查看更多>>摘要：The temperature dependence of helium bubble evolution in 316L stainless steel (SS) fabricated by selective laser melting (SLM) has been studied by performing 500 keV helium ion irradiation at temperatures between 350 and 900 degrees C. The number density of helium bubbles decreased steadily whereas the bubble diameter gradually increased with increasing temperature. The material swelling caused by helium bubbles had no obvious change at the temperature below 800 degrees C. However, at the maximum temperature of 900 degrees C, the swelling sharply increased to 0.17% mainly due to the rapidly growing bubble diameter. In particular, heterogeneous nucleation of bubbles occurred in the vicinity of dislocation structures at high temperatures of 70 0-90 0 degrees C, and the formed helium bubbles rapidly grew at the oxide-matrix interface at 900 degrees C. It was confirmed various intrinsic structures (dislocation structures or oxide particles) differently affected the bubble evolution process. The apparent activation energy (E-act) of helium bubbles determined by the Arrhenius model were used to elucidate the most probable helium bubble evolution mechanisms at various temperatures, as it is an experimentally observed activation energy, which contained information about the activity of helium atoms. It was found that the low-temperature regime with low apparent activation energies (E-Cb(act) = 0 . 15 eV, E-Rb(act) = 0 . 04 eV) and high-temperature regime with high apparent activation energies (E-Cb(act) = 1 . 50 eV, E-Rb(act) = 0 . 58 eV) favored different helium bubble evolution mechanisms. In the low-temperature regime, helium bubble evolution was controlled by the diffusion of helium atoms via a vacancy mechanism in the SLM 316L SS. On the other hand, the migration and coalescence of helium bubbles also occurred in the high-temperature regime which caused the apparent activation energy of density was higher than the theoretical value of helium atoms diffuse via a replacement mechanism. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). It was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Accurate evaluation of displacement damage is important for studying the nuclear materials after irradiation. The present work investigates neutron-induced displacement damage evaluation via damage cross sections. There are potential discrepancies between different evaluations on the damage cross sections induced by the (n,gamma) reaction and potential discontinuity at 20 MeV incident neutron energy. These two issues are however not important for evaluating neutron irradiation damage in fission and fusion reactors. Due to some issues associated with the recoil distributions in evaluated nuclear data files, the corresponding damage cross sections are questionable, such as those of Cr-52 in JEFF-3.3, Mn-55 in JEFF-3.3 and ENDF/B-VIII.0, and W-180,W-182,W-183,W-184,W-186 in ENDF/B-VIII.0. Therefore, the present work recommends using two body kinematic considerations and the differential cross sections of light particles for evaluating neutron irradiation damage (via either displacement cross sections or PKA spectra), unless the recoil distributions have been pre-validated/verified. The recommended method reduces the 70% discrepancy between different nuclear data evaluations to 13% for evaluated displacement damage of tungsten in ITER tokamak. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Thermal evolution and elevated temperature annealing of the dislocation microstructure of an irradiated metal, represented by an ensemble of elastically interacting interstitial dislocation loops, is explored using discrete dislocation dynamics simulations. The two fundamental microscopic processes driving the evolution of dislocations are the pipe diffusion of atoms along the dislocation lines, giving rise to the dislocation self-climb, and bulk diffusion of vacancies, resulting in the conventional dislocation climb. Simulations show that the coalescence and coarsening of the prismatic dislocation loop microstructure, observed at lower temperatures, is driven primarily by the dislocation self-climb. In tungsten, dislocation self-climb gives rise to a pronounced change in the dislocation loop microstructure at temperatures close to 800 C , see Ferroni et al. (2015) [1], whereas a similar microstructural transformation of the dislocation network driven by self-climb in alpha-iron is predicted to occur at ~270 C . Simulations also show that the diffusion of vacancies in the crystal bulk is able to explain the observed annihilation rates of interstitial loops in tungsten.Crown Copyright (C) 2022 Published by Elsevier B.V. All rights reserved.
