查看更多>>摘要：Pilot-scale melter testing has shown that Al and Fe sources in high-level nuclear waste feeds influence the glass production rate. To examine melting behaviors of these feeds, we employed X-ray diffraction, differential scanning calorimetry, thermo-gravimetric analysis, particle size analysis, and the feed expansion experiments. Both the chemical form and the particle size of the Al and Fe sources affect the rate of melting through their effects on the conversion enthalpy and the primary foam formation that control the energy demand for melting and the heat accepted by the cold cap. The particle size of gibbsite controls the rate of alumina incorporation in the initial glass-forming melt that in turn, through its effect on viscosity, affects the rate of dissolution of silica particles, thus governing the glass melt fraction, open pore closure, and primary foam formation. Since the temperature of primary foam collapse limits the heat flow to the cold cap, the gibbsite (and generally Al source) particle size ultimately influences the glass production rate that increases as the particle size increases. Variation in the Fe source affects the glass production rate mainly through their content of chemically bound water.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：In the hierarchically-arranged crystallographic structure of reduced-activation ferritic/martensitic (RAFM) steel, the smallest but the most abundant microstructural unit, i.e., the lath, plays an important role in dislocation plasticity. Due to the multi-scale complexity in the lath-martensitic microstructure, miniaturized mechanical characterization at very fine scale is required to understand the deformation mechanism associated with laths. In this study, uniaxial micro-compression tests combined with rigorous crystallographic analysis were performed to figure out the plastic deformation mechanism of lath boundary sliding in RAFM steels. These experimental results were further interpreted via molecular dynamics simulations to discover the underlying dislocation mechanisms. We found that the amount of lath boundary sliding is controlled by the crystallographic orientation of the lath boundary plane, the direction of Burgers vectors of the interfacial dislocations, and the magnitude of resolved shear stress on the lath boundary plane. Also, the effect of normal stress on the lath boundary plane was investigated. (C) 2021 Elsevier B.V. All rights reserved.
Tom, Phongsakorn PrakMurakami, KentaLuu, Vu NhutNguyen, Ba Vu Chinh...
6页
查看更多>>摘要：The dislocation loop is known as one of the irradiation defects that contributes to hardening and accompanying embrittlement in reactor pressure vessel steels. In this study, the effect of solute elements such as Ni and Mn in Fe-based alloys on dislocation loop evolution was investigated using ion irradiation. Pure Fe and Fe model alloys, Fe-0.6Ni and Fe-0.6Ni-1.4Mn, were irradiated by 2.8 MeV Fe ions to a peak dose of 3 displacements per atom (dpa) at 400 degrees C. Quantitative analysis of transmission electron microscopy (TEM) was conducted to analyze the evolution of dislocation loops. As a result, the presence of solute elements in Fe alloys increased the number density of dislocation loops significantly and decreased their mean size slightly. Both types of dislocation loops, 1/2< 111 > and ( 100 ) Burgers vectors, were observed in all specimens. Solute elements reduced the mobility of 1/2< 111 > dislocation loops, resulting in an increase in their number density and restricting the growth of < 100 > dislocation loops. Furthermore, an increase in the number density of < 100 > dislocation loops of small size was found with higher solute contents. (c) 2021 Elsevier B.V. All rights reserved.
Desgranges, LionelCanizares, AurelienSimon, Patrick
6页
查看更多>>摘要：The exploitation of the data acquired by Raman spectroscopy on UO2 related materials requires the understanding of the so-called defect peaks. To do so, we measured the Raman spectrum of a He-implanted UO2 sample as a function of temperature. Two annealing steps are evidenced that correspond to known annealing temperatures reported in literature. These results are discussed in order to identify what defect could be at the origin of the Raman defect peaks. (C) 2021 Elsevier B.V. All rights reserved.
Kocevski, VanchoRehn, Daniel A. A.Andersson, David A. A.Cooper, Michael W. D....
12页
查看更多>>摘要：Uranium mononitride (UN) is a promising nuclear fuel that combines the advantageous properties of readily used UO2 and uranium alloys. Various properties of UN have been previously studied using different density functional theory (DFT) methodologies; however, there are still inconsistencies when it comes to the dynamical stability and defect properties of UN. We address these inconsistencies by studying the UN phonons and defect properties using DFT calculations employing two generalized gradient approximation (GGA) exchange-correlation functionals: PBE and AM05, with and without an added on-site Coulomb repulsion term (+U). Furthermore, we investigate the importance of spin-orbit coupling (SOC) when calculating the properties of UN. We use the different methodologies to determine the preference of UN to have antiferromagnetic (AFM) ordering, as seen in experiments, or ferromagnetic (FM) ordering of the uranium spins. We compare the crystallographic properties, density of states, the DFT X-ray photoelectron spectra and phonon dispersions calculated using the different methodologies. We demonstrate that GGA + U reproduces the AFM ordering in UN, but the crystal structure is dynamically unstable. We also show that magnetic ordering is important in finding the lowest energy defective structure, and that SOC has a distinct influence on the energy of the different uranium interstitial defects. Lastly, we discuss the point defect formation energies under U-rich and N-rich conditions, and the stoichiometric formation energies calculated with the different methodologies, providing insight into the observed tendency for forming hypostoichiometric UN. (C)& nbsp;2021 Elsevier B.V. All rights reserved.& nbsp;
Shaw, Tim L. L.Jordan, Matthew S. L.Wilkinson, SamanthaRamsay, Paul G....
