The radiation-induced swelling of SiC and its composites, including strong dependencies on temperature and dose, can drive significant lateral bowing in the presence of temperature and/or dose gradients. In recent years, simulations have been performed to assess the extent of bowing in SiC composite light-water reactor (LWR) fuel cladding and boiling water reactor (BWR) channel boxes. However, to date, no integral experimental data exist to validate these models. This work provides the first experimental bowing evaluation of three similar to 380 mm long SiC composite specimens irradiated under varying neutron dose gradients (similar to 50 degrees C-60 degrees C, 0.03-0.06 dpa): two tubes (similar to 9.8 mm diameter) and a miniature BWR channel box (similar to 30 mm square). The measured radiation-induced length swelling (similar to 0.3%-0.7% linear) was consistently 10%-21% higher than values obtained from 3D finite element structural analyses with inputs from 3D radiation transport calculations. This discrepancy could be at least partially explained by differences in dose rate (similar to 10(-8) dpa/s) compared to the literature data (similar to 10(-6) dpa/s) used to establish the dose-to-swelling correlations in the model. Nevertheless, the modeled bowing magnitudes (<2 mm) obtained from finite element analyses and simple analytical equations were within the bounds of the experimental measurements for all specimens. With improved confidence in the ability to predict the structural response and measure the macroscopic deformations, future experiments will target transient bowing under neutron flux gradients at representative LWR temperatures and assess whether grid spacers can mitigate the tens of millimeters of bowing that would otherwise be expected in similar to 4 m long LWR components.

Silicon carbide (SiC) and SiC fiber-reinforced SiC matrix composites (SiC/SiC) are receiving renewed attention for use in next-generation fusion reactors due to their ability to withstand extreme conditions, including high temperatures, neutron irradiation, and plasma interactions. General Atomics Electromagnetic Systems (GA-EMS) has demonstrated significant progress in scaling up the fabrication of SiC/SiC, achieving high mechanical uniformity and meeting dimensional requirements in components up to 12 feet in length. Key developments are discussed including scale-up of the chemical vapor infiltration (CVI) process from lab-scale to full sized parts, high-dose (100 dpa) irradiation testing, nuclear-grade ceramic joining technologies, and production-focused quality control with the collective aim to establish SiC/SiC as a reliable solution for structural and functional components in fusion systems. Beyond manufacturing, the paper addresses supply chain barriers, particularly the limited availability and high cost of nuclear-grade SiC fiber. GA-EMS is developing a novel SiC fiber production method based on a thermochemical cure step that is anticipated to reduce costs compared to traditional approaches. Additionally, advancements in engineered SiC materials, such as SiC foams and tungsten-graded SiC composites, are discussed as promising solutions for specific fusion reactor components.

This paper describes results from the mechanical evaluation of unirradiated SiC fiber–reinforced SiC matrix composite tubes under a controlled high-temperature steam environment. The experiments were aimed at identifying key material degradation behavior under environments relevant to loss-of-coolant accidents of light water reactors. Mechanical tests of the SiC composite tubes at 1000°C under steam and inert environments were conducted using a unique test capability. The material tested was a duplex tube with a thick monolithic SiC layer on the outer surface. The tubes were subjected to preloading at ∼100 MPa in tension before exposure to high-temperature steam with up to 75% of the preload at a constant displacement. In the presence of matrix cracks, the steam exposure caused embrittlement of the SiC composite tubes and failure at a stress level below the pretest stress. The material degradation was explained by a fiber oxidation model, which can be applied to various SiC cladding concepts. The embrittlement could be a limiting factor for using SiC cladding subjected to loss-of-coolant accident conditions.

Demonstration of hermetic SiC fiber–reinforced SiC matrix composite cladding under normal operating environments has been identified as one of the most critical feasibility issues for accident-tolerant fuel cladding in light-water reactors. This study provides critical experimental data needed for understanding the effects of irradiation on hermeticity. SiC composite and monolithic tubes were neutron-irradiated to 2 displacements per atom with and without a nominal radial heat flux of 0.6 MW/m2 to produce a simulated in-pile stress state for the normal operation of a light-water reactor. The through-thickness temperature gradient under irradiation results in a gradient in swelling, which causes a significant stress buildup. Such irradiation-induced stress was modeled using a commercial finite element analysis code. The radial heat flux–irradiation synergism was experimentally investigated by constructing a special irradiation capsule and evaluating the helium hermeticity of the specimens. The simulated stress state exhibited a near equi-biaxial tensile axial and hoop stress of ∼150 MPa at the inner surface of the SiC composite tube. This stress level is potentially beyond the matrix cracking stress. Degradation of hermeticity of the SiC composite tubes was observed after irradiation, indicating irradiation-induced cracking, whereas the irradiated monolithic SiC tubes remained hermetic. The results indicate that loss of hermeticity caused by radiation-induced microcracking is a potential issue for SiC composite cladding, depending on the magnitude of the temperature gradients. Coating the outer surface of the cladding was identified as a mitigation strategy that might overcome the cracking issue.

