Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. In this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO2 formation via Cr2O3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al2O3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H2 generation by orders of magnitude relative to bare Zry-4.

Here, we employ a combination of advanced synchrotron-based scattering characterization techniques to understand and unravel the atomic origins of the colossal stored energy release in neutron-irradiated silicon carbide. The quantification of the neutron irradiation-induced defects and their impact on the structure-property relationship are important for the design and safe operation of advanced fission and fusion reactors. Our experimental results show that the atomic structure in the as-irradiated samples is significantly perturbed due to a large fraction of vacancy- and interstitial-type defects that lead to complex microstructures and additional components in the x-ray diffraction and pair distribution function (PDF) results. We directly correlate the stored energy release to the recovery of the sublattices with PDF analysis, highlighting that the carbon interstitial- and vacancy-type defects contribute to stored energy more than those of silicon. We find these results to be striking and believe our discoveries to be timely and noteworthy given the technological importance of silicon carbide to the nuclear fission and fusion communities.

Characterization methods capable of providing critical information across multiple structural length scales are essential in materials exposed to the extreme environments such as anticipated fusion power systems. Complementary techniques capable of uncovering the complicated microstructural irradiation-induced evolution are also important to verify and validate advanced computational models. To date, the primary microstructural tools informing such lower-length scale models have included analytical electron microscopy, positron annihilation spectroscopy, atom probe tomography, and small-angle neutron scattering. In this paper, we discuss the application of state-of-the-art synchrotron-based x-ray characterization methods in fusion material research. Specifically highlighted are opportunities in leveraging synchrotron-based techniques to address fundamental and applied materials science challenges at various length scales and in support of modeling efforts. Examples presented in this article include: a combined small angle x-ray scattering and x-ray diffraction study of transmutation-induced precipitation in neutron irradiated tungsten, and the identification of size and structure of nm-scale transmutation precipitates and voids; quantitative characterization of thermodynamically predicted minor precipitate populations in advanced reduced activation ferritic-martensitic steels through high energy x-ray diffraction; and a review of recent synchrotron-based studies dedicated to quantifying the radiation response of fusion relevant materials. The latter includes a pair distribution function analysis investigation of neutron irradiated SiC with insights into the different radiation response of the silicon and carbon sublattices, and a dose dependent decrease in the size of defect free material.

Neutron irradiation tests were carried out on 3D-printed SiC derived from binderjet additive manufac-turing and chemical vapor infiltration. Irradiation was carried to 2.3 dpa over a temperature range of 400-850 degrees C. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the ma-terial. No degradation in strength was observed in the material before or after irradiation, at various temperatures, or in different orientations. Electron microscopy of the microstructure after neutron irradi-ation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed. (c) 2021 Elsevier B.V. All rights reserved.

Swelling of SiC at 300 degrees C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of -50V, -75V, and -100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. At the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress. (C) 2021 Elsevier B.V. All rights reserved.

Titanium carbide (TiC) is an ultra-high temperature ceramic with potential as a structural material candidate for advanced reactor concepts. However, the irradiation tolerance of TiC is not well understood. Here, we reveal the key irradiation damage microstructure degradation processes in TiC using mixed spectrum neutron irradiations at dose of ∼2 displacements per atom (dpa) at temperatures of ∼220, 620, and 1115 °C, combined with state-of-art microstructure characterization using transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). The helium injection (∼65 atomic parts per million) produced by neutron transmutation also occurred in the sample. TiC was observed to form irradiation-induced interstitial-type dislocation loops and He-stabilized cavities. At 220 and 620 °C, the analysis of the electron diffraction patterns, rel-rod imaging and HRTEM revealed that the dislocation loops were faulted Frank loops with Burgers vector bFrank = 1/3<111> lying on {111} planes. A detailed Burgers vector identification performed by the g.b technique revealed that the dislocation loops forming at 1115 °C were unfaulted, edge-type, with Burgers vectors b = a<100> and a/2<110> with corresponding {100} and {110} habit planes. Using continuum mechanics, we estimated the critical radius at which a faulted dislocation loop transitions to a perfect loop to be 9 nm. Further, no amorphization occurred in TiC under-investigated irradiation conditions while macroscopic swelling under point defect swelling regime was observed. He-stabilized cavities were detected at 1115 °C, but not at lower temperature irradiation. These observations indicate the onset of long-range vacancy migration in TiC occurs between 620 and 1115 °C.

