查看更多>>摘要：Uranium carbides (UCs) are prevalent inclusions in U metal that form during melting operations because of interactions with crucible walls and the casting chamber atmosphere. Although UCs have been studied extensively since the beginning of U metal foundry operations, there are still unknowns regarding the effects of thermomechanical processing on their sizes and morphology. Here, we present the results of a series of controlled cooling experiments with molten uranium to elucidate the effect of cooling rate on inclusion morphology in α-U. Samples were melted using a vacuum induction melter and manually cooled at rates of 2.5, 1.7, 1.1, 0.8, and 0.3 (± 1%) K/s from ～1600 K to < 700 K in under 1 h. Subsequent scanning electron microscopy (SEM) was performed on cross sections of the samples, revealing a complex mixture of UC morphologies that are indicative of diffusion and growth influenced by the thermal processing of the U matrix. Image analysis using the morphological analysis of materials (MAMA) software showed that UC sizes generally grew larger with slower cooling rates, and the two slowest cooling rates noticeably affected the inclusion circularity and ellipse aspect ratio. These results indicate that UC morphology is sensitive to short cooling rates (< 1 h) and could therefore be controlled in the production of metallic nuclear fuels. Additionally, inclusion speciation and morphologies could potentially provide forensic clues about the processing history of unknown metal samples. Understanding the driving forces involved in UC morphology evolution is beneficial for evaluating metal fuels for next-generation nuclear reactors and for identifying signatures for nuclear forensics.
查看更多>>摘要：A comprehensive thermodynamic investigation has been performed on irradiated uranium dioxide CANDU fuel under representative severe accident conditions. Two cases have been explored in this work: Case I) the fuel is in contact with varying combinations of hydrogen and steam (i.e., mainly reducing), and Case II) the fuel is in contact with varying combinations of air and steam (i.e., oxidizing). The system thermochemistry has been computed for a wide range of gas mixtures for both cases, fuel to gas ratios, temperatures, and hydrostatic pressures. The calculations predict fission gas speciation and fuel volatilization based on equilibrium thermodynamics, including the formation of several secondary phases, changes in UO2±x stoichiometry, fission product solubility in various phases, melting, and vaporization. The overall objective of this work was to inform source term analyses in the industry standard toolset SOURCE under severe accident conditions of a CANDU station.
查看更多>>摘要：Rhenium (Re), a non-radioactive surrogate for technetium (Tc), was incorporated into the lattice of SnO2 by reductive co-precipitation process at room temperature, and Re volatility from Re(SnO2) under different conditions (in air and N2 and with glass frit) at elevated temperatures was carefully investigated. Re began to evaporate from Re(SnO2) both in air and N2 at 400 °C, while Re exhibited a faster volatilization in air than in N2, as all Re was gone at 1200 °C in air, but ～40% of Re still remained at 1200 °C in N2. Additionally, Re did not volatilize from Re(SnO2)-containing glass frit until 800 °C and only ～15% of Re volatilized at 1200 °C, suggesting the loss of Re decreased by ～30% points through incorporation of Re into SnO2 before vitrification. The analysis of Re valence transition revealed that the reduced Re4+ was easily oxidized to volatile Re7+ in air, but in N2, Re4+ disproportionated to form Re7+ and the lower valence states of Re at temperatures from 600 °C to 1000 °C. In Re(SnO2)-containing glass frit, the formation of NaReO4 during vitrification inhibited Re7+ volatilization before 800 °C, while the presence of Re4+ in the glass melt enhanced Re retention at temperature above 800 °C.
查看更多>>摘要：Self-ion irradiation of pure tungsten with 2 MeV W ions provides a way of simulating microstructures generated by neutron irradiation in tungsten components of a fusion reactor. Transmission electron microscopy (TEM) has been used to characterize defects formed in tungsten samples by ion irradiation. It was found that tungsten irradiated to 0.85 dpa at relatively low temperatures develops a characteristic microstructure dominated by dislocation loops and black dots. The density and size distribution of these defects were estimated. Some of the samples exposed to self-ion irradiation were then implanted with deuterium. Thermal Desorption Spectrometry (TDS) analysis was performed to estimate the deuterium inventory as a function of irradiation damage and deuterium release as a function of temperature. Increase of inventory with increasing irradiation dose followed by slight decrease above 0.1 dpa was found. Application of Positron Annihilation Spectroscopy (PAS) to self-irradiated but not deuterium implanted samples enabled an assessment of the density of irradiation defects as a function of exposure to high-energy ions. The PAS results show that the density of defects saturates at doses in the interval from 0.085 to 0.425 displacements per atom (dpa). These results are discussed in the context of recent theoretical simulations exhibiting the saturation of defect microstructure in the high irradiation exposure limit. The saturation of damage found in PAS agrees with the simulation data described in the paper.
