查看更多>>摘要：Characterization of the fuel debris has been required for decommissioning Fukushima Daiichi Nuclear Power Station. To understand the reaction mechanisms involved for mixed UO 2 - ZrO 2 compounds, these materials were treated at high temperatures (1473 to 1873 K) under oxidizing, inert, and reducing atmospheres. The reaction products were analysed by a combination of powder X-ray diffraction (PXRD) and X-ray absorption spectroscopic measurements of the U L III - and Zr K-edges. Principal component analysis of the X-ray absorption near edge structure and extended X-ray absorption fine structure of U L III - and Zr K-edges provided a breakdown of the composition of each species within the products, these results were further supported by PXRD. Under an oxidizing atmosphere, the formation of U 3 O 8 and U 2 Zr 5 O 15 was observed in equilibrium with UO 2 , monoclinic-ZrO 2 , and tetragonal-ZrO 2 . However, when O 2 gas was purged through the reaction tube during the cooling process to room temperature, pentavalent U in ZrU 2 O 7 was produced by the oxidation of solid solution UO 2 formed at > 1774 K during the temperature dropped at < 1473 K. Under the inert atmosphere, mixed oxides of uranium were found to form at > 1673 K due to a low concentration of O 2 impurity with the Ar gas. Although the oxidized UO 2 was able to form in such a system, tetravalent UO 2 and its solid solution were instead present throughout the whole temperature range examined under a reducing atmosphere (H 2 gas). This study can pave the way for understanding the interaction between the nuclear fuels and the cladding materials in damaged reactors enabling further simulation of possible decontamination procedures. (c) 2021 Elsevier B.V. All rights reserved.
Lach, Timothy G.Le Coq, Annabelle G.Linton, Kory D.Terrani, Kurt A....
7页
查看更多>>摘要：The SiC fuel matrix for advanced gas-cooled high temperature reactors as part of the Transformational Challenge Reactor program serves as the fuel particle container structure, a barrier to fission gas release, and a heat transfer medium. Its performance is particularly important because the fuel matrix must demonstrate good structural stability and thermal behaviors. An additive manufacturing methodology combining a binder jet 3D printing process with chemical vapor infiltration (CVI) for the production of SiC was recently developed. In this study, post irradiation examination by transmission electron microscopy shows that defect accumulation within the printed particles is very similar to other forms of high-purity SiC. However, damage accumulation was not directly observed in the CVI matrix because black spot damage and dislocation loops are difficult to image within the nanoscale highly faulted CVI matrix and because interstitial defects may rapidly annihilate at the stacking faults. Therefore, electron energy loss spectroscopy (EELS) analysis was used to analyze defect swelling in both the printed particles and the CVI matrix. The EELS analysis helped reveal that the radiation-induced swelling in the CVI matrix is similar to that of the printed SiC particles. This work shows that 3D printed SiC has behavior that is comparable to SiC processed by other means and that 3D printing could serve as a suitable processing technique for high-purity SiC for nuclear applications.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Arguably one of the most important factors in the fast deployment of advanced nuclear reactors, with major improvements in safety, is the development and qualification of radiation and corrosion tolerant materials, that serve as the structural components in reactor cores. However, the discovery, improvement, and assessment of materials resistant to radiation and corrosion in the advanced reactors' extreme environments is quite demanding, time-consuming, and costly, which represents a significant barrier to materials innovation and qualification for nuclear energy. This short review highlights a novel, integrated, high-throughput (HTP) research framework to develop understanding and predictive models for irradiation at high doses and molten salt corrosion responses of structural Compositionally Complex Alloys (CCAs), with the objective to accelerate materials discovery for high-temperature nuclear structural applications. Using a novel in situ alloying technique, arrays of additively manufactured bulk CCAs are processed, heat-treated, and characterized, while still attached to the build plate. Leveraging recent development in automation of heavy-ion irradiation experiments at the University of Wisconsin Ion Beam Laboratory, arrays of CCAs can be rapidly irradiated to hundreds of dpa up to 800 degrees C. An innovative droplet corrosion method is also used to test molten salt corrosion behavior of CCAs arrays. Automated and rapid characterization methods are used to assess the irradiation and molten salt corrosion resistance