Silicon carbide (SiC) fiber?reinforced SiC matrix (SiC/SiC) composites have been widely investigated for potential fusion reactor applications. In this present investigation, the high-temperature creep performance of five types of SiC fibers is evaluated and microstructural analysis is performed. The creep behavior of the fibers was assessed by the bend stress relaxation method at various applied strains at 1500?C and 1700?C. The fibers tested include developmental-grade fibers with different residual silicon amounts (similar to 0%, 2% to 3%, and 5% to 6%) fabricated by laser chemical vapor deposition at Free Form Fibers. Generally, the creep behavior of the Free Form (FF) fibers was similar to Hi-Nicalon Type S and/Tyranno-SA SiC fibers currently used for fabrication of SiC/SiC composites for fusion applications. However, all FF fibers exhibited the formation of pores after the creep tests at 1700?C regardless of residual silicon amount, which can be improved by further development via optimization of the composition and microstructure.

Neutron irradiation in mixed spectrum reactors is an effective approach to investigate the synergistic effects of solid transmutations and atomic displacement damage in tungsten. In this article the findings and interpretations from a large-scale tungsten irradiation campaign conducted using the High Flux Isotope Reactor of Oak Ridge National Laboratory are summarized. The response of originally-unalloyed tungsten to mixed spectrum neutron irradiation is characterized by enormous hardening, degradation of modulus of toughness eventually leading to embrittlement, and decreased thermal conductivity. These property changes may be attributed primarily to production, segregation, and precipitation of rhenium and osmium and assisted by displacement damage. The stages describing the evolution of the microstructural development and property changes with the neutron dose and the accompanying accumulation of transmutation products are proposed. This manuscript has been co-authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan). (C) 2019 Elsevier B.V. All rights reserved.

SiC ceramic matrix composites are a potential replacement for current light water nuclear reactor fuel cladding material. However, loss of fission gas via micro-cracks and corrosion remain an issue. Cathodic arc Cr, CrN, and TiN coatings were deposited on SiC tubes and plates to provide hermeticity and corrosion resistance. These coatings were characterized to determine as-deposited quality. Cross-sectional microscopy, X-ray diffraction, glow discharge optical emission spectroscopy, and scratch tests were performed to evaluate the purity, structure, and mechanical performance of the coatings. Nitride coatings had stable interfaces, but larger defects in the coatings as compared to the Cr coatings which showed cracking at the interface, but less deposition-induced defects. Despite the local state of the interface, the mechanical properties of the metallic coatings versus ceramic coatings enabled the Cr coatings to resist loads three times that of the nitride coatings during scratch tests. Glow-discharge optical emission spectroscopy showed that improvement in elemental purity is needed for future coatings.

Incorporating Raman spectroscopy with transverse lift-out specimens is demonstrated to effectively characterize depth-dependent ion-irradiation damage in nuclear ceramics, such as SiC/SiC composites irradiated up to 1, 10 and 50 displacements per atom (dpa) at 350 degrees C using 10 MeV Au ions. Raman spectroscopy reveals irradiation-induced structural disorder saturation in both SiC-fiber and SiC-matrix at doses as low as 1 dpa, despite vastly different microstructures, inferred from similar longitudinal optical (LO) and transverse optical (TO) phonon peak shifts. Diamond (D) and graphitic (G) peaks from SiC-fibers disappear under irradiation, revealing irradiation-induced carbon packet loss. The irradiation-induced carbon packet loss is also verified by conducting TEM on same FIB foils used for Raman spectroscopy. In a previous study, the irradiation-induced SiC-fiber shrinkage is known to occur due to carbon packet loss in fibers. (C) 2019 Elsevier B.V. All rights reserved.

The initial results of a post-irradiation examination study conducted on a SiC-SiC tubular specimen irradiated under a high radial heat flux are presented herein. The elastic properties of the specimen were evaluated before and after the irradiation using the resonant ultrasound spectroscopy (RUS) technique. The composite tubular specimen was considered as an orthotropic elastic with nine elastic constants (Young's moduli, shear moduli and Poisson's ratios—three components of each) for representing its full elastic deformation behavior. All the elastic moduli decreased after irradiation; the reduction was as high as 35% in one of the moduli. The significant decrease in the moduli indicates the presence of microcracks. The results from a computational study show significant stress development in the specimen due to irradiation, primarily caused by differential swelling across the thickness of the specimen. The evaluated stresses exceed the proportional limit stress of the material, indicating the likelihood of matrix microcracking, and thus corroborating the results obtained from RUS. X-ray Computed Tomography (XCT) study confirmed the presence of cracks in the irradiated specimen. These cracks occurred at the inner region of the specimen and propagated in axial and hoop directions. These XCT results are in agreement with the RUS results and stress distribution results from the computational study.

The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. The optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.

Silicon carbide (SiC) fiber reinforced SiC matrix composites continue to undergo development for fusion applications worldwide because of inherent advantages of the material including low activation, high temperature capability, relatively low neutron absorption, and radiation resistance. This paper presents an international overview of recent achievements in SiC-based composites for fusion applications. Key subjects include applications in fusion reactors, high-dose radiation effects, transmutation effects, material lifetime assessment, and development of joining technology (processing, test method development, irradiation resistance, and modeling capability). This paper also discusses synergy among research for fusion materials and non-fusion materials (for fission and aerospace applications). Finally, future research directions and opportunities are proposed.

