查看更多>>摘要：V-4Cr-4Ti-xY alloys with different yttrium (Y) contents (x = 0, 0.1, 1 wt. %) were prepared and irradiated with Fe2+ at 550 degrees C to the peak damage dose of 35 dpa (displacement per atom). The microstructure and hardness of the alloys before and after irradiation were obtained with transmission electron microscopy (TEM), scanning electron microscopy (SEM) and nano-indentation test. With the increase of Y contents, the grain size decreased monotonously, and the formation of needle-like Ti-rich precipitates was sup -pressed. After irradiation, voids were not observed in Y added alloys but in the V-4Cr-4Ti alloy, showing that the addition of Y could inhibit the formation of voids and irradiation-induced swelling. Dislocation loops and obvious irradiation hardening were observed in all V-4Cr-4Ti-xY alloys, and the size of the dis-location loops and the irradiation hardness increment decreased with the increase of the Y content. The irradiation-induced growth of Ti-rich precipitates was observed, which may be an important factor con-tributing to the irradiation hardening besides the dislocation loops and voids. The possible mechanism of the effect of the Y element on the irradiation behavior was discussed. Among the three Y contents, V-4Cr-4Ti-1Y alloy exhibits the best performance in the suppression of the formation of Ti-rich precipitates and irradiation hardening. (C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：In recent years, high entropy alloys (HEAs) have attracted significant attention due to their excellent physical, chemical, mechanical properties, and good irradiation resistance, thus are considered as potential candidates for fission and fusion structural applications. CoCrFeMnNi HEA was irradiated by high intensity pulsed ion beam (HIPIB) to investigate the effects of thermal shock irradiation on its microstructure, surface morphology and mechanical properties. It was found that CoCrFeMnNi HEA maintained the face-centered cubic single-phase structure after HIPIB irradiation. The ion beam effect of irradiation produced numerous defects such as vacancies and stacking faults within the range of carbon ions. While the thermal effect reduced vacancy concentration beyond the ion range by promoting the recombination of vacancies with interstitials, and decreased the nano-hardness of CoCrFeMnNi HEA. The thermal effect and shock wave effect promoted the migration of vacancies and formed defects such as stacking faults etc. far beyond the ion range in CoCrFeMnNi HEA. Because of the very compositional complexity, the high-level chemical disorder and local lattice distortion of CoCrFeMnNi HEA, the lattice parameter was almost unchanged after HIPIB irradiation. Even if the temperature reached the melting point of CoCrFeMnNi HEA, there was no crack on the surface after surface remelting and rapid cooling. CoCrFeMnNi HEA showed good thermal stability and thermal shock irradiation resistance. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The objective of this study is to understand the thermal stability of microstructure and its effect on the creep behavior of additively-manufactured 316L stainless steel (AM 316L SS). Creep specimens were fabricated from rods printed by a laser powder bed fusion process. Six different heat treatments, namely 650 degrees C/1 h, 700 degrees C/1 h, 750 degrees C/1 h, 800 degrees C/1 h, 900 degrees C/1 h and 1050 degrees C/1 h were applied to the creep specimens. The heat-treated specimens were creep-tested under the same condition, 550 degrees C/275 MPa to evaluate the effect of post-build heat treatment on the creep behavior of AM 316L SS. In the temperature range of 650-750 degrees C, dislocation density within cells, and cell size and wall thickness were affected by the heat treatment, while elemental segregation at boundaries remained unchanged and Mn-enriched Si oxide particles remained stable. In the temperature range of 750-900 degrees C, concurrent changes were observed in dislocation cell structures, elemental segregation and oxide particles. The 1050 degrees C-heat treatment removed cell structures and boundary solute segregations, leading to formation of equiaxed grains; Mn-enriched Si oxide particles in the as-built specimen were replaced by more stable Mn-enriched Cr oxides. The creep life of AM 316L SS increased after the heat treatment at 650 degrees C and then decreased with increasing temperature up to 900 degrees C where cell/subgrain structures still existed. The creep rate and creep elongation followed an opposite trend. The creep rate can be correlated with the structural parameters of cell/subgrain size and cell wall/boundary thickness and can be rationalized by different strengthening mechanisms of dislocation cells and subgrains. The heat treatment affected strongly the primary and secondary creep but had a minimal effect on the tertiary creep of AM 316L SS.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：This work provides a simple analytical formula with a set of best-fit coefficients for evaluating formation free energies of prismatic dislocation loops as a function of temperature in body-centered-cubic iron. The parameterization is made based on the anisotropic elastic self-energies and corresponding non-linear core-energies of edge dislocations. The variation of these two energy terms with temperature is estimated by including temperature-dependent elastic stiffness tensors in the anisotropic elasticity calculation, and based on the stability of C15 clusters and dislocation loops, respectively. The new analytical formula explicitly contains temperature and is valid for temperatures up to 1100 K.(c) 2021 Karlsruhe Institute of Technology. Published by Elsevier B.V. All rights reserved.
