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In this study, a technique that allows electron backscatter diffraction (EBSD) observation in the same point of interest under uniaxial tensile testing, i.e., a quasi in situ EBSD analysis technique, was developed and performed to investigate the interaction between slip and { 10 1 ¯ 2 } tensile twinning in a Zr alloy. Both the slip-independent and slip-assisted { 10 1 ¯ 2 } twins were investigated. The former twin is formed when the global stress surpassed the required critical stress, and the latter one is formed due to the local stress concentration near the grain boundary accumulated in the soft-oriented adjacent grain, although the global stress remains insufficient to activate the twinning. The formation of a slip-independent or slip-assisted { 10 1 ¯ 2 } twin seems to be closely linked to the crystal orientation relative to the loading direction. It is due to the bimodal basal texture exhibited in the present alloy that the slip-assisted twins can be observed. Furthermore, compared to slip-independent twins, the slip-assisted twinning is more likely to form at a higher strain, because substantial strain accumulation is essential to trigger its activation.

The purpose of this study is to investigate the response of the precipitates against irradiation in F82H steels which are considered as main candidate structure material for the fusion DEMO reactor blanket. M23C6-type precipitates contained F82H-IEA steel, and MX- and M23C6-type precipitates contained F82H-BA12 steel were selected as experimental specimens. Iron irradiation at 673 K using ion accelerator was performed on both steels up to 50 dpa. After irradiation, the instability behavior of precipitates was characterized using transmission electron microscopy. It is found that in F82H-IEA specimens, the number density of radiation defects at 20 dpa was 3.26 × 1021 m−3, and the amorphization of M23C6 was undetected up to 50 dpa. Although Ostwald ripening occurred in M23C6 precipitates, the average size of M23C6 mildly decreased with irradiation dose within this study. For F82H-BA12 specimens, the shrinkage rate of MX- and M23C6-type precipitates was estimated respectively as 0.06 and 0.03 nm/dpa. Hence, the MX and M23C6-type precipitates in F82H steel exhibit an excellent irradiation resistance under the present irradiation condition. Moreover, the behaviors like amorphization and weakening in contrast were not observed in M23C6 precipitates in both specimens, which is different from electron irradiation, presumably due to the factors such as different damage rate. Keywords: F82H, Ion irradiation, Precipitates, Instability

The purpose of this study is to investigate the response of the precipitates against irradiation in F82H steels which are considered as main candidate structure material for the fusion DEMO reactor blanket. M23C6-type precipitates contained F82H-IEA steel, and MX- and M23C6-type precipitates contained F82H-BA12 steel were selected as experimental specimens. Iron irradiation at 673 K using ion accelerator was performed on both steels up to 50 dpa. After irradiation, the instability behavior of precipitates was characterized using transmission electron microscopy. It is found that in F82H-IEA specimens, the number density of radiation defects at 20 dpa was 3.26×1021 m−3, and the amorphization of M23C6 was undetected up to 50 dpa. Although Ostwald ripening occurred in M23C6 precipitates, the average size of M23C6 mildly decreased with irradiation dose within this study. For F82H-BA12 specimens, the shrinkage rate of MX- and M23C6-type precipitates was estimated respectively as 0.06 and 0.03 nm/dpa. Hence, the MX and M23C6-type precipitates in F82H steel exhibit an excellent irradiation resistance under the present irradiation condition. Moreover, the behaviors like amorphization and weakening in contrast were not observed in M23C6 precipitates in both specimens, which is different from electron irradiation, presumably due to the factors such as different damage rate.

In this study, the mechanical properties in a Zr-Nb alloy plate with bimodal basal texture microstructure have been evaluated by tensile, Vickers hardness and nano-indentation tests. Based on the results, the strength-hardness relationship is discussed with special attention paid to its textured microstructure. It is found that the correlation between strength and hardness varied greatly depending on whether the loading direction is parallel or perpendicular to the normal direction of the plate, demonstrating that the strength-hardness relationship is texture-dependent. In addition, the relation between strength and nano-hardness is obtained by subjecting specimens to cold rolling at various reduction rates. This allows for the evaluation of the strengthening effect using convenient and non-destructive nano-indentation tests instead of elaborate tensile tests, which may have particularly useful applications for nuclear engineering components when considering the irradiation strengthening occurred in these materials.

