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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Scanning Transmission Electron Microscopy–Atom Probe Tomography Correlative Analysis for the Characterization of Solute-defect Interactions

Atom probe tomography (APT) and (scanning) transmission electron microscopy ((S)TEM) are complementary techniques that provide spatially resolved chemical and structural information at the atomic scale. Here, in this study, we employ two different STEM/APT correlative analysis methods to investigate Cr segregation at dislocation loops in ultra-high purity Fe–Cr alloys. APT needles for the correlative analysis were extracted either from bulk material or from thinned TEM lamellae. STEM analysis was used to determine the Burgers vectors of ion-irradiation-induced dislocation loops, while APT reconstruction of the same region revealed the Cr segregation to these loops. We extended the g•b = 0 invisibility criterion of dislocation loops from TEM mode in a lamella to STEM mode in a needle-shaped specimen. STEM and APT analysis on the same needle provide straightforward correlative analysis, although it is limited by a small observation volume. In contrast, iterative STEM analysis of TEM lamellae, followed by the selective extraction of specific regions of interest for APT analysis, expands the observation area by up to 100 times but requires additional time-consuming steps for APT needle extraction from the lamellae.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Dynamic observation of dual-beam irradiated Fe and Fe-10Cr alloys at 435 °C

The attractive mechanical properties and superior resistance to void-swelling make ferritic/martensitic alloys a promising structural material for advanced nuclear reactors. However, one anomaly that has intrigued researchers for more than 50 years is the proportion of two types of dislocation loops in Fe and Fe-Cr alloys with Burger vectors b=½<111> and b=<100>. Although the possible mechanisms responsible for the presence of <100> loops continue to be the subject of intense modeling studies, there remains incomplete experimental understanding of fundamental irradiation processes in Fe(Cr) alloys. Here, the dose dependence of the irradiation-induced microstructural evolution was examined from 0 to 20 displacement per atom (dpa) in high purity Fe and Fe-10Cr during simultaneous dual-beam (1 MeV Kr + 10 appm He/dpa) irradiation at 435 °C. We experimentally revealed that the mechanism for the formation of <100> loops may not follow the conventional simple dislocation reaction between two 1½<111> loops. Real-time dynamic formation and evolution of defects including black dot loops, loop coarsening, loop decoration, network dislocations, and cavities were demonstrated. Several results indicated that the addition of Cr and He could impede dislocation loop motion. The evolution of the defect size/density and relative fraction of ½<111> vs <100> loops were quantitatively summarized. With increasing dose, ½<111> loops became the dominant type of loop in both materials. Notably, <100> loops were predominantly observed near grain boundaries only for pure Fe, while arrays of nanoscale black dot defects composing the <100> loop strings were observed in plenty in Fe-10Cr.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Evolution of extended defects in UO 2 during high temperature annealing

The evolution of thermophysical characteristics of uranium oxide (UO 2 ) under irradiation is a critical aspect of nuclear fuel performance of light water reactors. This study examines the effect of high-temperature annealing at 1000 °C - 1600 °C on defect evolution in krypton irradiated UO 2 at room temperature. Transmission electron microscopy analysis reveals that the size of irradiation-induced bubbles and dislocation loops increases but their density decreases with annealing temperatures. Perfect dislocation loops with = <110> dominate, regardless of the annealing temperature. A rate theory model for extended defect evolution is applied to conclude that current understanding of dynamical loop and bubble evolution at high-temperature annealing is governed by defect coalescence.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Temperature and radiation effects on brittle versus ductile fracture behavior in miscible phase boundaries: insight from atomistic simulations

Temperature- and irradiation-assisted failure mechanisms in miscible phase boundaries are clarified via atomistic calculations. We first establish the temperature-dependent brittle-to-ductile transition in U–Zr miscible phase boundaries. Our results confirm that these boundaries are mostly brittle at low temperatures and ductile at elevated temperatures. We then investigate the changes induced by irradiation on the fracture mechanisms in such phase boundaries. The irradiation-induced defect accumulation follows a multi-stage process that starts with the accumulation of isolated small dislocation loops before transitioning to the saturation and growth of larger dislocation loops and end up with a reorganization into forest dislocations. The accumulation of loops is the primary feature to participate in the delineation between brittle and ductile interfacial fracture in irradiated phase boundaries. At low damage levels, radiation defect interactions with the crack tip are limited and U–Zr miscible boundaries fail through the emission of dislocations ahead of the crack tip followed by brittle cleavage in agreement with the classical Griffith’s criterion for crack stability. At higher damage levels, the failure mode transitions from brittle crack growth to ductile void growth. In this case, the microcrack tip is blunted by the high density of pre-existing, radiation-induced defects in the vicinity of the crack. This interaction leads to the development and growth of a cavity at the interface as opposed to interfacial cleavage.