of CCAs. Finally, a brief discussion of the results is presented considering future use of machine-learning-based methods to develop useful trends and highlight features of importance. Using this novel HTP approach, a robust and reliable database containing literally hundreds of data points for irradiation and corrosion responses of CCAs can be established within a year, which is considered a significant increase in the pace of nuclear structural materials research and discovery. (C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：To deeply understand chemical interactions among nuclear materials during core meltdown, pre-oxidized Zircaloy-4 was re-heated under oxygen starvation to form thermodynamically stable ZrO2/alpha-Zr structure and then brought into contact with 316SS-B4C melt at 1300 degrees C. During re-heating, initial ZrO2 layer was partially dissolved due to thermodynamic instability, forming a new alpha-Zr layer at the external surface. Dissolution of ZrO2 layer could be linearly fitted with fourth root of re-heating time. When pre-oxidized and then re-heated Zircaloy-4 contacted with 316SS-B4C melt, the external alpha-Zr layer became oxygen-unsaturated and was dissolved first by 316SS-B4C melt. Initial ZrO2 layer was divided into two sublayers from the position where Zr5Sn3 phase was located. Due to direct contact with oxygen-unsaturated alpha-Zr layer, outer ZrO2 sublayer became thermodynamically unstable and was dissolved more rapidly than inner sublayer. The reaction zone consisting of multiple sublayers grew approximately linearly with time. Consequently, pre-oxidized and then re-heated Zircaloy-4 showed no significant improvement in delaying ZrO2 dissolution and reaction zone formation under the current experimental conditions, compared with only pre-oxidized Zircaloy-4. (C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Grain boundary (GB) plays a crucial role in the mechanical properties and irradiation resistance of nuclear materials. It is thus essential to understand and predict the defect properties near GBs. Here, we present a framework for predicting vacancy formation energy (E-V(f) ) near GBs in tungsten (W) by machine learning (ML) technique. The E-V(f) values of 4496 atomic sites near 46 types of [001] symmetry tilt GB (STGB) in W are calculated as database and eight appropriate variables are selected to characterizing the surrounding atomic configuration and location of atomic sites. Via the support vector machine with the radial basis kernel function (RBF-SVM), the good predicted results of cross validation (CV) and generalized verification prove the suitability and effectiveness of the selected variables and RBF-SVM method. Beside, due to their big differences in dislocation arrangement and atomic configuration, the STGBs need to be divided into three types, high angle, low angle-I and low angle-II STGBs, for adopting the Separate CV, and their predicted accuracies were found to have big improvements. Because the present method adopts geometrical factors, such as spatial size characteristic, density and location, as descriptors for the ML analysis, it is robust and general to other materials such as alpha-Fe, and beneficial to predict and understand the vacancy formation near interfaces. (C) 2021 Elsevier B.V. All rights reserved.
Harte, A.King, D. J. M.Knowles, A. J.Bowden, D....
27页
查看更多>>摘要：This review considers current Zr alloys and opportunities for advanced zirconium alloys to meet the de-mands of a structural material in fusion reactors. Zr based materials in the breeder blanket offer the potential to increase the tritium breeding ratio above that of Fe, Si and V based materials. Current com-mercial Zr alloys might be considered as a material in water-cooled breeder blanket designs, due to the similar operating temperature to fission power plants. For breeder blankets designed to operate at higher temperatures, current commercial Zr alloys will not meet the high temperature strength and thermal creep requirements. Hence, Zr alloys with an operational temperature capability beyond that of current commercial fission alloys have been reviewed, specifically: binary Zr alloy systems Zr-Al, Zr-Be, Zr-Cr, Zr-Nb Zr-Ti, Zr-Si, Zr-Sn, Zr-V and Zr-W; as well as higher order Zr alloys Zr-Mo-Ti, Zr-Nb-Ti, Zr-Ti-Al-V and Zr-Mo-Sn. It is concluded that, with further work, higher order Zr alloys could achieve the required high temperature strength, alongside ductility, while maintaining a low thermal neutron cross-section. However, there is limited data and uncertainty regarding the structural performance and microstructural stability of the majority of advanced Zr alloys for temperatures 50 0-70 0 degrees C, at which they would be expected to operate for helium-and liquid metal-cooled breeder blanket designs.Crown Copyright (c) 2021 Published by Elsevier B.V. This is an open access article under the CC BY license ( http://creativecommons.org/licenses/by/4.0/ )
