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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Microstructural evolution of 3C-SiC exposed to simultaneous neutron irradiation and helium implantation

The synergism of transmutant helium and irradiation defects is critically important to understand the performance of SiC when in service in nuclear environments. In this study, we exposed 3C-SiC in contact with a 2 μm Ni foil to neutron irradiation at 500 °C to 29 dpa in the high flux isotope reactor to simulate the simultaneous introduction of helium and irradiation defects into SiC. The combination of TEM observations and thermal desorption measurements helps to elucidate the impact of helium on the microstructural evolution in 3C-SiC. Thermal helium desorption spectra from ion-implanted and neutron-irradiated SiC showed a completely different desorption behavior in terms of the peak position and helium desorption flux. The identification of possible helium trapping sites was attempted by applying the first order dissociation model to the measured thermal helium desorption spectra. TEM observations of the neutron-irradiated samples showed that the presence of helium facilitated the stabilization of the defect clusters and promoted the formation of visible helium bubbles in SiC subject to simultaneous neutron irradiation and in-situ He implantation. Following the subsequent heat treatment during the thermal desorption measurements, large faceted helium bubbles were found along grain boundaries in the neutron-irradiated SiC with in-situ He implantation, while the helium bubbles in the grain interior were relatively smaller than those found in the neutron-irradiated sample without in-situ He implantation due to the reduced mobility of He-defect clusters. In conclusion, the He contents contained in the neutron-irradiated samples were quantified based on the experimental data from TEM and thermal desorption measurements.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Irradiation of UC1+x kernels using the MiniFuel vehicle: Microstructure, phase analysis, and initial post-irradiation examination

Uranium carbide is a candidate fuel form for a wide range of advanced reactors, including larger Generation IV reactors as well as small modular reactors and microreactors. However, its commercial deployment timeline faces challenges via traditional qualification approaches. To address this issue, an accelerated fission rate irradiation test was performed to collect basic fuel performance data to inform fuel performance models and potential future integral tests. Hyperstoichiometric UC (UC1+x) kernels were irradiated in the High Flux Isotope Reactor using the MiniFuel irradiation vehicle. The test matrix spanned two temperature regimes (700 °C and 800 °C) and burnup levels (1.8 % FIMA and 2.8 % FIMA). Between 21 and 63 kernels were tested at each unique temperature and burnup condition. As-fabricated microstructural analysis revealed a multiphase composition with UC, UC2, UC2−y, and U-C-O bearing phases for the irradiated kernels. Following irradiation, fission gas release, swelling, and microstructures were analyzed. Measured fission gas release was below 5 % for all irradiation conditions, reaching a maximum at the highest temperature and burnup condition. A binary swelling response was observed; the lower burnup and temperature conditions resulted in negligible swelling, but the higher burnup and temperature conditions produced significant anisotropic swelling and densification in a subset of kernels. The basic microstructural exams of kernels following irradiation were not capable of showing a correlation between kernels that exhibited excessive swelling and those that did not. Characterization of a subset of samples using the Advanced Photon Source and more detailed microstructural examination of unirradiated kernels revealed that a subset of kernels contained very high UC2 phase fractions. The anomalous swelling response is hypothesized to have been driven by this chemical variation. The results of this irradiation highlight the potential of accelerated fission rate irradiation testing to explore such behaviors and inform the development of fuel specifications.

Adorno Lopes, Denise [ORNL] (ORCID:000900023705987↗
DOE OSTI2019

Automated Quantitative Analysis of Extended Irradiation Defects - Dislocations, Voids and Precipitates in Neutron Irradiated HT-9 Steel

