier transport in red InGaN MQW LEDs

Red InGaN-based light-emitting diodes (LEDs) exhibit lower internal quantum efficiencies than violet, blue, and green InGaN LEDs due to a reduction in radiative recombination rates relative to non-radiative recombination rates as the Indium composition increases. Additionally, the larger polarization and band offset barriers between high indium content InGaN quantum wells and GaN quantum barriers increase the forward voltage. In blue and green LEDs, random alloy fluctuations and V-defects play a key role in reducing the forward voltage. When V-defects are present, either naturally or intentionally introduced, they create an alternative path for carrier injection into the MQWs through the V-defect sidewalls. This injection mechanism explains the turn-on voltages of green LEDs. However, in InGaN red LEDs, these two phenomena do not reduce the forward voltage as effectively as in blue and green LEDs, and consequently, the computed forward voltage remains significantly higher than the measured one. Furthermore, currents are observed at low voltages before the turn-on voltage (V<ℏω/e=2.0 V) of red LEDs. To address this, we introduce dislocation-induced tail states in the modeling, suggesting that leakage current through these states may play a significant role both below and at turn-on voltages. The simulation also indicates that leakage carriers below turn-on accumulate, partially diffuse in the QWs, screen the polarization-induced barrier in the low injection regime, and further reduce the forward voltage. Despite these beneficial effects, a drawback of dislocation-induced tail states is the enhanced nonradiative recombination in the dislocation line region. This study provides a detailed analysis of device injection physics in InGaN QW red LEDs and outlines potential optimization strategies.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2026

TEM Approaches for Microstructure-Informed Prediction of Mechanical Properties in Structural Alloys

Predicting the mechanical performance of structural alloys from their evolving microstructure remains a major challenge in materials science, particularly for nuclear structural materials, where irradiation-induced defects span multiple types and length scales and interact through complex mechanisms. The dispersed barrier hardening (DBH) [1] and Friedel–Kroupa–Hirsch (FKH) [2,3] models have been widely used to evaluate the hardening contributions of individual obstacles and to estimate tensile strength from quantified microstructures; however, when multiple size-dependent obstacles coexist and evolve, predicting temperature-dependent tensile strength becomes significantly more complex, and a fully consistent hardening model is still lacking. Transmission electron microscopy (TEM) plays a central role in refining hardening models and enabling predictive assessments of tensile strength evolution by providing quantitative characterization of dislocations, irradiation-induced defects (e.g., dislocation loops and cavities), precipitates, and grain structure (Fig. 1.). These experimentally measured defect densities are incorporated into physically based hardening models with size- and shape- dependent obstacle strengths [4], using root-sum-square superposition for obstacles of comparable strength and linear superposition for dissimilar ones [5]. In addition, recent advances in TEM [6-8], including high-resolution imaging, 4D-STEM strain mapping, EDS/EELS elemental analysis, and flash-polishing-based TEM specimen preparation and extraction-replica methods (Fig. 2), further improve the accuracy of microstructural quantification. By comparison with prior studies as well as our own results, we show that when TEM-derived microstructural information is carefully integrated with physically grounded hardening models, yield strength (or irradiation-induced hardening) measured at room temperature can be predicted with good quantitative agreement across multiple alloy classes. In-situ TEM combined with high-temperature mechanical testing represents an important next step for refining hardening models by directly probing dislocation–obstacle interactions across varying irradiation doses and temperatures [9]. Because the barrier strength factor (α) depends on both temperature and obstacle size, it should not be treated as a constant fitting parameter; rather, it must be explicitly evaluated to achieve physically meaningful predictions of mechanical behaviour at operating temperatures. This presentation therefore discusses why all strengthening contributions (e.g., Peierls stress, solid-solution strengthening, voids, bubbles, dislocation loops, dislocation lines, and grain boundaries) must be considered collectively, why appropriate superposition methods are essential when obstacles possess different barrier strength factors, how hardness measurements can be meaningfully related to tensile properties, and how TEM-derived microstructural information can be systematically incorporated into hardening models. More broadly, it outlines a pathway toward microstructure-informed prediction of mechanical properties and supports the goal of establishing science-based tools for evaluating structural materials in extreme environments [10].

Lin, Yan-Ru [ORNL] (ORCID:0000000339991473)↗
DOE OSTI2019

DefectSegNet

DefectSegNet for semantic segmentation of defects in Steels The first and most important step of automating defect analysis is perceptual defect identification. In terms of digital image processing, semantic segmentation best emulates human recognition of defect features – it identifies the crystallographic defects and where they are located in a TEM micrograph. We solved this problem by developing DefectSegNet, a novel hybrid CNN algorithm for robust and automated semantic segmentation of three crystallographic defects, including line dislocations, precipitates and voids, that are commonly observed in structural metals and alloys.

Sainju, Rajat↗
DOE OSTI2024

Post Irradiation Examination Dislocation Defect Detection Software

This software provides dislocation-type defect identification and segmentation using a standard open source computer vision model, YOLOv8, that leverages transfer learning to create a highly effective dislocation defect quantification tool while using only a minimal number of expert annotated micrographs for training. This model demonstrates the ability to segment both dislocation lines and loops concurrently in micrographs with high pixel noise levels and on multiple alloys. It includes multiple layers of frozen layers used for transfer learning from multidisciplinary data and is extensible to alloys that are not included in the training dataset.