In order to clarify the instability of M23C6 in F82H steel under irradiation, both electron irradiation using a high voltage electron microscope (HVEM) and ion irradiation using an ion accelerator were performed. For the electron irradiation, in-situ observation under 2 MV electron irradiation and ex-situ high resolution electron microscopic (HREM) analysis were utilized to evaluate the response of M23C6 against irradiation. The temperature dependence of the irradiation induced instability of the carbide was first confirmed: 293 K < T < 573 K, by observation of lowering in contrast at the periphery of carbides, 698 K < T < 773 K, fragmentation at the interface between carbides and matrix, and at 773 K, formation and coarsening of new particles near the periphery of M23C6. HREM analysis showed the loss of the lattice fringe contrast at the pre-existing M23C6 precipitates at temperatures ranging from 473 to 773 K, indicating severe loss of crystallinity due to dissolution of the constituent atoms though irradiation- enhanced diffusion under the vacancy diffusion by the focused electron beam irradiation. For the ion irradiation, 10.5MeV-Fe3+ ion was applied to bombard the F82H steel at 673K to achieve the displacement damage of 20 dpa at the depth of 1.0 µm from surface. Cross-section TEM specimens were prepared by a focused ion beam technique. The shrinkage of carbide particles was observed especially near the irradiation surface. Besides, the lattice fringes at the periphery of carbide were observed in the irradiated M23C6 by the HREM analysis, which is different from that observed in the electron irradiation. It was clarified that the instability of M23C6 is dependent on the irradiation conditions, indicating that the flow rate of vacancy type defects might be the key factor to cause the dissolution of constituent atoms of carbide particles into matrix under irradiation.

Tensile properties and microstructure of a Zr–1.8Nb alloy subjected to 140-MeV C 4+ ion irradiation at 573 K up to 5.3 dpa have been investigated by using 0.18 mm-thick tensile specimens. Irradiation damage was introduced into the tensile specimens homogeneously over entire thickness using a variable energy degrader. Irradiation-induced strengthening and embrittlement (loss of ductility) were successfully evaluated. The yield strength of irradiated specimens (3.1- and 5.3-dpa) was 137% and 145% of unirradiated specimen, whereas the total elongation of those irradiated specimens was less than a half of unirradiated specimen. It appears that the rate of embrittlement is fairly faster than the rate of strengthening. Transmission electron microscopy (TEM) observation on the 3.1-dpa specimen revealed that the bcc-Nb/Zr precipitates originally contained in the alloy were stable in terms of crystal structure and size. The size and number density of dislocation loops were 7.2 nm and 1.2 × 10 21 /m 3 for the -loops, 15.4 nm and 6.5 × 10 20 /m 3 for the -type loops, respectively. The increase of yield strength expected from these dislocation loops was ∼64% of the actual irradiation-induced strengthening, indicating that the dislocation loops are the main contributors for the irradiation-induced strengthening.

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

In this work, the mechanical behavior of as-received and hydrogenated Zircaloy-4 fuel claddings was investigated by the newly developed advanced expansion due to compression (A-EDC) test and the conventional uniaxial tension (UT) test at room temperature, in order to, respectively, understand the hydride-induced embrittlement in tube longitudinal and hoop directions. The UT experimental results showed that the mechanical strength in the longitudinal direction slightly increased with hydrogen content, whereas the maximum strain decreased greatly with hydrogen increasing. In the case of A-EDC tests, the mechanical performance in the hoop direction seemed insensitive to the hydrogen content; no obvious decline in maximum strain was observed until 800 ppm H. The comparison between these two tests clearly reveals that the hydride-induced embrittlement is preferential to occur in the longitudinal direction, compared with the sluggish response in the hoop direction, which implied the enhanced ductility a...

