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1.
Yue-Lin Liu 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2009,267(18):3193-3196
Using a first-principles computational tensile test (FPCTT), we have investigated the effect of helium (He) on the structure and bonding properties of tungsten (W), which is a promising plasma-facing material in nuclear fusion Tokamak. Density of states results reveal the underlying reason that the substitutional site for He is the most energetically favorable, while the tetrahedral interstitial site is more favorable than the octahedral interstitial one. The FPCTT shows that the ideal tensile strength is 29.1 GPa at the strain of 14% along the [0 0 1] direction for intrinsic W, while it decreases to 28.2 GPa at the same strain when one impurity He atom is introduced. A local bond-breaking region around He forms in the tensile process due to the presence of He, which suggests He will have a large effect on the bonding properties of W. 相似文献
2.
Lattice parameter, electronic structure, mechanical and thermodynamic properties of ThN are systematically studied using the projector-augmented-wave method and the generalized gradient approximation based on the density functional theory. The calculated electronic structure indicates the important contributions of Th 6d and 5f states to the Fermi-level electron occupation. Through Bader analysis it is found that the effective valencies in ThN can be represented as Th+1.82 N−1.82. Elastic constant calculations show that ThN is mechanically stable and elastically anisotropic. Furthermore, the melting curve of ThN is presented up to 120 GPa. Based on the phonon dispersion data, our calculated specific heat capacities including both lattice and conduction electron contributions agree well with experimental results in a wide range of temperature. 相似文献
3.
《Journal of Nuclear Materials》1987,151(1):1-9
A series of tensile and strain controlled low-cycle fatigue tests were conducted at temperatures ranging from RT to 900°C on a nickel-base heat-resistant alloy, Hastelloy XR-II, which is one of the candidate alloys for applications in the process heating high-temperature gas-cooled reactor (HTGR). Fatigue tests at room temperature and all tensile tests were conducted in air, while fatigue tests at and above 400°C were conducted in the simulated HTGR helium environment. In those tests the effect of test temperature on tensile and fatigue properties was investigated. The ductility minimum point was observed near 600° C, while tensile and fatigue strengths decreased with increasing test temperature. The fatigue lives estimated with the method proposed by Manson were compatible with the experimental results under the given conditions. For the specimens fatigued at and above 700°C, the percentage of the intergranular fracture mode gradually increased with increasing test temperature. 相似文献
4.
The stability and migration behavior of helium and self defects in vanadium and V-4Cr-4Ti alloy are studied by first-principles calculations. The tetrahedral site is found as the most stable configuration for interstitial He, followed by the octahedral and substitutional sites. Among the self defects, the monovacancy has lower formation energy (1.71 eV for V and 2.14 eV for V-4Cr-4Ti alloy) than the self interstitial ones. The migration energies for He hopping between the tetrahedral sites are 0.06 and 0.09 eV for vanadium and V-4Cr-4Ti alloy, respectively. Our calculations reveal strong repulsion between two interstitial He atoms and strong attraction between He and vacancy, suggesting that vacancy acts as a trapping site for He impurity and a seed for further bubble formation. 相似文献
5.
Toshio Kainuma Nobuhiko Iwao Tadashi Suzuki Ryoji Watanabe 《Journal of Nuclear Materials》1979,80(2):339-347
Effects of oxygen, nitrogen and carbon additions on the mechanical properties at room temperature of vanadium and V-Mo alloys containing up to 25 at% molybdenum were studied. The V-O, N or C and V-Mo-O, N or C alloy systems, respectively, were prepared by heating vanadium and the V-Mo alloys with oxygen, nitrogen or propane gas in sealed quartz capsules. The prepared alloys were homogenized, argon-quenched and analyzed for oxygen, nitrogen and carbon content. Then they were examined by tensile and hardness tests at room temperature, SEM observations and TEM studies. Oxygen and nitrogen additions to vanadium and the V-Mo alloys raise the hardness by solid solution, with nitrogen being more effective. Carbon additions to vanadium form coarse precipatates of V2C along the grain boundaries, which do not raise the hardness, while those to the V-Mo alloys form fine precipitates of vanadium carbide, homogeneously distributed in the matrix, which raise the hardness. Threshold concentrations of oxygen, nitrogen and carbon for embrittlement (fracture without plastic deformation) of these alloy systems decrease with increasing molybdenum concentration, those of carbon being most detrimental. Also, it was shown that the embrittlement of the V-Mo alloys after exposure to liquid sodium may be explained not only by solid solution hardening due to absorbed oxygen but also by combined effects of absorbed oxygen, nitrogen and carbon. 相似文献
6.
Plutonium dioxide (PuO2) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO2 at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO2 were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO2 were investigated up to 1500 K. 相似文献
7.
