Results 1 - 10 of 16163
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[en] Single crystal Fe thin films (∼250 nm) were grown on MgO substrates and irradiated with 2.0 MeV Fe+ ions at 10 and 50 dpa, and the defect evolution was studied using high resolution Transmission Electron Microscopy (HR-TEM) and Doppler Broadening Positron Annihilation Spectroscopy (PAS). It was shown that irradiation induced or exacerbated a thin oxide layer at the outer interface and produced substantial Fe/Mg mixing at the film/substrate interface, particularly for the higher dose. Modeling of the PAS data allowed interpretation of the defect types at different distances from the Fe surface, and included several types of MgO substrate damage and annihilation condition changes, indicative of damage due to ballistic effects of the Fe atoms as well as chemical changes due to implantation and subsequent diffusion. This detailed PAS study compared with TEM and energy dispersive spectroscopy (EDS) provides significant insight into depth-dependent defect creation. These results will be useful for predicting defect creation in Fe-based materials under irradiation conditions, for extension to neutron irradiated structural materials.
[en] Thorium dioxide (ThO2), is a nuclear fuel that is expected to play a vital role in the Generation IV nuclear reactors. One of the challenges to implementation of thoria for the fuel cycle has been the difficulty in fabricating dense pellets via conventional sintering techniques. In this study, the non-conventional sintering of thoria using spark plasma sintering (SPS) was explored. A systematic investigation into the influence of processing parameters on the densification, microstructure, grain size and thermal conductivity of ThO2 is presented. The range of sintering temperature, pressure, and hold time has been systematically varied between 1500 and 1800 °C, 50–70 MPa and 5–15 m, respectively. The results revealed that without the help of any sintering aid, pellets with a relative density of 95% theoretical density (TD) were fabricated at a sintering temperature of 1600 °C, sintering pressure of 50 MPa and a hold time of 10 min. Furthermore, the characterization of these specimens clearly indicates that by carefully selecting the processing parameters, the density, microstructure, grain size and the thermal conductivity of ThO2 can be suitably controlled. This study shows that the use of the SPS technique can potentially solve one of the primary concerns in the front end of the thoria fuel cycle.
[en] Stable structures of hydrogen atoms trapped in a divacancy in tungsten and their binding energies are presented on the basis of first-principle calculations. The hydrogen atoms are favorable sitting in the vicinity of octahedral interstitial sites (O-sites) next to the divacancy. Besides, hydrogen atoms preferentially occupy O-sites located in the center of the divacancy. As hydrogen atoms increases, O-sites located in the periphery of the divacancy are also occupied by the hydrogen atoms. The divacancy in tungsten is energetically unstable, compared with two isolated monovacancies. However, the divacancy is extremely stabilized by the hydrogen atom trapping. The binding energy of the divacancy depends on the sort of the hydrogen isotope.
[en] Low dose neutron irradiation has been conducted on a NiFeMnCr (Co-free) high entropy alloy (HEA) near room temperature. Post-irradiation examination (PIE) at room temperature revealed that this HEA exhibited qualitatively similar change in mechanical properties (hardness, strength and ductility) as conventional austenitic stainless steels. Isochronal annealing was performed to determine the hardness, electrical resistivity and defect property evolution with respect to annealing temperature. Different annealing behavior was observed. Irradiation hardening nearly completely anneals by 650 °C, although the annealing trend is different following exposure to 0.1 and 1 displacements per atom (dpa). Positron Annihilation Spectroscopy (PAS) measurements indicate that the temperature to initialize vacancy cluster (stage V) recovery is ∼400 °C, and that the density of vacancy-type defects is reduced to that of the unirradiated control specimen at 500 °C. No significant change in the chemical ordering near defects was measured by PAS after neutron irradiation or after isochronal annealing up to 700 °C. Electrical resistivity measurements revealed a large increase in resistivity following irradiation, and this resistivity increase does not significantly recover even after the 700 °C anneal. Overall, these annealing results indicate the HEA vacancy cluster annealing behavior is similar to conventional face centered cubic (FCC) alloys whereas solute diffusion is limited up to 700 °C (significantly more sluggish than conventional FCC alloys). X-ray diffraction (XRD) and transmission electron microscopy (TEM) results indicated that this HEA exhibit good phase stability upon neutron irradiation up to 1 dpa and post-irradiation annealing up to 700 °C.
[en] The goal of achieving higher thermal efficiency in nuclear power systems, whether fission or fusion based, has invariably led to the study and development of refractory metals, ceramics, and their composites. Silicon carbide materials, owing to their favorable neutronic and high-temperature properties, have seen extensive study for over half a century in support of this goal. Currently, our community has a relatively deep understanding of the irradiation effects on this system and has developed irradiation-hardened materials that are currently in use for fission reactor fuels and available as structural composites for next generation reactors. Outside of the nuclear arena SiC has also enjoyed significant development with a wide range of ordinary and high-value product now in use including very high temperature commercial aerospace installations such as turbine engines. The paper presents a brief history of the development of SiC, focused on but not limited to irradiation applications that has led to our present understanding of the system for nuclear application.
