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Silicon carbide (SiC) ceramic matrix composite (CMC) cladding is currently being pursued as one of the leading candidates for accident-tolerant fuel (ATF) cladding for light water reactor applications. The morphology of fabrication defects, including the size and shape of voids, is one of the key challenges that impacts cladding performance and guarantees reactor safety. Therefore, quantification of defects' size, location, distribution, and leak paths is critical to determining SiC CMC in-core performance. This research aims to provide quantitative insight into the defect's distribution under multi-scale characterization at different length scales before and after different Transient Reactor Test Facility (TREAT) irradiation tests. A non-destructive multi-scale evaluation of irradiated SiC will help to assess critical microstructural defects from production and/or experimental testing to better understand and predict overall cladding performance. X-ray computed tomography (XCT), a non-destructive, data-rich characterization technique, is combined with lower length scale electronic microscopic characterization, which provides microscale morphology and structural characterization. This paper discusses a fully automatic workflow to detect and analyze SiC-SiC defects using image processing techniques on 3D X-ray images. Following the XCT data analysis, advanced characterizations from focused ion beam (FIB) and transmission electron microscopy (TEM) were conducted to verify the findings from the XCT data, especially quantitative results from local nano-scale TEM 3D tomography data, which were utilized to complement the 3D XCT results. In this work, three SiC samples (two irradiated and one unirradiated) provided by General Atomics are investigated. The irradiated samples were irradiated in a way that was expected to induce cracking, and indeed, the automated workflow developed in this work was able to successfully identify and characterize the defects formation in the irradiated samples while detecting no observed cracking in the unirradiated sample. These results demonstrate the value of automated XCT tools to better understand the damage and damage propagation in SiC-SiC structures for nuclear applications. [ABSTRACT FROM AUTHOR]

Abstract This paper presents results of studies characterizing the mechanical properties of irradiated U 10Mo fuel in support of low-enriched fuel conversion of several United States high-performance research reactors (USHPRR), specifically the monolithic base fuel qualification. Specimens were tested from ten test fuel plates irradiated at the Advanced Test Reactor (ATR) from the RERTR-12 and AFIP-6 Mk. II irradiation campaigns. Tests were conducted at Idaho National Laboratory (INL) in the Hot Fuel Examination Facility (HFEF) inert atmosphere Main Cell. The monolithic fuel plates consist of a U 10Mo fuel foil meat covered with a Zr diffusion barrier layer fabricated by co-rolling, and clad in 6061 Al using a hot isostatic press (HIP) bonding process. Initial unirradiated specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and irradiated (as tested) fuel plate average burnup ranged from approximately 0.36 × 1021 to 6.2 × 1021 fissions/cm3. After sectioning individual specimens from the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre-dissolution measurement of width and post-dissolution measurement of thickness were conducted on test specimens to facilitate accurate analysis of bend test results. The final thickness values varied from 0.3 to 0.8 mm. Four-point bend testing was conducted on the HFEF Remote Load Frame (RLF) using custom-designed test fixtures and calibrated force and displacement sensors. All specimens exhibited dominantly elastic behavior and failed in a brittle manner. Analysis of test results showed decreasing elastic modulus and bending failure strength with increasing fission density. The trend of modulus decrease intersected with unirradiated modulus at zero fission density. The trend of decreasing bending strength with increased fission density is apparent, but has significantly higher uncertainty levels than the modulus results over the fission density range evaluated. A linear trend fit of the calculated bend strengths versus fission density does not coincide with unirradiated tensile strength, which is significantly higher. Porosities measured on the test specimen material are used with several model equations to predict irradiated modulus values. Over the range of available porosity information, comparison to the experimental modulus values are favorable. Other factors and their potential influence on modulus and strength calculations are discussed. [ABSTRACT FROM AUTHOR]