Ultra-high temperature transition-metal ceramics are potential candidates for fusion reactor structural/plasma-facing components. We reveal the irradiation damage microstructural phenomena in Boron-11 enriched titanium diboride (TiB2) using mixed-spectrum neutron irradiations, combined with state-of-art characterization using transmission electron microscopy (TEM) and high resolution TEM (HRTEM). Irradiations were performed using High Flux Isotope Reactor at -220 and 620 degrees C up to 2.4 x 10(25) n.m(-2) (E> 0.1 MeV). Total dose including contribution from residual Boron-10 ((10) B) transmutation recoils, was similar to 4.2 displacements per atom. TiB2 is susceptible to irradiation damage in terms of dislocation loop formation, cavities and anisotropic lattice parameter swelling induced micro-cracking. At both 220 and 620 degrees C, TEM revealed dislocation loops on basal and prism planes, with nearly two orders of magnitude higher number density of prism-plane loops. HRTEM, electron diffraction and relrod imaging revealed additional defects on {10 (1) over bar0} prism planes, identified as faulted dislocation loops. High defect cluster density on prism planes explains anisotropic a-lattice parameter swelling of TiB2 reported in literature which caused grain boundary micro-cracking, the extent of which decreased with increasing irradiation temperature. Dominance of irradiation-induced defect clusters on prism planes in TiB2 is different than typical hexagonal ceramics where dislocation loops predominantly form on basal planes causing c-lattice parameter swelling, thereby revealing a potential role of c/a ratio on defect formation/aggregation. Helium generation and temperature rise from residual B-10 transmutation caused matrix and grain boundary cavities for the irradiation at 620 degrees C. The study additionally signifies isotopic enrichment as a viable approach to produce transition-metal diborides for potential nuclear structural applications. (C) 2018 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

This white paper examines novel high temperature materials that have the potential to enable break-through concepts for plasma-facing, blanket, and/or structural components in fusion reactors. Such materials include 1) emerging materials of high interest in the general materials science and engineering (MSE) community [such as MAX phases, ultra-high temperature ceramics (UHTC), and high entropy alloys (HEA)], 2) novel structural materials that may be specifically developed for fusion based on recent materials science advancements [such as castable nanostructured alloys (CNA) and MAX-phase ceramic matrix composites (CMC)], and 3) tungsten (W)-based refractory multi-functional composites that potentially enable extensive use of W in fusion reactors. The US fusion energy sciences program is uniquely positioned to explore and potentially adopt new materials due to its strong connections to the broader US materials science community and the current timeline flexibility for DEMO development. These emerging materials exhibit several clear and outstanding advantages over the current reference PFC and blanket materials, presenting potential game-changing opportunities for improved fusion energy systems.

Monolithic silicon carbide (SiC) to SiC plate joints were fabricated and irradiated with neutrons at 270 -310 degrees C to 8.7 dpa for SiC. The joining methods included solid state diffusion bonding using titanium and molybdenum interlayers, SiC nanopowder sintering, reaction sintering with a Ti-Si-C system, and hybrid processing of polymer pyrolysis and chemical vapor infiltration (CVI). All the irradiated joints exhibited apparent shear strength of more than 84 MPa on average. Significant irradiation-induced cracking was found in the bonding layers of the Ti and Mo diffusion bonds and Ti-Si-C reaction sintered bond. The SiC-based bonding layers of the SiC nanopowder sintered and hybrid polymer pyrolysis and CVI joints all showed stable microstructure following the irradiation. (C) 2017 Elsevier B.V. All rights reserved.

SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.

Fabrication of large, or complex, components from silicon carbide, or other technical ceramics, used in heat exchanger devices, energy production and chemical synthesis systems, and for components within fusion and fission reactors require robust joining processes. Ceramatec has developed a novel method for achieving bonds using an air brazing process. For silicon carbide joining, the braze acts under certain conditions to promote diffusion bonding. The resulting joined regions are thought to form by rapid interdiffusion of the diffusion enhancing braze material and silicon and carbon species, resulting in a microstructure more similar to one formed by diffusion bonding than brazing. Processing of these joints is accomplished at relatively low temperatures, 900'C -1200'C in air, with minimal applied load. The brazed joint strength was found to be statistically equivalent to monolithic control samples at room temperature. Oxidation testing, using dry oxygen and saturated steam, was conducted at 1000 C for 1000 hours on joined specimens, resulting in further microstructural development of the joint, with subsequent shear testing showing no appreciable reduction in strength. Torsion tests on irradiated joined samples show that the joint's mechanical integrity is resistant to radiation degradation.