11页
查看更多>>摘要：The Young's modulus, Poisson's ratio and flexural strength of 218 beams of radiolytically oxidised nu-clear graphite have been determined by combining electronic speckle pattern interferometry, ESPI, and inverse finite element modelling techniques. The graphite was extracted from three operational advanced gas cooled reactor stations where it had been subjected to fast neutron irradiation and radiolytic oxi-dation resulting in mass loss. The 6 x 6 x 19 mm beam samples had densities ranging from 1063 - 1840 kg.m( -3), equivalent to weight losses up to 46%, with cumulative neutron doses of up to 161 x 10 20 n.cm -2 equivalent Dido nickel dose, or 21.1 displacements per atom, significantly extending the datasets available in the open literature to higher levels of both dose and oxidation. In situ ESPI measurements were collected during three-point bend tests, creating full field displacement maps of the evolving de-formation to which simulations were matched by chi-squared minimisation of the elastic properties. Po-tential loading abnormalities were accounted for in the fitting process to ensure the simulations were representative, with total uncertainties in the determined Young's modulus and Poisson's ratio values of +/- 4.5% and +/- 0.08, respectively. Consistent trends were observed in the behaviour of the material from the different stations, with strong positive dependences on density for both the flexural strength and Young's modulus. The strength and Young's modulus remained well correlated even to high levels of ox-idation mass loss. Conversely, the Poisson's ratio was practically invariant with oxidation with mean val-ues of 0.21 +/- 0.03, 0.24 +/- 0.05 and 0.25 +/- 0.05 for the three stations. The combined ESPI-finite element measurement approach is demonstrated as robust for irradiated small specimens, making it suitable for monitoring graphite in future reactors. The collected data support the safety assessments of the current reactors, enabling predictions of the responses to progressively higher mean core mass loss. (C)& nbsp;2021 Elsevier B.V. All rights reserved.
Turner, J.Buckley, J.Worth, R. N.Salata-Barnett, M....
9页
查看更多>>摘要：As part of a scoping study to investigate the initial feasibility of UN-UB 2 composites, pellets of 5, 25 and 50 % UB 2 within UN were produced via Spark Plasma Sintering alongside a high density UN material. The pellets were seen to have a well-formed and dense microstructure which did not appear to change after annealing for 24 hours at 1273 K, which consisted of UN, UB 2 and a UBN phase, with an additional UB 0 . 1 N 0 . 9 phase suggested via EBSD. Steam exposure in an STA demonstrated that an increase in reaction onset temperature of approximately 150 K can be achieved via the addition of 25 % UB 2 , and an increase of 120 K in onset temperature was evident in pellets containing only 5 % UB 2 . High density UN was tested alongside these composites and behaved similarly, but microstructural examination of composites postoxidation suggest that the oxidation rate is retarded in porous regions in addition to highly dense regions. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The concept of high entropy materials (HEMs) provides a fertile ground for developing novel irradiationresistant structural materials. In HEMs, the vast and complicated configurational space induced by extreme disorder poses grant challenges to understanding defect dynamics and evolution. Machine learning (ML) techniques, which can exploit implicit relationships between diverse descriptors and observations, exhibit great potential in uncovering the governing factors for irradiation damage and modeling local environment dependence of defect dynamics. Herein, three applications of ML in understanding radiation damage in HEMs are summarized and discussed, including ML-based irradiation response prediction, MLbased interatomic potential development, and ML-informed defect evolution.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Helium plasma is known to affect recrystallization in tungsten, with lower temperatures during plasma exposure leading to slower crystal grain growth. To understand why this occurs, tungsten samples were first exposed to helium plasma at surface temperatures between 300 degrees C and 800 degrees C, before annealing at temperatures between 1100 degrees C and 1400 degrees C. Annealing after helium exposure at 300 degrees C was confirmed to lead to smaller crystal grains than annealing after exposure to helium at 500 degrees C. Small 1-2 nm radius nanobubbles formed readily in tungsten after helium plasma exposure, but disappear after annealing at temperatures of 1100 degrees C and above. The formation of cracks and open volumes beneath the surface was observed exclusively in tungsten exposed to helium-plasma at 300 degrees C, with extensive surface cracks visible after annealing. These cracks were not observed for higher temperature helium exposure and likely form due to the strong tendency of bubbles to cluster along grain boundaries for helium exposure at 300 degrees C. Despite this, nano-mechanical testing revealed a similar influence of annealing conditions on tungsten hardness for all plasma exposure conditions studied. The crack formation is likely caused by interactions between solute helium and residual defects from surface polishing. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：To study the effect of the content of chromium (Cr) in the tungsten (W) matrix on the vacancy formation and retention of hydrogen isotopes, the samples of the W-0.3 at.% Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523-1273 K to a damage level of 0.26 displacement per atom (dpa). These displacement-damaged samples were exposed to D 2 gas at a temperature of 673 K and a pressure of 100 kPa to decorate ion-induced defects with deuterium. The addition of 0.3 at.% Cr into the W matrix resulted in a significant decrease in the retention of deuterium compared to pure W after irradiation especially at high temperature ( >= 773 K). Positron lifetime in W-0.3 at.% Cr alloy irradiated at 1073 K was almost similar to that for non-irradiated one. These facts indicate the suppression of the formation of vacancy-type defects (monovacancies and vacancy clusters) by 0.3 at.% Cr addition, which leads to the significant reduction in deuterium retention in W-0.3 at.% Cr alloy.(c) 2021 Elsevier B.V. All rights reserved.
NSTL
Elsevier