Structural alloys that make up the core components of nuclear reactors are exposed to radiation that can substantially alter the microstructure and mechanical properties of the material. Irradiation can cause the formation of extended defects including dislocation line and loops, voids and various types of precipitates. To understand the irradiation effects, it is critical to characterize these defects individually and collectively to quantify how they evolve throughout irradiation stages. Transmission electron microscopy (TEM) is a standard tool for providing extensive size, number density and spatial distribution of extended irradiation defects, however, manual measurement of defect size and number density requires time-intensive effort to analyze enough defects that provide a statistically meaningful result and are prone to errors of subjectivity. In this work, we developed a set of MATLAB algorithms offering robust and automated identification and quantification of extended irradiation defects like dislocation lines, voids and precipitates commonly observed in neutron irradiated cladding alloys [1]. Using a neutron-irradiated HT-9 martensitic alloy irradiated at 412°C to 111.8 dpa, we demonstrate the capability of these algorithms for quantitative defect analysis. Figure 1 presents defect images of the HT-9 using an advanced diffraction contrast imaging scanning transmission electron microscopy (DCI STEM) that suppresses bend contours and provides better defect contrast compared to conventional TEM [2]. The first step of defect analysis is defect identification, a process that can be digitally achieved by contrast thresholding (for simple defect contrast) and by machine learning computer vision for defect semantic segmentation [3]. Here, labeled binary defect images (Figures 2a and 2c) were used. For dislocation density quantification, we employed an optimized circular-grid intersection method first introduced by Dr. John Steeds [4]. This method is shown in Figures 2a and 2b, where the number of intersections of dislocation lines with circular traces was measured. This produces an unbiased sampling of dislocation lines, regardless of their orientation. This process was optimized with automated digital intersection number and grid length registration. Interestingly, it was found that a relatively large number of circles in a set circular gird is was needed to produce an invariant result (Figure 2b), with a low systematic measurement error. This demonstrates the main advantage of computer-based defect analysis, in performing repetitive operations quickly with a reduced systematic error. For the quantitative analysis of voids and precipitates a challenge encountered with automated defect analysis is feature overlap, particularly when the density of voids/precipitates is high. To mitigate this problem, as shown in Figure 2d, we modified the popular circular Hough transformation and treated ‘circular shape’ as the particle descriptor for voids/precipitates, efficiently resolving the overlapping particles. In summary, this automated digital quantification together with machine learning enabled defect identification [3] opens new opportunities for standardizing the characterization workflow for quantitative analysis of extended irradiation defects.

Sainju, Rajat↗
DOE OSTI2024

Understanding Formation of Irradiation-Induced Defects through 4D-STEM, Electron Tomography, and WBDF-STEM

A major challenge in advancing nuclear materials for next-generation fission and proposed fusion reactors is to comprehensively understand the formation of irradiation-induced defects. Here it is essential to correlate the evolution of irradiation-induced defects and the degradation of mechanical properties, as they collectively dictate the material's lifespan and ensure nuclear safety. Scanning transmission electron microscopy (STEM) based techniques have emerged as indispensable tools for irradiation-induced defect characterization, offering high spatial resolution imaging and chemical analysis, such as electron energy loss spectroscopy (EELS) and energy dispersive X-ray spectroscopy (EDXS). These techniques have been effectively used to obtain an atomic-scale view of the defect structure. Recent advances in electron microscopy, particularly in 4D-STEM, offer detailed insight into microstructural evolution by capturing full 2D diffraction patterns at every pixel position. Using high-speed direct electron detectors, this technology generates a four-dimensional dataset, overcoming the limitations of traditional STEM imaging.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Final Report on MEGAPIE Target Irradiation and Post-Irradiation Examination

Megawatt pilot experiment (MEGAPIE) was successfully performed in 2006. One of the important goals of MEGAPIE is to understand the behaviour of structural materials of the target components exposed to high fluxes of high-energy protons and spallation neutrons in flowing LBE (liquid lead-bismuth eutectic) environment by conducting post-irradiation examination (PIE). The PIE includes four major parts: non-destructive test, radiochemical analysis of production and distribution of radionuclides produced by spallation reaction in LBE, analysis of LBE corrosion effects on structural materials, T91 and SS 316L steels, and mechanical testing of the T91 and SS 316L steels irradiated in the lower part of the target. The non-destructive test (NDT) including visual inspection and ultrasonic measurement was performed in the proton beam window area of the T91 calotte of the LBE container, the most intensively irradiated part of the MEGAPIE target. The visual inspection showed no visible failure and the ultrasonic measurement demonstrated no detectable change in thickness in the beam window area. Gamma mapping was also performed in the proton beam window area of the AlMg 3 safety-container. The gamma mapping results were used to evaluate the accumulated proton fluence distribution profile, the input data for determining irradiation parameters. Radiochemical analysis of radionuclides produced by spallation reaction in LBE is to improve the understanding of the production and distribution of radionuclides in the target. The results demonstrate that the radionuclides of noble metals, 207 Bi, 194 Hg/Au are rather homogeneously distributed within the target, while radionuclides of electropositive elements are found to be deposited on the steel-LBE interface. The corrosion effect of LBE on the structural components under intensive irradiation was investigated by metallography. The results show that no evident corrosion damages. However, unexpected deep cracks were found in the EBW (electron beam weld) of the LBE container in the intensive irradiation zone of the target, which should be formed during irradiation. In the SS 316L steel of the flow guide tube, inclusions or precipitates enriched with O, Si, S, Ca, Ti and Mn were observed. Many of them are very long, up to a few mm, and located on grain boundaries along the extrusion direction of the tube. The degradation of the mechanical properties of the T91 and SS 316L steels has been investigated by conducting tensile tests on the specimens extracted from the T91 and SS 316L components in the intensive irradiation region. The results obtained from the proton beam window of the T91 calotte exhibit a good ductility of T91 steel after irradiation at 6-7 dpa (displacement per atom) in contact with flowing LBE.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Interpretation of Ion Irradiation and Neutron Irradiation Damage in Additively Manufactured 316 Stainless Steel using Multiscale Modeling