Silicon carbide (SiC)-based nuclear fission fuel rod cladding has been considered as one of the possible designs for accident tolerant fuels. It is in the form of a SiC fibre reinforced SiC matrix composite tube (SiCf-SiCm) with monolithic SiC outer and/or inner coating layers. This study focuses on the deformation and fracture processes in this material using in situ X-ray micro-computed tomography (XCT) at room temperature (RT) and 1200 °C in an inert gas environment in a C-ring compression loading configuration. Prior to testing, local properties and residual stresses were characterised using nanoindentation and Raman spectroscopy since they can impact the mechanical behaviour of the material. The 3D strain distribution, crack formation and propagation processes including the toughening mechanisms (e.g., crack deflection, micro-cracking, crack bridging and bifurcation) are investigated in the coating and underlying composites at RT and 1200 °C. There is no particular sequence which toughening mechanism occurs first – this is very different from the conventional toughening theory in ceramic-matrix composites under uniaxial tension loading. Indeed, no evidence of fibre pull-out or fibre fracture was observed in this SiCf-SiCm nuclear cladding material in the current C-ring compression configuration. The correlation between local measurements and bulk mechanical behaviour are discussed.

SiC-fiber-reinforced SiC matrix composite cladding for light water reactor fuel elements must withstand high-temperature steam oxidation in a loss-of-coolant accident scenario (LOCA). Current composite designs include an outer monolithic SiC layer, in part, to increase steam oxidation resistance. However, it is not clear how such a structure would behave under high-temperature steam in the case when the monolithic layer cracks and carbon interphases and SiC fibers are exposed to the environment. To fill this knowledge gap, stress-rupture tests of prototypic SiC composite cladding at 1000 degrees C under steam and inert environments were conducted. The applied stress was similar to 120 MPa, which was beyond the initial cracking stress. The failure lifetime under steam was 400-1300 s, while 75% of the composite specimens did not fail after 3 h of total exposure under inert gases. Microstructural observations suggest that steam oxidation activated slow crack growth in the fibers, which led to failure of the composite. The results from this study suggest that stress rupture in steam environments could be a limiting factor of the cladding under reactor LOCA conditions.

Grid-to-rod-fretting (GTRF) in pressurized water reactors (PWRs) is known to cause wear and surface damage on the fuel claddings, potentially leading to radioactive leakage. One of the accident-tolerant fuel (ATF) concepts is to use advanced cladding materials that could withstand higher temperatures. This study investigated the wear behavior of candidate silicon carbide (SiC)-based composite claddings with different levels of surface finish in fretting against a commercial ZIRLO alloy grid using a unique bench-scale autoclave GTRF rig. The experiments mimicked the environment in an industrial full-assembly PWR simulator. Fretting tests were conducted with a realistic load (similar to 0.5 N) in deionized water under a pressure of 20-23 bar at 204 degrees C for 100 h. While the SiC/SiC composite claddings showed significantly higher wear resistance than the commercial ZIRLO alloy cladding as expected, the smoother versions experienced surprisingly higher wear than the much softer counterface, ZIRLO grid. The wear mechanism of the SiC/SiC cladding was attributed to the SiC wear debris that was trapped at the fretting interface causing both 3-body and 2-body (embedded into the grid surface) abrasion of the cladding. Rougher SiC/SiC claddings had less material loss but caused more wear on the ZIRLO grid. Pre-oxidized ZIRLO grid showed better compatibility with the SiC/SiC cladding to protect both the cladding and grid as a result of reduced wear debris trapping.

