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Results for “Irradiation Defects”

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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

The effects of high energy deuteron ion beam irradiation on the tensile behavior of HT-9

Ion beam implantations are widely performed to understand the effects of irradiation-induced displacement damage on nuclear structural materials. However, the volume of material that can be investigated is often limited by the maximum energies of accelerator facilitates, leading to a limitation in the thickness of samples whose mechanical properties can be evaluated. The Lawrence Berkeley National Laboratory’s 88-Inch Cyclotron offers a wide range of ions and energies, allowing for material ion implantations at larger scales than typical. Four HT-9 SS-J-geometry tensile specimens were polished and then irradiated with deuterons at the 88-Inch Cyclotron to doses of approximately 0.2 dpa prior to small scale tensile testing. Here, the results from this study show irradiation hardening characterized by the tensile test results and black dot irradiation defects. Additionally, a comprehensive look at low temperature irradiations of high-Cr F/M steels is presented and our results show agreement with the available data.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Quantifying dislocation-type defects in post irradiation examination via transfer learning

The quantitative analysis of dislocation-type defects in irradiated materials is critical to materials characterization in the nuclear energy industry. The conventional approach of an instrument scientist manually identifying any dislocation defects is both time-consuming and subjective, thereby potentially introducing inconsistencies in the quantification. This work approaches dislocation-type defect identification and segmentation using a standard open-source computer vision model, YOLO11, that leverages transfer learning to create a highly effective dislocation defect quantification tool while using only a minimal number of 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 two alloys not represented in the training set. Inference of dislocation defects using transmission electron microscopy on three different irradiated alloys relevant to the nuclear energy industry are examined in this work with widely varying pixel noise levels and with completely unrelated composition and dislocation formations for practical post irradiation examination analysis. Code and models are available at https://github.com/idaholab/PANDA.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Status Report on Characterization of High Burnup Fuel with Advanced Nondestructive Pulsed Neutron PIE

Characterizing irradiated or spent nuclear fuels with pulsed neutron techniques provides microstructural data such as phase fractions as well as crystallographic data, e.g. lattice parameters, from diffraction analysis. Diffraction characterization is complemented by spatially resolved mapping of isotope densities from energy-resolved neutron imaging, in particular neutron absorption resonance imaging, and overall bulk isotope assay with better sensitivity for minority isotopes from neutron absorption resonance spectroscopy without spatial resolution. Furthermore, after characterization at ambient condition, heating of irradiated or spent fuel will allow to characterize differences of e.g. lattice thermal expansion or phase transition temperature and kinetics compared to fresh fuel as well as enable the study of disappearance of irradiation defects. This data enables benchmarking of predictions of properties of irradiated fuels for which otherwise experimental data is sparse. The effort described here strives to characterize a section cut from a high-burnup fuel. Volumes smaller than entire fuel pellets or rodlets as proposed here, e.g. sections cut from a fuel pellet, to pave the way to characterize entire pellets or rodlets in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Indirectly probing the structural change in ion-irradiated Zr-Based metallic glasses from small scale mechanical tests

Ion irradiation was applied to tailor the structural heterogeneities in Zr-based metallic glasses at room temperature. Experimental methods of X-ray diffraction, nanoindentation, and micropillar compression were conducted to examine the irradiation effects on their structural and mechanical property changes. It is found that the irradiated materials retained amorphous structure after room-temperature Ni ion irradiation. The reduction of elastic modulus and hardness measured by nanoindentation indicated the irradiation-induced mechanical degradation. A unified statistic model was employed to quantitatively predict the density and strength of irradiation defects, although their specific structural and physical nature is not explicitly included in this model. The transition of non-intersecting shear bands to multiple intersecting shear bands was observed on compression tests of irradiated micropillars with the increase of irradiation dose. Furthermore, the analysis of the displacement excursion of micropillar compression tests indicated a different deformation mode from the unirradiated state. These results from nanoindentation pop-in and micro-pillar compression tests suggested that irradiation eventually leads to a new state with different types and characteristics of structural heterogeneities from violent displacement cascades and non-equilibrium energy deposition/dissipation processes, which also proves ion irradiation as an effective method to tune the structure and mechanical properties of metallic glasses.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Generalized kinetic model for defect evolution in irradiated materials