First-principles calculations based on density functional theory have been performed to investigate the behaviors of He in hcp-type Ti. The most favorable interstitial site for He is not an ordinary octahedral or tetrahedral site, but a novel interstitial site (called FC) with a formation energy as low as 2.67 eV, locating the center of the face shared by two adjacent octahedrons. The origin was further analyzed by composition of formation energy of interstitial He defects and charge density of defect-free hcp Ti. It has also been found that an interstitial He atom can easily migrate along 〈0 0 1〉 direction with an activation energy of 0.34 eV and be trapped by another interstitial He atom with a high binding energy of 0.66 eV. In addition, the small He clusters with/without Ti vacancy have been compared in details and the formation energies of HenV clusters with a pre-existing Ti vacancy are even higher than those of Hen clusters until n ? 3. 相似文献
8.
《Fusion Engineering and Design》2014,89(2):137-141
An oxide dispersion strengthened ferritic steel with a nominal composition of Fe–14Cr–2W–0.3Ti–0.3Y2O3 (in wt.%) was consolidated by hot isostatic pressing at 1150 °C under various pressures in the range of 185–300 MPa for 3 h. The microstructure, microhardness and high temperature tensile properties of the steel were investigated. With increasing compaction pressure the density of specimens also increased, however OM and SEM observations revealed residual porosity in all tested specimens and similar ferritic microstructure with bimodal-like grains and numerous of large oxide particles, located at the grain boundaries. Mechanical testing revealed that compaction pressure has negligible influence on the hardness and tensile strength of the ODS steel, however improves the material ductility. 相似文献
9.
This paper presents experimental results concerning the tensile properties of JIS Type SUS 316 stainless steel. The test was carried out at room temperature, 400°C and 550°C at strain rates of 10−3 1/s and 102 1/s. Base metal, weld joint and weld metal specimens were chosen for the test. The aim of this test is to clarify the effects of strain rate and test temperature on the mechanical properties such as 0.2% yield strength, ultimate tensile strength and elongation of JIS Type SUS 316 stainless steel. 相似文献
10.
Normalized-and-tempered 9 Cr-1 MoVNb steel tensile specimens were irradiated in the Experimental Breeder Reactor-11 (EBR-11) at 390, 450, 500, and 550°C to ~2.1 and 2.5 × 1026 neutrons/m2 (> 0.1 MeV), which produced displacement damage levels of ~10 and 12 dpa, respectively. Tensile tests were conducted at the irradiation temperature and at room temperature. In addition to the irradiated specimens, as-heat-treated specimens and as-heat-treated specimens thermally aged at the irradiation for 5000 h were also tested.Thermal aging had no effect on the unirradiated tensile properties. Irradiation at 390°C increased the 0.2% yield stress and the ultimate tensile strength above those of the unirradiated control specimens. The ductility decreased slightly. After irradiation at 450, 500, and 550°C, the tensile properties were essentially the same as the unirradiated values. The hardening at 390°C was attributed to the dislocation and precipitate structure formed during the irradiation. The lack of hardening at 450°C and higher correlates with an absence of an irradiation-induced damage structure. 相似文献
11.
The effect of neutron irradiation on the tensile properties of normalized-and-tempered Cr-1 Mo steel was determined for specimens irradiated in Experimental Breeder Reactor II (EBR-II) at 390 to 550°C. Two types of unirradiated control specimens were tested: as-heat-treated specimens and as-heat-treated specimens aged for 5000 h at the irradiation temperatures. Irradiation to approximately 9 dpa at 390° C increased the strength and decreased the ductility compared to the control specimens. Softening occurred in samples irradiated and tested at temperatures of 450, 500, and 550 °C; the amount of softening increased with increasing temperature. The tensile results were explained in terms of the displacement damage caused by the irradiation and changes in carbide precipitates that occur during elevated-temperature exposure. 相似文献
12.
The current program plan for the development of materials for fusion reactors requires testing candidate materials in both fission reactors and high energy neutron sources. Because of the volume limitations of available facilities, both current and near term, and because of the relatively large number of materials and test conditions that will need to be explored, it is essential that test techniques be developed to extract mechanical property information from small volume specimens. A variety of such test techniques are under development at the University of California, Santa Barbara. These include instrumented microhardness, bulge, shear punch, indentation creep and load relaxation and miniaturized fracture tests for obtaining strength, ductility, time-dependent flow, and fracture behavior information on specimens as small as TEM discs. 相似文献
13.