[en] Highlights: • Metallurgical and electrochemical formation of Bi-Ce alloys at 500 °C. • Vertical entire cross-sectional microscopic images of the Bi-Ce alloy. • Characterization of floating CeBi2 in the Bi-Ce alloy by SEM-EDS and XRD. • The effect of Ce concentration and cooling rate on the vertical distribution of CeBi2. • Suggestion on density-based separation of An and Ln in liquid Bi for pyroprocessing. - Abstract: This study presents the spatial distribution of CeBi2 in Bi-Ce alloys formed in various conditions to investigate density-based separation using liquid Bi between the intermetallic compound of actinides and lanthanides in used molten salt from pyroprocessing. It was experimentally identified that CeBi2 clearly formed and floated at the top of Bi-Ce alloy. Four Bi-Ce alloys were metallurgically prepared to investigate the feasibility of floating intermetallic compounds with the consideration of Ce concentrations in the liquid Bi alloy and cooling rates. Cyclic voltammetry (CV) of CeCl3 in LiCl-KCl was conducted to examine the electrochemical behavior of Ce ion on liquid Bi pool electrode at 500 °C. The Bi-Ce alloy formed by the galvanostatic electrolysis of LiCl-KCl-CeCl3 on Bi cathode at 500 °C. The applied cathodic current was determined to be 20 mA/cm2 based on the CV results. The spatial distribution of intermetallic compounds was obtained by scanning electron microscope images which focused on the vertically entire cross sections of all Bi-Ce alloys. The intermetallic phase was characterized by both energy dispersive spectroscopy and X-ray diffractometry. From the experimental results, we suggest the feasibility of the density-based separation process and its flowsheet for the application to pyroprocessing technology.
[en] With the goals of exploring the feasibility of replacing Zr with Fe and P in Cu–Cr–Zr alloys and developing Cu–Cr–Fe–P alloys with excellent mechanical and electrical properties, the effects of Fe and P additions on the microstructures and properties of the as-cast, cold-rolled and aged Cu-0.8Cr alloys are investigated. The results indicate that adding Fe and P into the Cu-0.8Cr alloy can promote the precipitation of a large amount of fine, uniformly dispersed Cr and Cr3P particles. From Cu-0.8Cr to Cu-0.8Cr-0.17Fe-0.048P, the area fraction of precipitates increases from 3.1% to 6.3% while their average diameter reduces from 1.6 μm to 1.4 μm. When subjected to 95% cold rolling plus 450 °C aging for 1 h, the Cu-0.8Cr-0.17Fe-0.048P alloy exhibits a Vickers hardness of 140.4 HV and an electrical conductivity of 73.2%IACS, higher than 96.7 HV and 58.4%IACS in the Cu-0.8Cr alloy. Among three different processing routes investigated, the one with initial 50% cold rolling → aging at 500 °C for 1 h → final 95% cold rolling allows the Cu-0.8Cr-0.17Fe-0.048P alloy to gain most in both its tensile strength (569 MPa) and electrical conductivity (64%IACS). The findings in this work provide insights to composition and processing designs for new Cu–Cr alloys with high strength and electrical conductivity.
[en] Interaction of steel specimens such as SS 316 LN, D9, 9Cr–1Mo, oxide dispersed steel (ODS) and pure iron with liquid lithium has been studied at 823 and 973 K (550 and 700 °C) for various durations from 250 to 1000 h at 250 h intervals. The relevance of this study is to identify a suitable container material for liquid lithium which has been proposed as a liquid poison to control the reactivity in nuclear reactors. The metallic specimens were characterised before and after exposure to liquid lithium by X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive X-ray analysis (EDX) and mechanical testing. The results indicate that except for pure iron, lithium interacts with all steel samples in varying degree. The depth of interaction varies from 20 to 60 μm depending on nature of specimen, temperature and time of exposure.
[en] WCrY Smart Alloys are developed as first wall material of future fusion devices such as DEMO. They aim at behaving like pure W during plasma operation due to depletion of the alloying elements Cr and Y. The Cr concentration gradients induced by preferential plasma sputtering cause Cr-diffusion. The exposure of WCrY and W samples to pure D plasma, with a plasma ion energy of , is simulated using the dynamic version of SDTrimSP. Cr-diffusion is included into the model. Simulation results are compared with experimental results. At sample temperatures of more than 600∘C and sputtering by D plus residual oxygen in the plasma ion flux, the Cr-transport to the surface leads to enhanced erosion for WCrY samples. A diffusion coefficient for Cr in WCrY of the order of is determined. The suitability of WCrY as first wall armour and the influence of further effects, considering especially Cr-diffusion, is discussed.
[en] The feasibility of utilizing the suspended bridge method, which was originally developed for one-dimensional or nearly-one-dimensional nanomaterials, to measure the thermal conductivity of lower-length-scale structures in nuclear materials is explored in this study. Nanoribbon specimens of stainless steel SS304, representing materials with well-known thermal conductivity, and atomized U-Mo alloy particles used in dispersion fuels for research reactors, representing new nuclear materials with limited thermal conductivity data reported, were made using focused ion beam (FIB). The contact thermal resistance was corrected by measuring a series of specimens with different bridge lengths. The measured thermal conductivity of SS304 was found to be consistent with that reported for bulk samples. The thermal conductivity of U-Mo particles measured using the suspended bridge method was also analyzed and compared with literature data of monolithic U-Mo alloys. Ab initio molecular dynamics (AIMD) was used to quantitatively demonstrate that the reduced specimen size only has marginal effects on the measured thermal conductivity compared to the bulk specimens. The novel concept of utilizing the suspended bridge method in nuclear material research is proven and future work is discussed.