A fundamental and open problem in materials science is to determine how the structural properties of irradiated and self-irradiated materials evolve in time. Structural defects generated by irradiation lead to changes in a material’s macroscopic properties, and these macroscopic changes, such as volumetric swelling, can in turn give rise significant alterations of the functionality of a material. Here, we develop a unified analytical model to understand the complex interplay between aggregation, fragmentation, and recombination processes involving mobile clusters of defects such as bubbles, voids, and interstitial atoms in materials subject to radiative sources. Specifically, we employ a mean field approach to derive a system of coupled kinetic equations that describes both the time evolution of the density of each type of defect and how clusters of each defect form and grow as the material ages.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiated Microstructure of Zorita Materials

Reactor baffle plate materials were harvested from the decommissioned Zorita nuclear power reactor after 38 years of service and were characterized with transmission electron microscope. The samples were Type 304 stainless steels and had been irradiated to 0.05, 16.9, ad 48.6 dpa. Faulted dislocation loops, voids, and irradiation-induced precipitates were observed in this material. While dislocation loops were the dominant irradiation defects for all samples, small voids and precipitates were also observed at 16.9 and 48.6 dpa, but not at 0.05 dpa. The size of dislocation loops increased with irradiation dose and saturated below 16.9 dpa. The loop density continued to increase between 16.9 and 48.6 dpa. For voids, both the density and mean size were small, and the volume swelling was negligibly small even at 48.6 dpa. In addition, carbides or G-phase precipitates were detected at 16.9 and 48.6 dpa, and a complex dose dependence was observed. A range of irradiation hardening was estimated at each dose, and the resulting yield strengths were consistent with the previous tensile results measured from the Halden and BOR-60 specimens.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Point defect evolution under irradiation: Finite size effects and spatio-temporal correlations

The evolution of point defect concentrations under irradiation is controlled by their diffusion properties, and by their formation and elimination mechanisms. The latter includes the mutual recombination of vacancies and interstitials, and the elimination of point defects at sinks. Two models are traditionally used to predict the defect concentration evolution, the standard rate theory (SRT) and the atomistic kinetic Monte Carlo (AKMC). We show in this work a large discrepancy in the defect concentrations between both methods when the average number of defects in the AKMC simulation box is close or lower than one. The reason is that AKMC naturally captures strong space and time correlations between vacancies and interstitials generated by the finite size of the periodic simulation box. These correlations strongly affect the recombination rate and the point defect concentrations. SRT fails to predict such correlations, and the corresponding solution deviates from the more accurate solution given by the AKMC simulation under similar conditions. These finite size correlation effects are strong when the elimination of point defects occurs by recombination only, but can still be significant in the presence of sinks. In order to account for the spacio-temporal correlation in a continuum framework, we introduce a Correlated Pair Theory (CPT). This theory fully takes into account the correlations between vacancy and interstitial pairs and predicts point defect concentrations in good agreement with AKMC simulations. Inversely, if the goal is for the AKMC to reproduce defect concentrations in bulk using small simulation boxes, the naturally occurring correlations need to be corrected. We show here that the CPT can be used to modify the elimination rates in the AKMC simulations, so as to yield point defect concentrations in agreement with SRT.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Microstructural characterization of nuclear graphite: from the microscale to the nanoscale

Multiple reactor designs use graphite as a moderator of the nuclear reactions and as structural support. During the lifetime of the reactor, multiple aging factors such as neutron irradiation, oxidation, and temperature along with others induce changes in the microstructure and crystal lattice of graphite components. The pore morphology and crystal structure of some phases in graphite can be used to trace the evolution of irradiation defects and mechanical properties of graphite. We present a combination of results from several microscopy techniques to investigate the differences between nuclear graphite grades and the effects of neutron irradiation and oxidation at multiple length scales. This multiscale approach is needed to understand the microstructural variations caused by the raw materials and manufacturing processes as well as how the different phases of graphite are affected by the reactor environment. The results provide insight into the oxidation- and radiation-induced changes of graphite and create a robust baseline of microstructure information that can be used for the selection of materials for the next generation of nuclear power stations. Moreover, the experiments conducted in this work provide an overview of the advantages and limitations of the most common techniques used to characterize nuclear graphite and how these techniques might be applied to study other carbon-based materials used in the nuclear industry.