查看更多>>摘要：The erosion behavior and retention properties of tungsten trioxide (WO3) and W films exposed to low energy (~50 eV) deuterium (D) ions were studied at surface temperature from 403 K to 758 K. The initial thickness of WO3 and W films are approximately 7.0 x 10(18) atoms cm(-2) and 1.8 x 10(18) atoms cm(-2), respectively. The erosion rate of oxygen (O) by deuterium is thermally enhanced, and tungsten (W) atoms can be sputtered rapidly in the form of WxOy molecules and/or W atoms at higher temperatures due to the relatively weak binding energy between W and O, resulting on an increase in the W sputtering yield with temperature. Nano-sized pinholes appear on the surface at 403 K, however, pinholes with much smaller sizes and cracks appear on the surface at higher temperatures, and their density increases with increasing temperatures. The synergistic effects of ion irradiation and temperature cause the structure change of tungsten oxide films according to Raman spectra. X-ray photoelectron spectroscopy shows that after irradiation with D ions, metallic W appears on the surface, but the tungsten oxide is still present. Trapped D is almost uniformly distributed in WO3 and its concentration is much higher than that in W at lower irradiation temperature, which is mainly related to D atoms chemically bonded to O atoms with the formation of deuterium tungsten bronze. The thermal desorption peak temperatures of D-2 (~465 K or 500 K) and D2O (~495 K, 730 K (or 790 K) and 825 K) are related to the heating rate (1 K s(-1)) and it implies that the deuterium tungsten bronze is decomposed under the low D ion irradiation temperatures. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The microstructure evolution of the magnetite layer on CLAM steel in lead-bismuth eutectic containing 10 (-6) wt% oxygen at 500 & nbsp;C was investigated. From 500 to 20 0 0 h, magnetite first changed from many islands to an intact and continuous oxide layer, which then fell off and regenerated; its main microstructure changed from Cr-containing powders to Cr-free crystals. The formation of island magnetite may be related to the enrichment of W in the spinel layer and the large size and poor uniformity of martensitic laths. Finally, an island growth, exfoliation and regeneration model of the magnetite layer is proposed based on the above phenomenon. (C)& nbsp;2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：ALCYONE simulations of integral RIA transients performed on UO2 fuel rods prior to or within the ALPS program in the Japanese NSRR facility with stagnant liquid water coolant conditions are presented in this paper. The 11 selected tests which cover a wide range of fuel enthalpy increases (200 J/g - 800 J/g) aimed at assessing/challenging ALCYONE capabilities on four quantities of interest: clad outer surface peak temperature (293 K - 1200 K), film boiling duration (0 s - 2 s), clad residual hoop strain (0% - 25%) and transient fission gas release (2% - 25%). Despite a global consistency between measurements and ALCYONE predictions, paths are investigated to explain and reduce discrepancies. In particular, the reevaluation of fuel enthalpies by JAEA has led to revisit ALCYONE clad-to-water coolant heat exchange models which were derived from previous values of fuel enthalpies and suggested a recalibration of some of their parameters. Then, modeling the delayed gas axial flow in the free volumes of the rod is shown essential to achieve better residual hoop strain predictions in case of clad ballooning if proper timing of fission gas release, rod internal pressure increase and clad temperature elevation can be simulated. Key experimental and/or modeling research areas are shown to be at stake for future work. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：To investigate the microstructural evolution during high temperature oxidation, Cr-coated Zr-4 alloy specimens are prepared by multi-arc ion plating and high temperature oxidation tests are carried out under 1000 degrees C, 1100 degrees C and 1200 degrees C in air atmosphere. Cr2O3 layers are formed on the surface after high temperature oxidation test. However, the thickness of Cr2O3 layer does not depend monotonously on the oxidation temperature. The oxidation temperature increases from 1000 degrees C to 1100 degrees C, the thickness of Cr2O3 layer increases from 2.35 mu m to 5.74 mu m, while the thickness is only 4.1 mu m when the oxidation temperature is 1200 degrees C. This is mainly due to the reduction of Cr2O3 to Cr, and oxygen diffuses into the substrate to form alpha-Zr. The oxidation weight gain for the three temperatures conforms to an exponential function, while the thickness of the Cr-Zr diffusion layers follows a linear relationship with the oxidation temperature. Regarding the Zr-4 alloy