36 MATERIALS SCIENCE↗
DOE OSTI2020

(S)TEM/EDS study of native precipitates and irradiation induced Nb-rich platelets in high-burnup M5®

We have investigated microstructure and microchemistry of precipitates and dislocation loops in high-burnup M5® using (scanning) transmission electron microscopy ((S)TEM) equipped with energy dispersive X-ray spectroscopy (EDS). Two (S)TEM lamellae were made by cryo-FIB from the same cladding sample. The Nb-rich native precipitates were found in the metal, in the suboxide and in the oxide. Upon diffraction analysis, most of the Nb-rich native precipitates in the metal matrix remain as β-Nb phase, while no β-Nb native precipitates were found in the oxide. Nearby the oxide and metal (O/M) interface, the native precipitates in the oxide were already oxidized into t-NbO2 phase. At further distance away from the O/M interface, the Nb-rich native precipitates were gradually oxidized and became amorphous. Besides the native precipitates, Nb-rich irradiation induced precipitates (IIPs) were found in the metal matrix. Using g = <0002> vector for imaging, the length of the IIPs was aligned with dislocation loops. However, no Nb segregation to the dislocation loops themselves was observed. For the first time, we report that IIPs indeed exist in the oxide but only within about 1.5 µm away from O/M interface. However, the oxidation state of the IIPs in the oxide is still unclear. The presence of both native precipitates and IIPs in the oxide may indicate the Nb concentration in the oxide solid solution remain low nearby the O/M interface, which may explain the reduced corrosion kinetics of in-pile M5®. On the other hand, no IIPs were observed in the oxide at further distance and this may indicate that they have eventually dissolved back into the oxide. A few mechanisms related to IIPs stability in the oxide are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Temperature-dependent mechanism on hardening mitigation in thermomechanically processed 800H alloys under neutron irradiation

Irradiation induced hardening is commonly observed in structural materials, and is often associated with the degraded ductility. This study explores the role of thermomechanical processing (TMP) in mitigating irradiation-induced hardening in Incoloy 800H. The radiation response of solution-annealed and TMP-treated samples was systematically investigated to elucidate the underlying mechanisms of hardening reduction. Neutron irradiation was performed on solution-annealed and TMP-treated samples at 359, 431, and 580 °C, revealing consistently lower hardening in TMP samples. Microstructural characterization compared dislocation loops, cavities, clusters, and precipitates to assess their contributions to hardening. The strengthening contributions from individual microstructural features were estimated, and the microstructure-derived yield strength is consistent with the hardness-converted yield strength. The microstructure–properties correlation revealed the temperature-dependent primary hardening mechanism: dislocation loops dominated at lower temperature, whereas Ni3(Al,Ti) γ′ precipitates were predominant at higher temperature. TMP reduced radiation-induced hardening through two mechanisms: (1) increased sink strength from dislocations and Ti(C,N) precipitates limited dislocation loop formation at lower temperatures, and (2) reduced Ti solute within the matrix in TMP samples suppressed γ′ precipitate formation at higher temperatures. These findings identify the mechanisms governing radiation-induced hardening across different temperatures, which lays the foundation for TMP optimization for the development of advanced radiation-tolerant materials.

Zhong, Weicheng [ORNL] (ORCID:0000000231589271)↗
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↗
DOE OSTI2024

Strengthening mechanisms for microstructures containing unimodal and bimodal γ' precipitates in ATI 718Plus

Here, the influence of γ' precipitate size distribution on the deformation mechanisms under tensile loading in ATI 718Plus was studied. A set of aging treatments within the temperature range of 720 °C–900 °C was performed on solution-treated samples to obtain various γ' precipitate size distributions. Unimodal and bimodal γ' precipitate size distributions were achieved through single-step and two-step aging sequences, respectively, and such microstructures were tensile tested to failure to assess their yield strength, ultimate tensile strength, and elongation-to-failure. Some of the tensile samples were interrupted after achieving 3–4 % plastic strain, and the deformed microstructures were examined using transmission electron microscopy to investigate the γ' precipitate-dislocation interactions. For the unimodal γ' precipitate size distribution samples with the smaller γ' precipitates (radius ~ 7 nm), dislocations sheared through the precipitates. Both dislocation loops and paired dislocations were observed for the microstructures containing larger γ' precipitates (radius ~ 24 nm). The microstructure containing a bimodal γ' precipitate size distribution, which included average γ' precipitate radii of ~6 nm and ~28 nm, exhibited shearing as the dominant deformation mechanism, and this microstructure exhibited the highest strength values. The experimental observations were rationalized based on the theoretically-calculated critical resolved shear stress values for shearing and looping and a modified model for predicting the yield strength for bimodal microstructures was introduced.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructural evolution of a silicon carbide-carbon coated nanostructured ferritic alloy composite during in-situ Kr ion irradiation at 300°C 450°C