The accelerated adoption of nuclear energy necessitates advanced manufacturing technologies, such as additive manufacturing, to meet heightened supply chain requirements and support innovative reactor technologies. Due to the unique microstructural characteristics of additively manufactured materials under distinct solidification conditions, comprehensive evaluation of their performance in reactor environments is essential. The Advanced Materials and Manufacturing Technologies program under the Department of Energy's Office of Nuclear Energy focuses on understanding the irradiation performance and damage evolution of laser powder bed fusion 316 stainless steel, with an emphasis on integrating ion and neutron irradiation data to accelerate the development and qualification of materials for advanced nuclear reactor applications. While ion irradiation is a cost- and time-effective method, modeling and simulation are required to interpret the data for the broader range of irradiation conditions encountered in advanced reactors. In fiscal year 2025, integrated multiscale modeling and simulations were conducted to assess irradiation damage in additively manufactured 316 stainless steel. Key outcomes include predictions of chromium enrichment at grain boundaries, nickel enrichment at dislocation cell walls and void surfaces, and heterogeneous void evolution under ion and neutron irradiation conditions. Cluster dynamics simulations revealed the coarsening of voids at high irradiation temperatures and the suppression of void growth by high network dislocation density, while also demonstrating significant growth and coarsening of voids and self-interstitial atom loops at low dose rates. Machine learning-accelerated atomistic simulations highlighted the impact of the local environment and chromium concentration on vacancy diffusivity, providing key insights on the influence of composition on void swelling and radiation-induced segregation. Additionally, molecular dynamics simulations demonstrated the presence of defect production bias and a significant effect of carbon content on defect cluster behavior. These combined efforts aim to predict the performance of additively manufactured materials under various reactor conditions, supporting their qualification for nuclear reactor applications by interpreting ion irradiation data. This report underscores the potential of integrated multiscale modeling to analyze ion irradiation data in the effort to accelerate the qualification of additively manufactured materials for nuclear reactor components.

316 stainless steel↗
DOE OSTI2021

Damage development in neutron-irradiated concrete in a test reactor: Hygro-thermal and mechanical simulations

This paper reports the development of a 3D mesoscale hygro-thermal-mechanical simulation approach to predict damage in concrete irradiated in a test reactor. This framework, developed in MOOSE, considers the effects of elevated temperature, moisture content, and high neutron fluence (energy threshold, E > 0.1 MeV) on the mortar and aggregates separately. The first-stage simulation implements hygro-thermal analysis to determine the temperature and RH inside the specimen as a function of imposed radiation energy. These are used as inputs to the second stage, which considers radiation-induced volumetric expansion (RIVE) of aggregates, and creep, shrinkage, and stress-strain response of mortar to predict the expansion, stresses, and damage in specimens made using different coarse aggregates and subjected to different irradiation times. The irradiation time-dependent damage in the mortar is expressed using an isotropic damage parameter. This multi-physics model serves as a predictive tool for damage quantification in concrete due to neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation Damages of Structural Materials under Different Irradiation Environments

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study’s Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

radiation damage, structural alloys, neutron irrad↗
DOE OSTI2016

Lattice damage and compositional changes in Xe ion irradiated In{sub x}Ga{sub 1-x}N (x = 0.32−1.0) single crystals

Lattice disorder and compositional changes in In{sub x}Ga{sub 1-x}N (x = 0.32, 0.47, 0.7, 0.8, and 1.0) films on GaN/Al{sub 2}O{sub 3} substrates, induced by room-temperature irradiation of 5 MeV Xe ions, have been investigated using both Rutherford backscattering spectrometry under ion-channeling conditions and time-of-flight secondary ion mass spectrometry. The results show that for a fluence of 3 × 10{sup 13 }cm{sup −2}, the relative level of lattice disorder in In{sub x}Ga{sub 1-x}N increases monotonically from 59% to 90% with increasing indium concentration x from 0.32 to 0.7; a further increase in x up to 1.0 leads to little increase in the disorder level. In contrast to Ga-rich In{sub x}Ga{sub 1-x}N (x = 0.32 and 0.47), significant volume swelling of up to ∼25% accompanied with oxidation in In-rich In{sub x}Ga{sub 1-x}N (x = 0.7, 0.8, and 1.0) is observed. In addition, irradiation-induced atomic mixing occurs at the interface of In-rich In{sub x}Ga{sub 1-x}N and GaN. The results from this study indicate an extreme susceptibility of the high In-content In{sub x}Ga{sub 1-x}N to heavy-ion irradiation, and suggest that cautions must be exercised in applying ion-implantation techniques to these materials at room temperature. Further studies of the irradiation behavior at elevated temperatures are warranted.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

Irradiation effects in tungsten-copper laminate composite

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410-780°C and fast neutron fluences of 0.02-9.0×1025 n/m2, E>0.1 MeV, 0.0039-1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22°C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22°C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile.