Journal Article Advanced Crack Analytics on 3D X-ray Tomography of Irradiated Silicon Carbide Claddings Get access Fei Xu, Fei Xu Idaho National Laboratory, Idaho Falls, ID, United States Corresponding author: fei.xu@inl.gov Search for other works by this author on: Oxford Academic Google Scholar Joshua J Kane, Joshua J Kane Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Peng Xu, Peng Xu Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Nikolaus Cordes, Nikolaus Cordes Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Jason L Schulthess, Jason L Schulthess Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Mahmut Nedim Cinbiz, Mahmut Nedim Cinbiz Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Sean Gonderman, Sean Gonderman General Atomics, San Diego, CA, United States Search for other works by this author on: Oxford Academic Google Scholar Christian Deck, Christian Deck General Atomics, San Diego, CA, United States Search for other works by this author on: Oxford Academic Google Scholar Jack Gazza Jack Gazza General Atomics, San Diego, CA, United States Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 28, Issue S1, 1 August 2022, Pages 208–210, https://doi.org/10.1017/S1431927622001696 Published: 01 August 2022

Due to their high resistance to radiation damage and elevated temperature, silicon carbide fiber-reinforced, silicon carbide matrix composites (SiCf-SiCm) are identified as potential cladding structures for use in nuclear reactors. In this study, four composite architectures with varying ply numbers along the thickness direction and different biaxial or triaxial plain weave orientations at either 45° or 60°, were systematically evaluated under various stress states to assess the influence of weave architecture on mechanical performance. Experiments were conducted on SiCf-SiCm composite tubes under tensile hoop, axial compression, and rotating flexural loading to evaluate the mechanical response and investigate the failure modes using high-speed imaging and digital image correlation (DIC) techniques. It was found that for tensile hoop burst and flexural loading, the braiding angle had the most significant influence on the strength of the composite, whereas the effect of fiber angle was more limited for compression testing. Under axial compression a unique failure mode where a microcrack nucleates and grows only to a length equal to the thickness of a single yarn was identified. This crack growth behavior is reflected as periodic oscillations in the load-displacement response. For both axial and hoop loading, regardless of weave angle and number of plies, failure always initiated parallel to the tube axis in a single yarn and the cumulative interaction of these microcracks lead to either axial burst or fracture at an angle to the tube axis along a yarn. These results point to the importance of customizing the design of tube architecture for enhanced performance in specified nuclear applications.

Silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composites are candidate materials for cladding of light water reactor (LWR) fuels. Loss of fission product gas retention due to the formation of microcrack networks is considered a potential failure mechanism for SiC/SiC-cladded fuels. In this study, a variety of SiC/SiC composite tubes were irradiated with and without an LWR-relevant radial heat flux in the High Flux Isotope Reactor, followed by detailed characterization with X-ray computed tomography (XCT). This first set of XCT data for neutron-irradiated samples confirmed that the internal stresses arising from a combination of temperature gradients and irradiation-induced swelling act as the primary driver for cracking. While the observed cracking patterns varied depending on the tube architectures, the sharp edges of relatively large pores were found to be the common stress concentrator. These findings are useful to help improve the design and manufacturing of SiC/SiC fuel claddings for reduced failure probability.

Cladding thermal conductivity is an important physical property in assessing the performance of silicon carbide (SiC)-cladded fuels for nuclear reactors. However, there is a significant lack of reliable data, particularly for irradiated materials, because the geometry complicates the measurement. This study investigates the thermal diffusivity of coupons with a curvature, machined from SiC fiber–reinforced SiC matrix composite tubes, with and without neutron irradiation under light water reactor–relevant temperature and dose conditions. The tested materials included full composite and duplex SiC composite tubes. The measurements were conducted using a modern flash diffusivity apparatus. The analyzed area on the specimen during diffusivity testing was reduced for improved measurement accuracy due to sample curvature. Post-irradiation measurements showed that the effects of neutron irradiation on thermal conductivity (e.g. thermal defect resistivity) are different between SiC composite plates versus tubes. The difference was explained by higher matrix density of the tube than the plate. This study provides reliable thermal properties of prototypic SiC composite tubes useful for fuel performance modeling of SiC-based cladding.