This work focuses on irradiation behaviors of a novel silicon carbide and carbon coated nanostructured ferritic alloy (SiC-C@NFA) composite for potential applications as a cladding and structural material for next generation nuclear reactors. The SiC-C@NFA samples were irradiated with 1 MeV Kr ions up to 10 dpa at 300 and 450 °C. Microstructures and defect evolution were studied in-situ at the IVEM-Tandem facility at Argonne National Laboratory. The effects of ion irradiation on various phases such as α-ferrite matrix, (Fe,Cr) 7 C 3 , and (Ti,W)C precipitates were evaluated. The α-ferrite matrix showed a continuous increase in dislocation density along with spatial ordering of dislocation loops (or loop strings) at >5 dpa. The size of the dislocation loops at 450 °C was larger than that at 300 °C. The nucleation and growth of new (Ti,W)C precipitates in α-ferrite grains were enhanced with the ion dose at 450 °C. This study provides new insight into the irradiation resistance of the SiC-C@NFA system.

(FeCr)(7)C-3↗
DOE OSTI2025

In-situ irradiation of uranium carbide

Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe + and 1 MeV Kr 2+ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO 2 or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO 2 and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

In situ study on radiation response of tungsten manufactured by laser powder bed fusion

Tungsten (W) produced by laser powder bed fusion (LPBF) was examined by in situ Krypton (Kr) ion irradiation at 400 °C up to 2.52 displacements per atom (dpa) to investigate its radiation response. Dislocation loops with identical Burgers vectors form aligned raft structures, inducing significant grain misorientation accumulation. Defect saturation was observed beyond 0.36 dpa, marked by constant loop density and raft spacing. WO 3 nanoparticles are found in the as-printed matrix and served as efficient defect sinks. Dislocation loops were absorbed at the W/WO 3 interface, facilitating defect annihilation and suppressing defect accumulation. In conclusion, these findings highlight the role of LPBF microstructure and oxide interfaces in mediating radiation-induced defect evolution, offering insights for designing radiation tolerant W-based materials.

Defect sink↗
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 OSTI2025

Characterization of in-situ and ex-situ ion-irradiated additively manufactured 316L and 316H stainless steels

Additively manufactured (AM) 316 stainless steel (SS) differs from its wrought counterpart in its unique dislocation cell structure and the presence of segregation and oxide particles at the cell walls. This work investigated the evolution of the microstructure in laser powder bed fusion (LPBF) 316L and 316H SS under in-situ 1 MeV Kr ion irradiation at 600 °C to 5 dpa, and ex-situ 4 MeV Ni ion irradiation at 300 °C and 600 °C from 0.2 dpa to 10 dpa, with a dose rate for all experiments of 10 -3 dpa/s. The results reveal that the dislocation cell structure results in heterogeneous formation of dislocation loops and voids, particularly at 600 °C, where loops tend to form within the cell interiors while voids form at the cell boundaries. LPBF 316H has a reduced level of swelling compared to LPBF 316L due to prolonged incubation. Energy Dispersive X-ray Spectroscopy (EDS) mapping indicates Ni and Si segregation at void surfaces due to radiation-induced segregation. At 300 °C, where voids are absent, the distribution of dislocation loops and stacking fault tetrahedra appears to be uniform. Dislocation cell structures mostly disappeared by 2 dpa for all conditions in this work. M 23 C 6 carbides were observed in LPBF 316H at 600 °C as early as 0.2 dpa, but not in LPBF 316L. Nanoindentation was performed to obtain the hardness of irradiated materials. In conclusion, this work illustrated the influence of additive manufacturing processes on microstructure evolution under irradiation, revealing the differences as well as the similarities as compared with wrought 316 SS, and the AM-related phenomenon that can potentially occur under neutron irradiation in nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Neutron irradiation induced defects and clustering in NF616 and T91

NF616 is a third-generation ferritic martensitic steel, developed to have better creep resistance than the prior generation T91. Unlike relatively numerous studies of T91, there is a lack of understanding of the irradiation effect on the microstructural evolution and mechanical response of NF616. This work evaluated the microstructures and radiation hardening of NF616 irradiated up to 8.2 displacement per atom (dpa) at 292 °C – 431 °C, compared with T91 from two heats. Dislocation loops were observed in all investigated samples. NF616 exhibited comparable loop size but slightly lower loop density than those in the general T91 heat at 430 °C. Cavities were only observed in NF616 at 431 °C but absent at lower irradiation temperatures (292 °C and 359 °C). Ni-rich clusters were also observed in NF616 at 431 °C, while only weak Ni-clustering were observed at lower irradiation temperatures. Compared to the general T91 heat, NF616 demonstrated better swelling resistance (e.g., one third of swelling in the general T91 heat at 430 °C), a slightly higher number density of Ni-rich clusters, and slightly lower radiation hardening. The low-carbon T91 showed the greatest hardening with the largest swelling and loop sizes, despite its lowest irradiation temperature and intermediate dose. The calculated hardening from loops, cavities and Ni-rich clusters using the classic dispersed barrier-hardening model had reasonable agreement with the experiment-derived results, with the primary hardening contribution attributed to dislocation loops.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Defect Production and Microstructural Feature Impact for Radiation Damage in Additively Manufactured 316 Stainless Steel

