查看更多>>摘要：Potassium-doped tungsten (KW) plate with a weight of 25 kg was prepared by hot rolling. Subsequently, part of the KW suffered annealing at 160 0, 180 0 degrees C for 0.5 h. For the rolled state, K bubble showed the size and number density of 71 nm, 7.6 x 10(18) m(-3) in the grain interior and 98 nm, 1.9 x 10(19) m(-3) at the grain boundaries. To figure out the influence of K bubbles on recrystallization resistance and mechanical properties, microstructure and tensile performance were examined on the stress relieved and high-temperature annealed KW. The results indicated that the KW displayed excellent recrystallization resistance as the K bubble strings inhibited the transverse movement of grain boundaries. Tensile tests at 100, 150, 200, 250, 300, 500 and 700 degrees C showed the DBTT was 50-100 degrees C for the stress relived KW and about 200 degrees C for the high-temperature annealed KW. Moreover, the KW plate exhibited large ductility as K bubbles in the grain interior can act as the source to emit dislocation and barrier to dislocation movement thereby lead to lots of mobile dislocation. Besides, the high number density of nano-sized K bubbles also improved the KW strength, and the strength contribution of K bubbles calculated by the Orowan mechanism was 245.6 MP at room temperature. (C) 2021 Elsevier B.V. All rights reserved.
Clay, Benjamin T.Donnelly, Stephen E.Greaves, Graeme
9页
查看更多>>摘要：Polycrystalline 3C SiC was irradiated and observed in-situ via Transmission Electron Microscopy with a 20 keV He ion beam at 40 0, 80 0, 10 0 0 and 120 0 degrees C at the Microscopes and Ion Accelerators for Materials Investigations facility. During the 40 0, 80 0, 10 0 0 and 120 0 degrees C irradiations, black-spot damage was observed at 3.1, 1.1, 2.1 and 2.1 dpa respectively. Helium bubbles were observed after 6.3 dpa at 400 degrees C and 2.1 dpa at 800 degrees C, and He platelets were seen after 1.1 dpa at 80 0, 10 0 0 and 120 0 degrees C but not observed during the 400 degrees C irradiation. This work shows for the first time, the preferential nucleation of platelets within stacking faults in 3C SiC. The dependence of He platelet diameter with temperature and dose has also been observed.(c) 2021 Elsevier B.V. All rights reserved.
Evans, Jordan A.Puente, Ashley E. Paz Y.Robinson, Adam B.Glagolenko, Irina Y....
12页
查看更多>>摘要：The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel's operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed incomplete bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB(2 & nbsp;)in its as-rolled condition. The porosity in the ZrB(2)foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.(C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：This paper presents a theoretical approach addressing plastic-strain spreading in post-irradiated BCC materials accounting for crucial sub-grain scale, dislocation-mediated plasticity mechanisms. The proposed model explicitly provides the number of shear-bands developed in irradiated (N-irr) versus non-irradiated (N-00dpa) grain cases, for fixed amounts of plastic deformation. Calculations carried out under various irradiation defect size and number density cases, which helps it appraising important material properties, in particular the dose-dependent, grain-scale uniform elongation threshold. The model ability to handle macro-scale effects is then evaluated using a simple stochastic calculation procedure, taking advantage of actual grain size and orientation maps. The dose-dependent embrittlement amplitude appears to critically depend on the shear band thickness and spacing variations, existing near the fracture surface of failing specimens. That perception allows comparing our predictions with adapted test results, for validation. (c) 2021 The Authors. Published by Elsevier B.V.& nbsp;
查看更多>>摘要：Pebbles made of lithium-containing ceramics with a few millimeters in diameter have been regarded as the main candidates for the tritium breeding materials. In this work, X-ray computed tomography (Xray CT) technique has been employed to observe the microstructure of tritium breeding pebbles. The 3D pore network of the pebbles was presented and compared with normal 2D images. Based on the CT results, simulations of the heat transfer in one single pebble have been performed employing lattice Boltzmann method and showed that the gas atmospheres and microstructure have apparent influences on the thermal conductivities of pebbles.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Available publications on crystalline materials used to immobilize HLW and actinides have been analyzed. Select compounds and solid solutions covered in this paper are phosphates of the following structural families: monazite, xenotime, kosnarite (NZP),apatite and britholite, langbeinite, whitlockite and thorium phosphate diphosphate (TPD). These materials have been shown to be fit for "nature-like" disposal (ex-cept TPD), they have a wide spectrum of isomorphic substitutions of cations and/or anions, and high heat, chemical, and irradiation stability. They can be also used in geoenvironmental engineering. The studies taken into consideration follow the materials science approach that rests on the "composition - structure - synthesis - property" basis. The analyzed papers describe materials synthesized using Spark Plasma Sintering, a process that results in rapid formation of virtually poreless ceramics and improves environ-mental safety both at stage of HLW immobilization and during long-term storage.(c) 2021 Published by Elsevier B.V.
查看更多>>摘要：To estimate the effect of displacement damage by neutron irradiation of an oxide film of zirconium alloy, a Zircaloy-2 sample was corroded in high temperature and high pressure water at 561 K for 10 0 0 h to form an oxide film. It was then irradiated with 3-MeV Zr2+ ions up to 1.3 x 10(20) ions/m(2) at 573 K. Subsequently, the crystalline properties of the oxide film and nature of secondary phase particles (SPPs) in the oxide film were investigated by X-ray diffraction (XRD), transmission electron microscopy (TEM), scanning electron microscopy (SEM), and secondary-ion mass spectrometry (SIMS). From the results off the XRD measurement, monoclinic-ZrO2 was found to be predominant at operation temperature of a light-water reactor, and slightly tetragonal ZrO2 was also confirmed. The diffraction peaks of the monoclinic ZrO2 were slightly clearer in the case of the irradiated sample than in the case of the unirradiated sample. Specifically, ratio of tetragonal ZrO2(011) diffraction to monoclinic ZrO2(-111) diffraction was increased by irradaiation. In addition, half width of monoclinic ZrO2(-111) diffraction was decreased with irradiation dose. This result suggests that the crystallinity of the oxide film was macroscopically improved by the ion irradiation. The diffracted wave corresponding to the (011) plane of the tetragonal ZrO(2 )was also revealed after irradiation. It may suggest that tetragonal ZrO2 is stabilized by accumulation of irradiation defects. However, the crystal structure difference between irradiated and unirradiated is small. Detailed TEM observation of the oxide film confirmed transition from crystalline to amorphous phase in the SPPs near the metal-oxide interface. Furthermore, the diffraction pattern taken from the SPPs near the surface of the oxide film could not be distinguished from that of ZrO2. That is, it was concluded that those near the surface of the oxide film were assimilated into ZrO2. Their iron concentration was less than that near the metal/oxide interface. Following the TEM observation, SEM observation was performed in the vicinity of the area where the TEM image was obtained. According to the SEM observation, the number of SPPs in the ion-irradiated oxide film tended to be lower than that in the unirradiated oxide film. This tendency was more significant near the surface than near the metal-oxide interface. It is therefore concluded that ZrO2 in the oxide film is not easily irradiated; however, the SPPs are affected by the irradiation, and dissolution of the alloying elements from the SPPs to the matrix of the oxide film is promoted by substitution irradiation damage. (C)& nbsp;2021 Published by Elsevier B.V.
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