查看更多>>摘要：Beryllium wastes are produced by nuclear industry. One way to manage them is their encapsulation in cements. The main risk of this conditioning is the aqueous corrosion, which leads to the hydrogen production and cracks causing a loss of radioactivity confinement. The corrosion can be limited by the formation of the hydroxide solid phase Be(OH)(2(s)). The stability domain of this phase was calculated in water as a function of the pH: M. Pourbaix has calculated a stability domain from 2.9 to 11.7 for a 10(-4) M beryllium concentration, while according to our calculation with more recent thermodynamic data, it is stable from 5.3 to 13.5. Based on Pourbaix results, beryllium cannot be conditioned in the mainly used cement for nuclear waste, Portland cement, while it is possible according to our calculations. Experimental measurements were achieved to select the data set most in agreement with the experimental observations. The beryllium reactivity has been examined in matrices having different pH pore solution: brushite cement (pH 1.75-6.44), magnesium phosphate cement (pH 5.6-8.4), calcium-sulfoaluminate cement (pH 10.9-12.3), Portland cement (pH 12.5-12.9) cements and activated slag (pH 12.9-13.8), by measuring the open circuit potential and by electrochemical impedance spectroscopy. The experimental results agree with the more recent thermodynamic data. Beryllium corrosion is too high in the brushite cement, leading to a high hydrogen production. This matrix can then not be envisaged for the conditioning of Be waste. If the beryllium is encapsulated in the activated slag, the highly alkalinity is too high in the early age, leading to a high aqueous corrosion. Activated slag are also not suitable for Be conditioning. The main conclusion of this paper is that beryllium can be encapsulated in safe conditions in Portland, magnesium phosphate and calcium sulfoaluminate cements. (C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The existence of hydrogen isotopes in tritium breeding blankets indicates important research value for the realization of tritium self-sufficiency in fusion. The positive states of deuterium trapped by different irradiation defects in LiAlO2 (one of the typical ternary solid tritium breeders) have previously been indicated as O-D vibration absorption peaks in situ FT-IR after deuterium ion irradiation. The present study focused on the possible non-O-D state of deuterium which could not be directly observed by FT-IR in irradiated LiAlO2. In situ FT-IR and thermal desorption spectroscopy (TDS) was performed to single-crystal LiAlO2 irradiated by 3 keV D-2(+) at 380 K, 440 K and 500 K respectively. The increase of the total intensity of O-D vibration absorption peaks with the release of deuterium as D-2 suggests the existence of non O-D state. The non-O-D state is considered to be related to the deuterium trapped by oxygen vacancy. According to the density functional theory calculation, the valence state of hydrogen trapped by the O vacancy in LiAlO2 is-0.865. The transitions of different states (positive versus negative, different positive states) with the irradiation temperature and their effect on the release of chemical forms of deuterium have also been discussed in this paper. All above will contribute to further understanding on existence and behavior of tritium in neutron-irradiated solid breeders. (C)& nbsp;2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：This study explores the evolution processes of initiation, migration, growth, annihilation, and aggregation for dislocation loops and dislocation complex formation in regions with and without pre-existing dislocations. In-situ TEM experimental observations are conducted for 508-III steel, commonly used as a pressure vessel steel in nuclear reactors, during 400 keV Fe + irradiation at 573 K. Loop initiation is always observed, even following the formation of dislocation complex. The average size and number density of dislocation loops, in regions both with and without pre-existing dislocations, are obtained as a function of the irradiation dose. At the initial stage of irradiation, pre-existing dislocations significantly influence the evolution of dislocation loops. However, with an increase in the irradiation dose, this influence gradually weakens. The in-situ TEM experimental observations of microstructure evolutions in this study are of significant value for furthering our understanding of irradiation damage to 508-III steel.