查看更多>>摘要：High-energy X-ray diffraction and electron backscatter diffraction techniques were employed to investigate the hydride precipitation in Zr-1Nb-0.01Cu cladding tube under different internal pressures up to 30 MPa. Hydride reorientation occurs when the internal pressure exceeds 10 MPa. Higher internal pressure applied to the cladding tube induced plastic deformation, which significantly increases the nucleation sites of the δ-hydrides. The different evolutions of the (111) D-spacing with the azimuth angles indicated that the δ-hydrides precipitated from the cladding tubes under different internal pressures possessed different residual stress fields. Detail analysis of stereo-graphic projections confirmed that both the circumferential and radial δ-hydrides precipitated under different internal pressures followed the {0001}α-Zr // {111}δ-hydride and <112ˉ0>α-Zr // <110>δ-hydride orientation relationship with their parent α-Zr matrix. The δ-hydride {111}<112ˉ> twins may form when the Shockley partial dislocations glide on the alternate close-packed (0001) basal planes and the alternating distribution of δ-hydride twins may result in a deflection of hydride packet. The δ-hydride precipitates along the radial direction of the cladding tube under internal pressure is more adaptable to the volume expansion induced by phase transformation.
查看更多>>摘要：Using a thermogravimetric analyzer and an oxygen sensor, relative oxygen potentials of mixed oxide (MOX) fuels with Pu contents of 45% and 68% were measured at 1473 K – 1873 K; the relative oxygen potentials provide the relation among the equilibrium oxygen-to-metal (O/M) ratio, temperature, and oxygen partial pressure of the flowing gas. To evaluate the effect of the Pu content on the relative oxygen potential of MOX and related defect formation energy, this work aims to bridge the gap in the results available in existing reports, where Pu contents of 0%-46% and 100% have been studied [1–11]. The measured data indicate that the trend of increasing relative oxygen potential with increasing Pu content, which can occur as a consequence of more Pu cations changing from being tetravalent to trivalent, reaches a saturation at a Pu content of 68%, in particular in the region of O/M < 1.96. As the formation of defects related to oxygen results in oxidation and reduction of nonstoichiometric MOX and the intrinsic ionization is the dominant defect formation process, the relation between the equilibrium O/M ratio and the oxygen partial pressure was well reproduced by introducing the equilibrium constants of the defect formation process. The highest concentration of electron and hole defects caused by the intrinsic ionization was observed in MOX with a Pu content of 68%, and the related properties, such as electrical conductivity and specific heat, also exhibited the highest values upon increasing the defect concentration at high temperature.
查看更多>>摘要：? 2021 Elsevier B.V.The creep properties, microstructural evolution, and fracture mechanism of 16Cr-3Al oxide dispersion strengthened (ODS) steel were systematically investigated at 600 °C under different uniaxial tensile stresses in the range of 160–200 MPa. The minimum creep rate had a linear relationship with the mean creep rate, which followed the modified Monkman-Grant relationship with a slope of 0.57. The threshold stress was determined as 145.09 MPa based on the modified Bird-Mukherjee-Dorn (BMD) equation. Both Laves phase and Cr-rich phase were identified after creep deformation. The evolution of Cr-rich phase during creep deformation included the dissolution of W-rich σ phase and the precipitation of Si-rich σ phase. Cross-sectional cracks could be observed near the rupture surface, and these cracks mainly formed near the Si-rich σ phase and the boundaries between coarse and fine grains. It was considered that the formation of Si-rich σ phase in fine grains was the main fracture mechanism of 16Cr-3Al ODS steel during creep deformation.
查看更多>>摘要：Metallic nuclear fuels have been extensively investigated for more than 60 years. Early studies noted that metallic fuels exhibit characteristic irradiation behaviors that set them apart from oxide fuels. By performing a critical review of historical data, the current gaps in the understanding of metallic U-Zr and U-Pu-Zr fuels (e.g., constituent redistribution, swelling and fission gas release) and recommendations for future research direction can be identified. This paper reviews and highlights the key aspects of the metallic fuel irradiation behaviors, with respect to the gaps in current understanding, based on the research performed in the past 3 years.
查看更多>>摘要：Uranium is traditionally stabilized in its ductile γ (BCC) phase by the addition of elements such as Mo, Nb or Zr (Mo is used in metallic fuel for reactor applications). Due to the resemblance of uranium alloys to ferrous alloys, an attempt was made to synthesize single phase U-based high entropy alloys (HEAs) in the U-Mo-Nb-Zr system, following empirical rules related to enthalpies of mixing and atomic radius differences. Microstructure, phase and elemental compositional were characterized, and mechanical properties were measured. This research showed that despite the expectation that a single γ (BCC) phase would be formed, these alloys usually presented two-phase structures; a U-rich γ phase and a Mo-Nb rich BCC phase or a Laves phase. Thermodynamic calculations were successful in predicting the content of phases in the alloys but were not in full agreement with the experimental results. Small punch tests (SPT) showed that most of the studied samples were hard and brittle, which could be attributed to the presence of the Laves phases or alternatively, could be correlated with the multi-component γ (BCC) phase, since other BCC HEAs also tend to be hard and brittle.
查看更多>>摘要：Silicone oil continues to be used as a gelation medium for internal gelation process. Typically, the silicone oil has been removed from the gel spheres using trichloroethylene (TCE) washes. However, TCE is no longer a viable option for large-scale operations because of environmental issues, health concerns, and disposal costs. During the development of a large gelation system at the Y-12 National Security Complex, two new TCE wash replacements were identified. NuSolv SOR-C and FluoSolv WS were used in a number of uranium runs and found to effectively remove silicone oil. NuSolv SOR-C was determined to be the best option.
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Elsevier