Arregui Mena, Jose'↗
DOE OSTI2022

Role of point defects in stress-induced martensite transformations in NiTi shape memory alloys: A molecular dynamics study

Shape-memory properties of equiatomic NiTi rely on the thermal- or stress-induced reversible martensitic phase transformation between the $\textit{B}$2 and $\textit{B}$19' phases. Irradiation defects suppress thermal-induced transformations, but their effects on stress-induced transformations are poorly understood. In this work, we use molecular dynamics to investigate the effect of vacancies, vacancy clusters, interstitials, and antisite defects on stress-induced transformations. All defects suppress the transformation, but vacancy clusters do so to the greatest extent, while antisite defects do so to the least extent. The responsible mechanisms are grain boundary pinning and chemical disordering.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Review of SiC material development for nuclear fusion applications: Cross-cutting research and emerging opportunities

The SiC-based materials, particularly SiC-fiber-reinforced SiC matrix (SiC/SiC) composites, show strong potential for structural and functional applications in future fusion power plants because they can operate at high temperatures with a range of coolants and breeders, thereby enabling higher energy conversion efficiency. Here, this paper presents recent advancements in the development of SiC-based materials, focusing on processing techniques and material performance and resistance under fusion-relevant environments. The processing activities have emphasized near-net-shape fabrication and the joining of SiC subcomponents, with processing methods and material compositions informed by previous irradiation experiments on various grades of SiC. Research on irradiation effects has remained focused on degradation mechanisms and the microstructural optimization of SiC/SiC composites irradiated to high neutron damage levels. Analysis of irradiation defects in SiC has advanced via the application of cutting-edge characterization methods, among which Raman spectroscopy is becoming a common tool to assess atomic-scale chemical disorder. Fusion–fission crosscutting irradiation research has explored combined effects in SiC/SiC composites with application-relevant geometries, including bowing of SiC/SiC composite channels under neutron flux gradients, stress evolution in SiC/SiC composite tubes under through-thickness temperature gradients, and irradiation-enhanced corrosion in SiC. Finally, research opportunities for component testing and assessment under fusion-relevant conditions, in support of emerging concepts from the private fusion sector, are discussed.

Advanced manufacturing↗
DOE OSTI2020

Plasticity of irradiated materials at the nano and micro-scales

Here, we review here our recent work on plastic deformation in irradiated materials at the nano- and micro-scales, as revealed by Discrete Dislocation Dynamics (DDD) simulations. Two methods of including irradiation effects in the DDD framework are presented. The first directly captures the atomistic interaction mechanisms, while the second can effectively study high-dose irradiation. Computer simulations lead to new understanding of the dynamics of collective dislocation-irradiation defect interactions, as well as the quantitative analysis of the temporal and spatial characteristics associated with plastic instabilities. Based on these insights, theoretical models are developed to predict the critical conditions for dislocation channel formation. A simple probability model is proposed and demonstrated to predict the width of dislocation channels in bulk irradiated materials with good agreement with experimental data. The fundamental understanding of the origins of plastic flow localization in irradiated materials sheds light on the design of future generations of radiation-resistant materials.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Understanding the effect of irradiation and silicon carbide microstructure on silver diffusion in TRISO particles

The silicon carbide (SiC) layer in tristructural isotropic (TRISO) fuel particles serves as a barrier to prevent the escape of fission from the fuel kernel. The release of silver (Ag) is a concern due to the long half-life of the 110mAg isotope. In this study, a combination of atomistic and phase-field methods focuses on quantifying the effect of irradiation and SiC microstructure on Ag diffusion. Atomistic calculations provide the Ag diffusivity in SiC bulk and GBs, also providing an estimate of the effect of irradiation defects on Ag diffusivity in bulk SiC. These lower length scale efforts are leveraged to develop a mesoscale effective Ag diffusion coefficient (Deff) in SiC. Deff is defined as a function of temperature, microstructure variables, and fluence. Deff is implemented in the fuel performance code Bison to predict Ag release from AGR-1 and AGR-2 TRISO fuel particles. We hereby quantify the impact of SiC grain size and irradiation on Ag release and improve Bison's predictions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
NASA NTRS1980

Study of radiation induced deep-level defects in proton irradiated AlGaAs-GaAs solar cells

Radiation induced deep-level defects (both electron and hole traps) in proton irradiated AlGaAs-GaAs p-n junction solar cells are investigated along with the correlation between the measured defect parameters and the solar cell performance parameters. The range of proton energies studied was from 50 KeV to 10 MeV and the proton fluence was varied from 10 to the 10th power to 10 to the 13th power P/sq cm. Experimental tools employed include deep-level transient spectroscopy, capacitance-voltage, current voltage, and SEM-EBIC methods. Defect and recombination parameters such as defect density and energy level, capture cross section, carrier lifetimes and effective hole diffusion lengths in n-GaAs LPE layers were determined from these measurements.

Li, S. S.↗