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Cr-incorporated UN (Cr-UN) composite fuels with different Cr amounts up to 10 wt% are fabricated by spark plasma sintering, and their microstructure and phase heterogeneity are analyzed. Highly densified microstructure and homogeneous Cr distribution are identified for the Cr-incorporated UN pellets, demonstrating a liquid sintering characteristic with a Cr enriched phase on the UN grain boundaries. A ternary phase U2CrN3 forms in the Cr-UN matrix with Cr amounts of 5 wt% and 10 wt%. The SPS densified Cr-UN composite pellets display greatly-improved thermal conductivity and simultaneously high hardness and fracture toughness. The fracture toughness of the Cr-incorporated UN pellet is similar to 5.5 MPa-m(1/2), representing almost 130% enhancement as compared to that of monolithic UN sintered at the same condition. The Cr-incorporated UN pellets also display enhanced oxidation resistance as evidenced by increased onset temperature for oxidation to 450 ? for the 5 wt% Cr-UN. These results highlight that Cr additive and the formation of a ternary phase can be useful to improve the thermal-mechanical properties and oxidation resistance of UN fuels with well-maintained high fissile element density. (C)& nbsp;2021 Elsevier B.V. All rights reserved.
Tummalapalli, Murali KrishnaSzpunar, Jerzy A.Prasad, AnilMalakkal, Linu...
8页
查看更多>>摘要：Uranium dioxide (UO 2 ) -molybdenum (Mo) composites have been proposed as accident tolerant fuel candidates due to their high thermal conductivity, high melting point, high boiling point, and ability to have low thermal neutron absorption cross-section. It is therefore essential to evaluate the method of fabrication and thermophysical properties of this system. In this work, UO 2 -Mo samples with a Mo fraction of 10 Vol% and different densities were fabricated using spark plasma sintering (SPS). This method allows sintering at relatively low temperatures and within a short time. Experimental analysis was performed on UO 2 -Mo specimens to characterize thermal diffusivity and thermal conductivity as a function of temperature up to 900 degrees C. Thermophysical properties were analyzed for various densities, and grain sizes as a function of temperature. The grain size and microstructural differences related to porosity were also analyzed using the scanning electron microscopic (SEM) and electron backscatter diffraction (EBSD). (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The bulk and surface properties of ThO2 are studied computationally using density functional theory within the LSDA + U approach. The computational method is benchmarked against bulk lattice parameter and band gap, the best combined description of which is with the PBE functional and the Liechtenstein method with U = 7 and J = 1, and without spin-orbit coupling. The optimised computational settings are used to study the reaction of water with the {111} surface, which we find to be hydrophilic. Molec-ular adsorption of a single water molecule is energetically favourable, and dissociation of this water to form surface hydroxyl groups is barrierless. This wetted surface can aid further water adsorption, with an even lower free energy difference than the first adsorption, although dissociation of the second water is much harder than the first. Reaction of H +, H . radical or OH- groups with the wetted {111} surface has a substantial energy barrier, and reaction with H + is very sensitive to temperature. (C)& nbsp;2021 Elsevier B.V. All rights reserved.
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