Here, we unveil a methodology for a novel assessment of the fracture mechanics of SiC/SiC ceramic matrix composites enabled by in situ stereoscopic digital image correlation to quantify in-process flexural strain and crack opening displacement measurements. This technique isolates individual cracks on the composite surface as discontinuities in the spatial displacement field and correlates key fracture characteristics with the flexural strain of composite specimens during coupled four-point bend / hermeticity testing. Fracture was observed along the specimen length, originating at the tensile underside and propagating around the circumference of the tubular specimens with generally uniform spacing. Multiple specimens were also tested after heat treatments to 1200 degrees C in open air, in vacuum, and in helium for 48 h to evaluate the environmental effects on the fracture mechanisms of SiC/SiC composites, which revealed degradation of flexural properties after treatment in open air resulting in brittle failure. Indentation-based fracture toughness measurements were performed, which confirmed a 25% reduction in toughness after open air heat treatment relative to the other heat treatments. This assessment indicated that significant oxidation may occur within the composites from these heat treatments and suggested that further protection of the composites may be necessary for high-temperature applications.

and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division. A complete fabrication package record is maintained by THIEG and is available upon request.

Plain weave planar and biaxially braided tubular SiC/SiC CMCs are evaluated in tension and four-point bending, respectively, at ambient conditions. Custom-designed fixtures for CMC testing are developed for each loading mode and are coupled with three-dimensional digital image correlation. Stereoscopic image correlation analysis reveals crack initiation and failure sites to provide insight into stress redistribution mechanisms. Scanning electron microscopy is performed postmortem to determine the influence of microstructural features on crack initiation and failure. Crack spacing is measured in situ by stereoscopic image correlation and confirmed by SEM measurements to relate to underlying tow-tow crossing points. Triangulated surface heights of plain weave tow architecture are used to determine that subtle differences in neighboring transverse tow angle, which vary within a range of +/- 4 degrees from horizontal, have no significant effect on final fracture location. The results presented reaffirm the state of current SiC/SiC CMCs developed for energy applications and will help to further improve SiC/SiC and other CMCs.

Silicon carbide (SiC) fiber-reinforced ceramic matrix (SiCf/SiCm) composites have been identified as potential materials for nuclear fuel cladding. These composites are fabricated by braiding the SiC fibers around a mandrel into a cylindrical tube and then densifying this preform with SiC matrix using chemical vapor infiltration (CVI). However, during this fabrication process residual stresses develop in the composite, either due to the mechanical braiding process or due to high temperature CVI deposition of the SiC matrix phase. Micro-Raman spectroscopy has been employed to measure the residual stress in the SiC fibers at various stages of the fabrication process. Samples of the composite were analyzed after mechanical braiding and at 33%, 66% and 100% CVI densification of the composites. Raman spectra obtained at the above stages revealed significant band shifts of the SiC peaks and these peak-shift corresponded to -716 MPa of residual stress following the braiding process and -1075 MPa after the densification process. Interestingly, the Raman spectra also revealed carbon bands whose origin was traced to the presence of nanoscale turbostratic graphite throughout the nanograined SiC microstructure. (C) 2021 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

The results of high-temperature nanoindentation testing on both a control and a neutron-irradiated silicon carbide matrix silicon carbide fiber composite sample are presented. The mechanical properties of the chemical vapor-infiltrated matrix were observed to have slightly increased in hardness and slightly decreased in elastic modulus after irradiation. Tyranno SA3 fiber behavior results are inconclusive, possibly because residual graphite in the fibers resulting from the manufacturing process produced a large scatter in the data. This work also demonstrates the capability to measure the individual components of fabricated composites at elevated temperature, which should provide inputs for modeling the macro-scale behavior of the composites.

Silicon carbide matrix reinforced with silicon carbide fiber (SiC-SiC) composite is an advantageous material for many high temperature applications due to retention of room temperature strength above 1100 degrees C, which is the current operational limit for many high temperature metals. For the nuclear industry, this property, along with excellent radiation tolerance, has made SiC-SiC an attractive cladding material for the accident tolerant fuel program. The mechanical strength at high temperatures of these composites needs to be evaluated to address improvements in the third generation of Nicalon fibers, fabrication processes, as well as tested in a form-factor relevant to fuel cladding. In this work, mechanical strength of SiGA (TM) composite tubes, a SiC-SiC composite fabricated at General Atomics using the current generation of Hi-Nicalon Type S SiC fiber and chemical vapor infiltration processing (CVI) of the matrix, was evaluated at temperatures up to 1100 degrees C in ambient air and similar to 1900 degrees C in an inert gas environment with nominal 30-60 min dwell time. The results of this study show full retention of room temperature strength at both test regimes. Some degradation of material toughness was observed at temperatures above similar to 1760 degrees C. During this study, effects of interphase, fiber, and matrix morphological changes as a function of temperature were investigated. (C) 2019 Published by Elsevier B.V.