查看更多>>摘要：It is widely accepted that the characteristics of hydrides in the Zr-based nuclear fuel cladding are critical factors determining the integrity of the spent fuel. Therefore, to mitigate any harm caused by them, we need to gain a deep understanding of the underlying mechanical behavior involving the precipitated hydrides. Using the molecular dynamics (MD) simulation method, we investigated the deformation mechanisms of embedded hydride in the polycrystalline zirconium matrix under uniaxial tensile loading. The plasticity of hydridized zirconium is controlled by the slip transmission from the metallic matrix into the embedded hydride, for which the considerable stress concentration at the metal-hydride interface needs to precede. The proposed mechanism manifests the size-dependent plasticity in the hydride, where the morphology of embedded hydride plays an important role in allowing the slip transmission through the matrix-hydride interface. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Hollandite ceramics are well-recognized as a promising host for immobilizing radioactive cesium. In the present paper, the [BaxCsy][(Al3+,Ti3+)(2x +y)Ti-8-2x-y(4+)]O-16 (0.4 <= x, y <= 0.8) ceramics were fabricated to in-vestigate the effect of incorporated Cs on structural stability and durability of (Ba,Cs)(Al,Ti)(8)O-16 ceramics with Cs-incorporated. It was found that the sintered samples at 1250 degrees C show a pure hollandite phase with tetragonal structure (I4/m) and high Cs retention. Moreover, the synthesized (Ba,Cs)(Al,Ti)(8)O-16 ceramics exhibit an excellent aqueous stability and the normalized Cs release rate is 2.82 (+/- 0.27) x10(-3) g m(-2) d(-1) after 28 days. All these results reveal that (Ba,Cs)(Al,Ti)(8)O-16 is a promising candidate as a Cs-waste form. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Aluminum (Al) matrix composites with boron carbide (B4C) reinforcements were fabricated by solid state powder metallurgy using the hot-pressing process. Composite materials were fabricated at different vol-ume fractions of B4C particles, ranging from 2% to 12%, to evaluate the impact of B4C reinforcements on the thermal , mechanical properties of the composite materials. Thermal properties, such as thermal conductivity (TC) and the coefficient of thermal expansion (CTE), were measured and modeled. The me-chanical properties were evaluated by Vickers macro-hardness (HV) and tensile tests to obtain the strain hardening threshold (sigma y), ultimate tensile stress (UTS) , elongation (A) of the developed composites. Microstructures were observed by scanning electronic microscopy (SEM) and transmission electron mi-croscopy (HRTEM) to show the homogeneity of composites materials with different B4C contents and to characterize the Al/B4C interface. This article shows that incorporating B4C particles until 12% in the Al matrix increased the hardness ( + 85%) and strain hardening threshold ( + 55%) of the composite material and decreased the ductility. An increase, up to 8 vol.% B4C, of mechanical properties which a decrease of the elongation at rupture is measured. The strain hardening threshold and the UTS strength increased up to 37% and 13%, respectively. For higher B4C volume fraction, Al/B4C become more brittle leading to very limited plastic phases. Moreover, both the TC and the CTE decreased as a function of the increase of the B4C volume fraction; 20% decrease of TC was measured for an Al/B4C (12 vol.%). The thermal and me-chanical properties were correlated with the microstructure of the Al matrix and of the Al-B4C interfacial zone. (C)& nbsp;2022 Elsevier B.V. All rights reserved.
Zhou, YufanVelisa, GihanSan, SaroCrespillo, Miguel L....