Garrison, L. M.↗
DOE OSTI2015

The dynamics of Al/Pt reactive multilayer ignition via pulsed-laser irradiation

Reactive multilayers consisting of alternating layers of Al and Pt were irradiated by single laser pulses ranging from 100 μs to 100 ms in duration, resulting in the initiation of rapid, self-propagating reactions. The threshold intensities for ignition vary with the focused laser beam diameter, bilayer thickness, and pulse length and are affected by solid state reactions and conduction of heat away from the irradiated regions. High-speed photography was used to observe ignition dynamics during irradiation and elucidate the effects of heat transfer into a multilayer foil. For an increasing laser pulse length, the ignition process transitioned from a more uniform to a less uniform temperature profile within the laser-heated zone. A more uniform temperature profile is attributed to rapid heating rates and heat localization for shorter laser pulses, and a less uniform temperature profile is due to slower heating of reactants and conduction during irradiation by longer laser pulses. Finite element simulations of laser heating using measured threshold intensities indicate that micron-scale ignition of Al/Pt occurs at low temperatures, below the melting point of both reactants.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Bubble Formation and Lattice Parameter Changes Resulting from He Irradiation of Defect-Fluorite Gd2Zr2O7

Pyrochlores have long been considered as potential candidates for advanced ceramic waste-forms for the immobilization of radioactive waste nuclides. This work provides evidence that Gd2Zr2O7, often considered the most radiation tolerant pyrochlore, could be susceptible to radiation damage in the form of bubble nucleation at the highest He doses expected over geological time. Ion irradiations were utilized to experimentally simulate the radiation damage and He accumulation produced by ..alpha..-decay. Samples were pre-damaged using 7 MeV Au3+ to induce the pyrochlore to defect-fluorite phase transformation, which would occur due to ..alpha..-recoil damage within several hundred years of storage in a Gd2Zr2O7 waste-form. These samples were then implanted to various He concentrations in order to study the long-term effects of He accumulation. Helium bubbles 1-3 nm in diameter were observed in TEM at a concentration of 4.6 at.% He. Some bubbles remained isolated, while others formed chains 10-30 nm in length parallel to the surface. GIXRD measurements showed lattice swelling after irradiating pristine Gd2Zr2O7 with 7 MeV Au3+ to a fluence of 2.2 x 1015 Au/cm2. An increase in lattice swelling was also measured after 2.2 x 1015 Au/cm2 + 2 x 1015 He/cm2 and 2.2 x 1015 Au/cm2 + 2 x 1016 He/cm2. A decrease in lattice swelling was measured after irradiation with 2.2 x 1015 Au/cm2 + 2 x 1017 He/cm2, the fluence where bubbles and bubble chains were observed in TEM. Bubble chains are thought to form in order to reduce lattice strain normal to the surface, which is produced by the Au and He irradiation damage.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Response of nanostructured ferritic alloys to high-dose heavy ion irradiation

A latest-generation aberration-corrected scanning/transmission electron microscope (STEM) is used to study heavy-ion-irradiated nanostructured ferritic alloys (NFAs). Results are presented for STEM X-ray mapping of NFA 14YWT irradiated with 10 MeV Pt to 16 or 160 dpa at -100°C and 750°C, as well as pre-irradiation reference material. Irradiation at -100°C results in ballistic destruction of the beneficial microstructural features present in the pre-irradiated reference material, such as Ti-Y-O nanoclusters (NCs) and grain boundary (GB) segregation. Irradiation at 750°C retains these beneficial features, but indicates some coarsening of the NCs, diffusion of Al to the NCs, and a reduction of the Cr-W GB segregation (or solute excess) content. Ion irradiation combined with the latest-generation STEM hardware allows for rapid screening of fusion candidate materials and improved understanding of irradiation-induced microstructural changes in NFAs.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2017

Initial results from safety testing of US AGR-2 irradiation test fuel

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO 2 -kernel TRISO particles have undergone 1600°C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ( 110m Ag, 134 Cs, 137 Cs, 154 Eu, 155 Eu, 90 Sr, and 85 Kr) were monitored while heating the fuel specimens to 1600°C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600°C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO 2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO 2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600°C. However, additional silver release was observed later in the safety testing due to the UO 2 TRISO with failed SiC. Failure of the SiC layer in the UO 2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluation of scenario-specific modeling approaches to predict plane of array solar irradiation

The two main objectives of this study are to evaluate the extent to which the prediction of plane of array irradiance is improved when diffuse radiation is predicted using location-specific regression models developed from onsite measured data as against using generalized models and to estimate the expected uncertainties associated with plane of array irradiance predictions under different data collection scenarios likely to be encountered in practical situations.

Moslehi, Salim↗