substrate, the Zr grains grow up and gradually change from elongated grain parallel to the interface to elongated grain along normal direction after high temperature oxidation. Moreover, different oxidation temperatures bring distinct recrystallization energy to Zr grains, and the defects and dislocations produced during the recrystallization of Zr grain are unequal. The kernel average misorientation values obtained from EBSD indicate that the overall density of defect and dislocation of Zr grains after high temperature oxidation under 1100 degrees C is higher than the other two temperatures, while Zr grains after high temperature oxidation under 1000 degrees C has the highest local density of defect and dislocation. The investigation in this study attempts to reveal the microstructure evolution of Cr coating and Zr-4 substrate during the high temperature oxidation. These results give some new perspectives on understanding the performance of Cr-coated Zr-4 alloy under high temperature environment. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：We performed a computational study to assess the suitability and equivalence of the irradiation conditions on several test irradiation facilities (either currently operating or planned to deploy in the future) aimed at the qualification of materials for nuclear fusion reactors such as ITER and DEMO. The degradation of the material's properties is driven by the changes in its microstructure and chemistry (transmutation). The primary objective of this study is thus to perform a comparison of the microstructural pattern as predicted by means of simulations. The focus of the study is put on two materials: Eurofer97 steels and tungsten. We considered operation conditions in fusion reactors (i.e. ITER and DEMO) and in test irradiation facilities such as material test reactors (fast and mixed neutron spectrum), IFIMIF-DONES, ESS and proton accelerators. Typical irradiation conditions are addressed according to the currently available design (for DEMO) and expected operation modes (for ITER). The study is realized by means of object kinetic Monte Carlo which is parameterized and configured using state-of-the-art knowledge on irradiation spectra, neutron cross-sections, primary damage states, lattice defects mobility/cohesion and interaction of the material's microstructure with lattice defects. The irradiation defects are singled out using a dedicated post-processing tools to enable a comparison with expected findings in transmission electron microscopy (TEM) and atom probe tomography (APT). The results are discussed in terms of the equivalence of the emerging irradiation microstructure predicted to occur in test irradiation facilities if compared with the one simulated in the nuclear fusion reactors. The summary and discussion provide information on the equivalence and deviations of the microstructural patterns suggesting the suitability of the test irradiation facilities for certain irradiation regimes, as well as pointing at some limitations, e.g., originating from the difference in the neutron spectra or flux. (c) 2022 Elsevier B.V. All rights reserved.
Cherubin, Igor J. S.Topping, MatthewDaymond, Mark R.
9页
查看更多>>摘要：A sample with a delta-zirconium hydride rim grown on a Zr2.5Nb pressure tube was proton irradiated to different damage levels and its mechanical properties were probed using nanoindentation. Irradiations were also carried out on non-hydrided Zr2.5Nb as a reference. A numerical model proposed by Dao et al. and developed by Wang et al. was used to calculate the yield strength and the work hardening exponent. The samples were subjected to 0.2, 0.4, and 0.8 displacements per atom (dpa), with the irradiation temperature kept below 100 C. To account for the indentation size effect, the Nix-Gao model was used to calculate the true hardness (H-0) of both materials. Both hydrided and non-hydrided samples show an increase in the true hardness and in the yield strength due to the irradiation hardening. The true hard-ness of the 8-zirconium hydride increased from 3 GPa at 0 dpa to 4.31 at 0.8 dpa. Similarly, the H-0 of the Zr2.5Nb increased from 2.53 at 0 dpa to 3.03 at 0.8 dpa. Additionally based on the model analysis, the yield strength increased from 731 MPa to 1146 MPa for the delta-zirconium hydride, and 662 MPa to 850 MPa for the Zr2.5Nb, with the latter in good agreement with literature values for Zr2.5Nb. Lastly, evaluation of the work hardening exponent with the different dpa levels suggests the defects density evolves differently with irradiation in the two materials. These novel results demonstrate a clear change in mechanical properties of the delta-zirconium hydride with irradiation. (C) 2022 Elsevier B.V. All rights reserved.
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