In this study, the thermal and mechanical characteristics are analyzed for the structural integrity evaluation of the instrumented capsule used for the irradiation test of reactor vessel materials in the research reactor, hi-flux advanced neutron application reactor (HANARO). The temperature of test specimens inserted in the capsule mainbody by γ-flux is calculated using a heat transfer code, HEATING 7.2f. The maximum temperature is 556.75 K at the center of the capsule mainbody, thus the temperature satisfies the user's requirement. To estimate the mechanical characteristics of the capsule due to the pressure and thermal loading, stress analysis is carried out with a finite element analysis program, ANSYS. The strength of the capsule's external tube is also evaluated by considering the buckling stress of the capsule mainbody under coolant pressure loading. The results of the analysis show that the temperature distributions are significantly affected by the gap size between the holder and the specimen. The calculated stresses of the capsule structure are well within the allowable stress values of the ASME code. It is expected that the results presented in this paper will be useful in the design and safety evaluation of instrumented capsules for material irradiation tests. 相似文献
14.
N. N. Davidenkov B. A. Sidorov L. M. Shestopalov N. F. Mironov N. M. Bogograd L. A. Izhvanov S. B. Kostogarov 《Atomic Energy》1965,18(6):768-776
Thanks to its high strength density ratio and good nucIear properties, beryllium is attracting increased attention from scientists as a construction material for use in aviation, rocket, and atomic technology. We here describe methods for mechanical (tensile and compression) tests in vacuum on beryllium and set out results for beryllium prepared by metal.powder and casting methods, measured at 20 to 1000° C.Translated from Atomnaya Énergiya, Vol. 18, No. 6, pp. 608–616, June, 1965 相似文献
15.
We have investigated permeation and transport of hydrogen (H) isotopes in tungsten (W) single crystal employing first-principles calculations in junction with Fick’ law. Permeability was approximately evaluated according to the solubility and diffusion coefficient of H. The solubility for H in bulk W from present calculation is consistent with the experimental results measured by Frauenfelder. The permeation fluxes of H isotopes are examined at the different thickness of W wall. The permeation fluxes of deuterium with the W thickness of 21 μm at the temperature of 770 K and with the W thickness of 50 μm at the temperature of 893 K were 0.68 × 1013 atom/m2s and 0.34 × 1014 atom/m2s, respectively. The dissociation coefficients of H isotopes are also evaluated. We believe that the present first-principles combined with Fick’ law method can be also generalized to investigate permeation and transport of H isotopes in most metals since such H isotopes behaviors in most metals are similar to those of H isotopes in W. 相似文献
16.
Meimei Li M. Eldrup T.S. Byun N. Hashimoto L.L. Snead S.J. Zinkle 《Journal of Nuclear Materials》2008,376(1):11-28
Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10−5 to 0.28 dpa at 80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10−5 to 1 × 10−2 s−1. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at 0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at 0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress. 相似文献
17.
Roman E. Voskoboinikov 《Journal of Nuclear Materials》1999,270(3):309-314
The paper deals with the crack nucleation and stability in strain fields of stress concentrators (e.g. voids, gas bubbles, secondary phase precipitates). A general equation describes critical and subcritical crack length as a function of external (applied loading) and internal (stress concentrator type, normal traction, elastic properties of matrix, etc.) parameters. For the critical crack an analog of the Griffith criterion is found. The reduction of fracture stress due to different types of internal stress concentrators was evaluated. 相似文献
18.
In terms of first-principles calculations, elastic stiffness constants Cij’s as well as the polycrystalline aggregates including the bulk, shear, Young’s moduli, Possion’s ratio, and anisotropy factors have been predicted for three technologically important polymorphs of ZrO2, i.e., monoclinic m-ZrO2, tetragonal t-ZrO2, and cubic c-ZrO2. Here, both the strain vs. stress (S-S) and the strain vs. strain energy (S-E) methods are adopted. In the first-principles calculations, both the local density approximation (LDA) and the generalized gradient approximation (GGA) are utilized. It is found that the more accurate structural and elastic properties are determined by LDA in comparison with experimental results and the S-S method is more effective than the S-E method although the two methods predict the similar results. The predicted negative values for C16, C36, and C45 of m-ZrO2 suggest that the certain normal or shear stress corresponds to an opposite shear strain for reducing the total energy. Small differences of shear and Young’s modulus between m-ZrO2 and t-ZrO2 suggest that their mechanical properties are comparable. 相似文献
19.
20.
The effect of heating rate, prior to tensile testing, on the elevated temperature flow stress of zircaloy-2 was investigated in the temperature interval 473 to 923 K. An increase in the heating time from 30 to 120 min caused a decrease in the 0.2% flow stress of approximately 4 MPa (3%) which is within the experimental error involved in the measurement of this property by a single test. Similarly, the holding time, prior to testing, was found to have only a minor influence on the flow stress. Recently published work has reported that thermal history had a much greater effect than this and reasons for the discrepancy are proposed. 相似文献