Yan-Ru Lin

Helium bubble formation was examined by scanning/transmission electron microscopy (S/TEM) in Fe-9/10Cr binary alloys and two dispersion strengthened nanostructured alloys (CNA3 and 14YWT containing 5–10 nm diameter carbide and oxide particles, respectively) after ex-situ and in-situ He implantation to ~10,000 appm at 500 to 900 °C. The combination of high-resolution STEM images and electron energy loss spectroscopy (EELS) revealed that the Y-Ti-O nanoparticles in 14YWT were uniformly distributed and exhibited a one-to-one relationship for bubble attachment to the nanoclusters. In the in-situ experiment at 900 °C, grain boundary cracking was severe in the Fe-10Cr model alloy, but not in the nanostructured alloys. From 500 to 900 °C, the bubble size generally increased with increasing irradiation temperature, while the bubble density decreased with increasing temperature. At the same temperatures, the bubble size in the implanted materials was in the order of Fe-9/10Cr > CNA3 > 14YWT, while the bubble density showed the opposite order. The observed bubble number densities for the nanostructured alloys are comparable to the nanoparticle density, suggesting that the nanoparticles in both alloys were effective in trapping He. Our results indicate that very high He concentrations can be managed in nanostructured alloys by sequestering the helium into smaller bubbles (which leads to a lower volume swelling value) and to shield He from the grain boundaries. This can be attributed to the much higher sink strength associated with the nanoclusters or the He trapping ability between different types of nanoclusters.

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe3+ and 3.42 MeV energy degraded He2+) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. The comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

Carbon materials have become increasingly diverse, finding applications in high-temperature and high-radiation environments. Glassy carbon, an allotrope known for its exceptional chemical inertness and desirable mechanical properties. However, understanding neutron irradiation effects in glassy carbon has proven challenging, primarily because of its unique nanopore structure. This study presents a highly detailed microstructural characterization investigation of neutron-induced changes in glassy carbon, revealing how changes in nanopore structure and crystallinity impact the irradiation-induced shrinkage. Aberration-corrected scanning transmission electron microscopy (STEM) reveals pore closure that leads to material densification in the irradiated samples. Dimensional analysis combined with comparison to historical data suggest significant length shrinkage to occur. Neutron and in situ electron irradiation experiments suggest that glassy carbon transforms into so-called carbon onions, supporting the concept of shrinkage saturation. Investigating irradiation temperature effects using STEM, electron energy loss spectroscopy, x-ray diffraction, and Raman spectroscopy revealed partial amorphization at 210 °C–230 °C and preserved order in glassy carbon at 860 °C, coinciding with pore closure. Thermal property measurements were also conducted to assess the effects of densification and other changes in the atomic structure of glassy carbon. The results of this study have broad implications in the deployment of glassy carbon to nuclear environments, based around the observed changes in the thermal properties, and demonstrates the operational window for the onset of densification.

Nanostructured ferritic alloys (NFAs), such as oxide-dispersion strengthened (ODS) alloys, play a vital role in advanced fission and fusion reactors, offering superior properties when incorporating nanoparticles under irradiation. Despite their importance, the high cost of mass-producing NFAs through mechanical milling presents a challenge. This study delves into the microstructure-mechanical property correlations of three NFAs produced using a novel, cost-effective approach combining severe plastic deformation (SPD) with the continuous thermomechanical processing (CTMP) method. Analysis using scanning electron microscopy (SEM)-electron backscatter diffraction (EBSD) revealed nano-grain structures and phases, while scanning transmission electron microscopy (STEM)-energy dispersive X-ray spectroscopy (EDS) quantified the size and density of Ti-N, Y-O, and Cr-O fine particles. Atom probe tomography (APT) further confirmed the absence of finer Y-O particles and characterized the chemical composition of the particles, suggesting possible nitride dispersion strengthening. Correlation of microstructure and mechanical testing results revealed that CTMP alloys, despite having lower nanoparticle densities, exhibit strength and ductility comparable to mechanically milled ODS alloys, likely due to their fine grain structure. However, higher nanoparticle densities may be necessary to prevent cavity swelling under high-temperature irradiation and helium gas production. Further enhancements in uniform nanoparticle distribution and increased sink strength are recommended to mitigate cavity swelling, advancing their suitability for nuclear applications.

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 ½<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.

Glassy carbon, a monoatomic allotrope of carbon, is a candidate material for components in fission nuclear power systems due to its radiation tolerance. This article presents comprehensive electron microscopy data revealing the effects of neutron and electron irradiation on glassy carbon. For comparison, additional data are provided for pyrolytic graphite and carbon fibers, materials that exhibit similar structural behavior under irradiation. In situ electron irradiation experiments further illustrate the real-time microstructural evolution of glassy carbon during exposure. The dataset is organized into five parts: (1) transmission electron microscopy (TEM) micrographs of as-received and neutron-irradiated glassy carbon; (2) TEM micrographs of neutron-irradiated graphite; (3) TEM micrographs of unirradiated and irradiated carbon–carbon composites; (4) TEM micrographs of pyrolytic carbon specimens in both conditions; (5) scanning transmission electron microscopy (STEM) micrographs of as-received and neutron-irradiated glassy carbon and (6) in situ electron irradiation data of a glassy carbon particle. These datasets provide valuable insights into radiation-induced structural changes in carbon-based materials relevant to nuclear applications.

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 [1]. 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 [2]. Scanning transmission electron microscopy (STEM) based techniques have emerged as indispensable tools for irradiation-induced defect characterization [3, 4], 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 [5]. Recent advances in electron microscopy, particularly in 4D-STEM [6], 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. In this presentation, we discuss the potential for combining 4D-STEM, weak-beam dark-field (WBDF) STEM, electron tomography, and EDXS for defect analysis, specifically focusing on irradiation-induced dislocation loops in proton-irradiated Fe-5Cr model alloys. This approach offers the advantage of obtaining the 3D distribution of dislocation loops in atomic scale, as well as identifying their type (100 or ⁠) and nature (interstitial or vacancy). Through 4D-STEM strain mapping and STEM-EDX elemental mapping, we can assess the local strain field and strain interactions between loops, while also detecting chemical composition changes near the loops within the same area of interest. 4D-STEM and atomic-resolution STEM revealed that when loops are small (diameter <5 nm), their centers may not precisely overlap with the "black-dot" features (Fig. 1). In the case of an 100 edge-on loop (Fig. 1b), its center may lie between a pair of black-dots, which represent high-strain areas at the two ends of the loop. It was observed that ⟨100⟩ loop strings are composed of ½⟨111⟩ and ⟨100⟩ loops arrayed along ⟨100⟩ directions (Fig. 2). STEM-EDX and analysis revealed Cr enrichment associated with the dislocation loops (Fig. 2f), supporting prior observations that Cr impedes dislocation loop motion, resulting in a more sluggish dislocation loop evolution process in Fe-Cr alloys than in pure Fe [7]. STEM and EDX tomography of the loops and Cr-enriched features will also be presented [8]. The experimental results from these advanced techniques validate simulation models [9], enhancing the understanding of irradiation effects on material properties, crucial for materials development and selection in nuclear applications or other extreme environments [10].