The microstructure of the 12Cr-ODS steel was characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), transmission electron microscopy (TEM), and electron backscatter diffraction (EBSD) techniques. The results showed that the microstructure consisted of fine and elongated grains, high density of dislocations, and large concentration of nm-scale oxide particles, which were identified as five types of crystal structures by high resolution TEM (HRTEM) imaging, such as monoclinic and cubic Y 2 O 3 , hexagonal and orthorhombic Y 2 TiO 5 , and cubic Y 2 Ti 2 O 7 . The dominant {001} component and weaker parallel to normal direction texture were observed. Besides, the yield stress was experimentally measured and quantitatively estimated. The findings indicated that theoretical calculation was in accordance with the experimental measurements, and strengthening contributions from solid-solution atoms and grain boundaries were linearly additive, whereas those from dislocations and oxide dispersoids were averaged by root mean square summation.

The effect of Mo addition on microstructures of β-quenched and subsequently annealed Zr-1.2Nb alloys is systematically investigated. Differential scanning calorimetry, electron backscatter diffraction, transmission electron microscopy, X-ray energy dispersive spectrometry techniques and microhardness tests are jointly utilized to characterize the microstructure and to evaluate the mechanical property. Martensitic microstructure, a mixture of lath martensites and lens martensites, is observed in the β-quenched specimens. Width of lath martensite reduces and fraction of lens martensite increases as the Mo content increases. This suggests that Mo addition retards slip deformation and enhances twinning deformation during the martensitic transformation, and that segregation of Mo atoms is also responsible for the lath width reduction. After annealing at 873 K, the martensitic structures are maintained, accompanying with the presence of numerous precipitates, which is attributed to the fact that the β-quenched microstructure is strain-relieved instead of being recrystallized. The distribution of the misorientation angle between the neighboring grains in both β-quenched and subsequently α-annealed specimens is predominantly around 60° regardless of the Mo content. Microhardness increases with increasing Mo content both in the β-quenched and the annealed specimens, and it drops greatly after annealing at a fixed composition. Keywords: Zr-Nb alloy, Zr-Nb-Mo, Martensite, Mo addition, Twinning, Precipitates

A potential factor dominating the obstacle strength of second phase precipitate particles in dispersion strengthening is the crystallographic mismatch between the matrix phase and the second phase; however, yet this concept has not been fully assessed by experiments and simulations. In the present study, we experimentally investigated the obstacle strength of body centered cubic (bcc) Nb particles and nanometric Nb clusters embedded in hexagonal close packed (hcp) Zr matrix. The bcc Nb is softer than the hcp Zr in terms of shear modulus, whereas from a crystallographic viewpoint, the bcc Nb particles can be nonshearable, strong obstacles because the slip plane inside the particles is not parallel with that in the matrix. Although the bcc Nb is thermodynamically the stable configuration for Nb atoms precipitating from the Zr matrix, in the very early stage of solute agglomeration, the crystal structure of Nb nanoclusters is possibly hcp rather than bcc. The obstacle strength ( α ) was no greater than 0.5 for the Nb nanoclusters, whereas 0.85 ≤ α ≤ 1 for the coarse bcc Nb particles; α = 1 was obtained with the Taylor factor ( M ) of 5.5 and α = 0.85 with M = 6.5, respectively. These results indicate that the bcc Nb particles are strong obstacles, and that the Nb nanoclusters are weak obstacles.

The alloying effects of Sn, Nb, Cr, and Mo on zirconium alloys were elucidated and compared. Electron backscatter diffraction, transmission electron microscopy, tensile test, and fractographic observation were jointly utilized to carry out detailed microstructural characterization and mechanical property evaluation. Results show that Mo is the most effective among these elements from the viewpoints of strengthening and reducing grain size. The strengthening mechanism for each element is also discussed. The order of solid-solution strengthening of these alloying elements is Cr > Nb > Sn, and the sequence is Cr ≈ Mo > Nb when precipitation strengthening is considered. Further, as far as the ability to impede dislocation motion is concerned, the sequence is Mo > Cr > Nb > Sn. The experimental results demonstrate that minor amount of Mo addition in zirconium alloys is greatly effective in strengthening the alloy and reducing the grain size.