12页
查看更多>>摘要：Understanding chemical disorder in many concentrated solid solution alloys (CSAs) at the levels of electrons and atoms has attracted increasing attention as a path forward to reveal and identify underlying mechanisms for extraordinary mechanical properties and improved radiation tolerance. Single-phase NiFeCoCr CSA is a common base for many high-entropy alloys (HEAs) that have shown improved mechanical strength and radiation tolerance. In this study, defect production and damage evolution in NiFeCoCr under ion irradiation at room temperature to dose over 20 dpa are determined using ion channeling technique along both < 100 > and < 110 > directions utilizing multiple probing beam energies. The results obtained from the multi-axial and multi-energy channeling analysis are compared with those previously obtained for Ni crystals irradiated under similar conditions. The influence of chemical complexity on defect production and clustering at early-stage under room temperature irradiation up to dose of 1 dpa is discussed based on positron annihilation spectroscopy results. Defect structure evaluation in Ni and NiFeCoCr is also discussed based on transmission electron microscopy results over a prolonged irradiation at both room and elevated temperatures. Compared with chemically complex NiFeCoCr, larger dislocation loops thus less lattice strain are expected to form in pure Ni. Moreover, the role of chemical disorder in this CSA is also investigated based on ab initio calculations using large supercells. To understand the impact of chemical complexity effect on defect structure evolution, this integrated research effort attempts to link the relatively large charge redistribution due to difference in valence electron counts resulting from alloying different 3d transition metal elements, moderate lattice distortion arising from similar adaptable atomic size, and notable suppressed or delayed damage evolution in NiFeCoCr. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The interest to immobilize & nbsp; Cs-137 can pursue different aims. When it is separated from spent nuclear fuels, it is possible to reduce the thermal loading of the host glass and increase the waste loading capacity of the matrix. Also, the separated Cs-137 can be used to produce radioactive sealed sources useful in industrial and medical applications. If reprocessing is not envisaged after the generation of electrical energy, the spent nuclear fuel is the most important radioactive waste generated in the nuclear fuel cycle. In this case, aqueous solutions obtained from the chemical dissolution of spent nuclear fuels are high-level wastes. Short-lived fission products, such as Sr-90 and Cs-137, constitute the main source of heat generation from beta decay, causing self-heating of the glassy matrix. It is well known that vitrification is the main used technology, capable to immobilize these wastes during hundreds of years generating a waste form with proved structural stability, thermal shock resistance, and high chemical durability. In this article we present in an analogous way the immobilization of an specific amount of stable cesium in a porous Si-rich glass matrix through adsorption and sintering. Sintering requires lower temperature than that required in immobilization by melting. The leaching behavior of the waste form obtained was studied from the procedure described by the MCC-1 test method. Considering Cs-137 only for simulations, this work also includes the thermal evolution calculation of a simulated silica glass block loaded with 2 wt.% of Cs-137. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The properties of the oxide films grown on 308 L SS cladding with various surface treatments and temperatures in simulated PWR primary water are studied. The thickness of the inner oxide layer increases with the decrease of the surface roughness. The high oxidation resistance of the ground surface is due to the thicker fine-grained layer with high dislocations density and subgrain boundaries in the near-surface. The ferrite-affected oxidation zone (FAOZ) expands with the increase in temperature, resulting from the thermally activated self-diffusion of Cr in ferrite. The effects of surface states, ferrite and temperature on oxidation performance of 308 L SS cladding are discussed.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Polycrystalline tungsten is a promising candidate as plasma-facing armour material in future fusion reactors. Below the DBTT tungsten shows brittle or semi-brittle behaviour with no or little plastic deformation before cleavage fracture. Moreover, tungsten has an inherent low fracture toughness as well as a large scatter within the strength properties. This requires a statistical treatment of the strength values. Therefore, four-point bending tests in the temperature range of room temperature (RT) to 400 degrees C were performed to allow the Weibull analysis in brittle failure regime and/or the Beremin model analysis in case of onset of plastic behaviour while approaching the DBTT temperature. The analyses are supported by FE simulations. Weibull and Beremin parameter values are determined from the simulation results using post-processing codes. (c) 2022 Elsevier B.V. All rights reserved.
Yee, Kay L.Harikrishnan, R.Perez-Nunez, DeliaJiang, W....