Anderson, MatthewW↗
DOE OSTI2017

Deformation Behavior in Neutron Irradiated Generation II FeCrAl Alloys for Accident Tolerant Fuel Cladding

Preliminary work has been completed to evaluate the deformation behavior of the Generation II FeCrAl alloy class. The Generation II alloy, C35M, was neutron irradiated within the High Flux Isotop e Reactor followed by post-irradiation examination to evaluate the mechanical properties, irradiated microstructure, deformation modes, and deformation microstructure. Complementary characterization techniques were used in this evaluation. It was found tha t aft er irradiation to1.8 dpa at 364 °C that the C35M specimen failed in a ductile manner. The observed deformation microstructure showed a high density of line dislocations, dislocation networks, and dislocation pile-ups. The observed performance and stru cture s are consistent with literature studies on high-Cr FeCr alloys that were thermally aged at 500°C. The observed results suggest that Generation II FeCrAl alloys retain ductility under low-dose, prototypical LWR irradiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Second Annual Progress Report on Correlation Between Microstructure and Mechanical Properties of Neutron-Irradiated Ferritic-Martensitic and Austenitic Steels

Ferritic-martensitic steels G92-2b (an optimized Grade 92 heat), NF616 and T91, and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) and the Advanced Test Reactor (ATR) of Idaho National Laboratory (INL). Selected G92-2b samples were irradiated up to 14.66 dpa in the HFIR at two temperature ranges: 400–496.7°C and 683.3– 720°C. NF616 and T91 were irradiated in the ATR up to 8.16 dpa with the irradiation temperatures ranged from 241°C to 447.5°C. Alloy 800H and 800H-TMP samples were irradiated in both the HFIR and the ATR. Selected 800H and 800H-TMP samples had HFIR irradiation to 1.28 dpa at 580°C and ATR irradiation up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurements, fractography, and microstructural characterization were performed on the selected samples in the Low Activation Materials Design and Analysis (LAMDA) laboratory. Radiation-hardening of G92-2b was observed at the lower doses and lower irradiation temperatures (400- 496.7°C), with GB03 (0.52 dpa at 400°C) and GB04 (7.44 dpa at ~490°C) showing ~12% and ~8% hardening, respectively. Softening by ~14% was observed for GB05 (14.66 dpa at 496.7°C). Radiationsoftening of G92-2b was more prevalent at the higher irradiation temperatures (683.3-~720°C), with GB10 (0.46 dpa at 683.3°C), GB11 (7.44 dpa at ~720°C), and GB12 (14.63 dpa at ~720°C) showing ~8%, ~8%, and ~40% softening, respectively. Radiation-hardening of NF616 and T91 was observed with the hardness increased by ~37% to ~65% depending on the irradiation doses and irradiation temperatures. Within the studied irradiation conditions of NF616 and T91, samples with a higher dose had a larger hardness after irradiation. All the tested alloy 800H and 800H-TMP samples in this work showed radiation-hardening by ~96±7% to ~152±10%. Alloy 800H-TMP tended to have slightly smaller radiation-hardening than alloy 800H. The fractography results of G92-2b sample GB03, GB10, and GB11, together with the previously characterized fractography of GB04, GB05, and GB12, indicated that the ductility of G92-2b was maintained up to 14.66 dpa at the lower irradiation temperatures of 400-496.7°C, while some loss of ductility (less necking) was observed for higher doses at the higher irradiation temperatures of 683.3-720°C. This agrees with the previously reported tensile test results of G92-2b, where the elongation of G92-2b was reduced at higher doses at the higher irradiation temperatures. Dimple sizes increased at higher doses, which are more evident at the higher irradiation temperatures of 683.3-~720°C. The fractography of NF616 sample D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for sample D2, while ductile failure for samples D4 and D6. Fractography of alloy 800H and 800H-TMP samples in various irradiation conditions showed ductile failure with obvious necking. Dimples were observed, with some of them containing large Ti-rich particles, in all the 800H/800H-TMP samples. Electron backscatter diffraction characterization of GB12 indicated the recovery of the lath structure, which was generally replaced by an equiaxed grain structure. Transmission electron microscopy (TEM) characterization showed the presence of frequent M 23 C 6 (M = primarily Cr), MX (M = primarily V), spherical Nb(C,N) precipitates, and occasional Laves phase precipitates in the G92-2b samples. MX precipitates with sizes of 20-30 nm were observed at boundaries of smaller grains, indicating the pinning effect of the V-rich precipitates. The lath structure recovery was more evident at the higher irradiation temperatures (683.3-~720°C), with decreased densities of line dislocations and M23C6 precipitates. The irradiated T91 (TA04) showed the growth of M23C6 precipitates to 101 ± 40 nm from the initial 68 ± 22 nm in the unirradiated condition. Dislocation loops of both {100} and {111} types were present in TA04. TEM characterization was also performed on the irradiated 800H (N4, N5, N6, and AR2) and 800H-TMP (P4, P5, P6, and HG1). Accumulation of M 23 C 6 precipitates at grain boundaries was observed in all the xii 800H/800H-TMP samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in the irradiated 800H-TMP. Some Ti(C,N) precipitates are embedded in the M 23 C 6 precipitates, maintaining specific orientation relationships between the precipitates and between the precipitate and matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all the 800H/800H-TMP samples, with EDS Si maps tending to have a lower contrast in 800H-TMP samples. Atom probe tomography was conducted on the same samples, supported by a Rapid Turnaround Examination project under Nuclear Science User Facilities. The results are being analyzed to be integrated with the TEM results for a confident description of the γ’ precipitates. Dislocation loops also formed in all the 800H/800H-TMP samples. The density and the average size of dislocation loops were quantified to be in the order of 10 22 – 10 23 m -3 and 11.7 – 15.9 nm, respectively, in the ATR-irradiated 800H/800H-TMP samples. The loop densities in irradiated 800H were higher than that in irradiated 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.

36 MATERIALS SCIENCE↗