This milestone presents multi-scale modeling research results for additively manufactured 316 stainless steel. A combination of phase field, cluster dynamics, molecular dynamics, and density functional theory with machine learning is used, allowing for predictions of radiation-driven microstructural evolution in additively manufactured 316 stainless steel over a range of temperatures, damage rates, neutron spectra, and microstructures, and supporting the development of combined ion and neutron irradiation for material qualification. Informed by ion irradiation and neutron irradiation results across the Advanced Materials and Manufacturing Technologies program, we investigate the unique aspects of radiation-driven microstructure evolution in additively manufactured 316 stainless steel. In particular, we focus on understanding the impact of carbon concentration (varying, for example, between the 316L and 316H standards) on void formation; radiation-induced segregation at dislocation cells and grain boundaries; and the evolution of dislocation loops and network dislocation populations. We find that the unique characteristics of the additively manufactured microstructures must be accounted for in understanding the evolution of dislocation populations under thermal and irradiation conditions, such as the variation in sink strengths arising due to the variation in dislocation density. We also find that the radiation-induced segregation of Cr and Ni to grain boundaries and cell walls differs due to the differences in their defect sink biases. We also find that increasing the Ni content can slow vacancy diffusion, which may provide a mechanism for the observed reduction in transient swelling rate for austenitic Fe-Cr-Ni alloys with increasing Ni content. In addition, ion irradiations have shown that increasing carbon content in 316 SS results in a larger population of smaller voids, suggesting reduced vacancy diffusion. Our results show that the carbon content of additively manufactured 316 SS has a significant impact on the migration rate of defect clusters. The presence of carbon atoms results in carbon-vacancy trapping, significantly reducing the diffusion rate of vacancies. Carbon atoms may also be trapped near the surface of a void, which may reduce void growth by trapping vacancies that diffuse toward the void.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Cluster dynamics modeling of irradiation growth in single crystal Zr

Irradiation growth of single-crystal Zr is modeled with a reduced set of cluster dynamics equations. Nucleation and growth of basal and prismatic dislocation loops are both accounted for in the model. Equations are developed for the time evolution of single point defects, a limited number of point defect clusters, interstitial and vacancy loop nucleation rates, as well as for the growth of vacancy loops on basal planes and interstitial loops on prismatic planes. Reduction of the usual infinite hierarchy of cluster dynamics equations to the simple set studied here is justified on the physical basis of the stability of small loops once nucleated. This simplified cluster dynamics model with a small number of adjustable parameters avoids the complexity of explicit representation of higher order point defect clusters. The model shows consistency with experimental observations of the following aspects: (1) the growth rates of Zr crystals along the a- and c-axes; (2) the onset dose for breakaway irradiation growth; (3) the saturation dislocation loop densities of vacancy and interstitial loops; (4) the effects of cold work and temperature on irradiation growth.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Local chemical ordering of a neutron-irradiated CrFeMnNi compositionally complex alloy

While ion-irradiation studies are a critical first step in studying compositionally complex alloys (CCAs) for nuclear applications, they do not capture all the microstructural changes occurring under the low irradiation dose rates and different particles’ scattering patterns experienced in a nuclear reactor setting. To explore these phenomena in reactor-relevant conditions for the first time in CCA, the single-phase solid-solution Cr 10 Fe 30 Mn 30 Ni 30 was neutron irradiated up to 6.61 displacements per atom at 395 and 579 °C. Irradiation-enhanced local chemical ordering (LCO) well beyond the range of short range ordering was observed, and is predicted to be the precursor to the precipitation of a coherent Ni-Mn L1 0 phase and a Cr-rich α’ phase, though TEM analysis did not indicate the presence of either in any irradiation condition. The line density of faulted dislocation loops decreased from 6.47 to 1.69 ∙ 10 15 m -2 from 3.43 to 6.61 dpa at 579 °C despite no appreciable faulted loop content in the unirradiated material. LCO is expected to increase the complexity of the energy landscape within this alloy, restricting interstitial point defect mobility and creating local regions of greater stacking fault energy. These contribute to the negative correlation between irradiation dose and faulted dislocation loop density in this study, as well as the lack of void swelling observed.

36 MATERIALS SCIENCE↗