15页
查看更多>>摘要：Uranium Dioxide (UO2) fuel powers almost all commercial Nuclear Power Plants (NPPs) worldwide, generating carbon-free energy and contributing to the fight against climate change. UO2 fuel incurs damage and fractures due to large thermal gradients that develop across the fuel pellet during normal and transient operating conditions. A comprehensive understanding of the underlying mechanisms by which these processes take place is still lacking. A combined experimental and computational approach is utilized here to quantify the behavior of UO2 fuel fracture induced by thermal shock. This work introduces both (1) an experimental study to understand the fuel fracturing behavior of sintered UO2 pellets when exposed to thermal shock, and (2) a Multiphysics phase-field fracture model capable of simulating this process. Parametric studies were conducted to evaluate the effects of uncertainties in fracture properties on the fracture behavior of UO2 due to thermal shocking. A set of energy release rate (or equivalently fracture toughness) and contract area (the part of the fuel pellet in direct contact with the cold bath) were able to capture the overall fracture trends of the corresponding experimental data. Our combined approach presents a new method for accounting for the effects of microstructure and sample size on the energy release rate/fracture toughness. The experimental data were collected from multiple experiments that exposed UO2 pellets to high-temperature conditions (589-676 degrees C) followed by a quench in sub-zero water. This work demonstrates that joint experimental and computational efforts are able to advance the understanding of thermal fracture in the primary fuel source for existing and future NPPs. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Grain morphologies such as grain size and aspect ratio in uranium-based metallic fuels are important microstructural features that can impact various fuel performance properties such as fission-gas-induced swelling, thermal transport, high burnup structure formation, and radiation resistance. Accurate prediction of the fuel grain morphologies requires knowledge of critical grain growth parameters such as grain boundary (GB) mobility and anisotropy. In this work, molecular dynamics (MD) simulations were performed to study the GB mobility and its anisotropy in pure body-centered-cubic (BCC) gamma uranium. Nine GBs with different combinations of misorientation angles (20 degrees, 30 degrees, 45 degrees) and rotation axes ( < 100 > , < 110 > , < 111 > ), as well as an additional < 111 > 38.2 degrees GB were studied using three interatomic potentials. It is found that the GB mobility anisotropy has complex trends, depending on both rotation axis and misorientation. However, in general the < 110 > rotation axis has the fastest GB mobility at the same misorientation. The results of this work can be used as not only a baseline for future studies of GB mobility in uranium-based alloys such as uranium-molybdenum (U-Mo) fuels, but also input for mesoscale modeling of grain growth in uranium-based alloys.
查看更多>>摘要：Three ferritic/martensitic alloys were studied to understand the synergistic effect between single ion beam (Fe 2 + ), dual ion beam (Fe 2 + + He 2 + and Fe 2 + + H + ), and triple ion beam (Fe 2 + + He 2 + + H + ) irradiations on cavity evolution. A commercial alloy, F82H, a castable nanostructured alloy, CNA3, and a high purity model alloy, Fe8Cr2W, were irradiated at 400 degrees C to 600 degrees C to a damage level of 50 dpa at a damage rate of 1 x 10 -3 dpa/s with He and H injection rates of 10 and 40 appm/dpa, respectively. Post-irradiation characterization via bright field transmission electron microscopy and high-angle annular dark-field scanning transmission electron microscopy was performed on all irradiated conditions to characterize the cavity size distribution and determine the effects of H/He injection on cavity microstructure. In all three alloys, hydrogen co-injection with helium resulted in an increased cavity number density and maximum cavity size, producing an increase in swelling over that from helium injection alone. Swelling in F82H appears to peak between 450 degrees C and 500 degrees C. At 600 degrees C, swelling was minimal and cavities of high density and small size were confined to grain boundaries and dislocations while at 400 degrees C, swelling is also low with a nearly homogeneous, high density, distribution of very small cavities throughout. Swelling was least in the commercial alloy F82H due to the high sink strength. The CNA3 alloy underwent dissolution of precipitates that lowered the sink strength and resulted in higher swelling than F82H, but less than the model alloy. Electron energy loss spectroscopy (EELS) elemental mapping revealed hydrogen forming a halo-like structure about the periphery of the cavities and helium residing within the cavities themselves. This observation suggests that hydrogen reduces the surface energy of helium-filled cavities which results in both increased cavity number density and cavity size in triple beam irradiation over dual beam irradiation.
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