查看更多>>摘要：Metal hydrides are crucial for the long-term storage of tritium but suffer degradation due to the buildup and release of helium decay products. Therefore, it is of interest to explore how dopants in these metal hydrides may impact helium bubble nucleation and distributions, as well as associated fracture and helium retention. Prior studies have focused on helium behavior in pure metal hydrides or with one or two types of impurities. Analytical models have also shown that the concentration of nucleated bubbles can impact the time to fracture of materials. This study utilizes high-throughput density functional theory calculations to identify the impact of transition metal substitutional dopants on helium binding energies in face-centered cubic metal hydrides, such as erbium hydride, holmium hydride, scandium hydride, titanium hydride, yttrium hydride, and zirconium hydride. This study also explores the impact of hydrogen vacancies on the binding energy of helium near the substitutional defects. Finally, this study presents an initial assessment of dopant stability and solubility limits at many temperatures and pressures. Several metals strongly bind to helium in these metal hydrides, making them promising for influencing helium nucleation and subsequent bubble growth. However, many of these potentially beneficial substitutional defects have low solubility limits. The calculations show that some strong binding dopants may be soluble in quantities that affect the bubble concentration and impact material performance.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Irradiation damage drives complex and coupled phenomena in materials at far-from-equilibrium conditions. The self-organization of nanoscale defects in materials under irradiation shows great potential to tailor the physical properties of materials by controlling nanopatterned microstructures. Irradiation induced gas bubble and void superlattices are two important ordered nanostructures of great scientific interest. Although both types of superlattices have been investigated extensively, a consensus has yet to be reached on their formation mechanisms. In this review article, the current research status of gas bubble and void superlattices in metals and alloys and their characterization, structural stability, and mechanistic modeling are summarized. The fundamental research goals to advance the mechanistic understanding of gas bubble and void superlattices are outlined.(c) 2021 The Author. Published by Elsevier B.V. This is an open access article under the CC BY license ( http://creativecommons.org/licenses/by/4.0/ )
Balooch, M.Allen, F. I.Popovic, M. P.Hosemann, P....
8页
查看更多>>摘要：The near-surface structural and mechanical changes of tungsten upon exposure to 25 keV helium ions at doses ranging from 1 x 10(2) to 1 x 10(4) ions/nm(2) (1 x 10(16) to 1 x 10(18) ions/cm(2)) are investigated using site-specific implantation, imaging and probing techniques. Helium Ion Microscopy and Atomic Force Microscopy are used to investigate surface topography changes due to swelling and blistering, and nanoindentation is used to study changes in material hardness. An analytical model has been developed to qualitatively explain bubble formation, surface swelling and the variation in local hardness with depth observed experimentally. The model assumes that the implanted helium has a Gaussian depth distribution, which the helium nanobubbles also follow due to their fast formation. There are two competing processes proposed: (1) a reduction in hardness due to the decrease in material density from bubble formation, and (2) an increase in hardness due to bubbles acting as pinning points that impede dislocation motion. (C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The use of single-crystal sapphire optical fibers has been considered to extend fiber-optic sensing to the extreme temperature (> 1000 degrees C) environments encountered in nuclear applications. However, before these sapphire fiber-based sensors can be deployed, their optical transmission and dimensional stability (which impacts drift of some sensors) must be characterized under representative testing conditions. Data regarding the optical transmission of sapphire following high-dose neutron irradiation at temperatures > 100 degrees C is extremely limited. This work provides measurements of optical density (i.e., attenuation) and directional dimensional changes in bulk single-crystal sapphire materials irradiated to a fast neutron fluence of 2.4 x 10(21) n/cm(2) (3.5 displacements per atom) at temperatures ranging from 95 to 688 degrees C. Optical density measured after irradiation at 95 and 298 degrees C showed ultraviolet and visible absorption bands corresponding to known defect centers and temperature trends that were generally consistent with previous ex situ and in situ measurements made at much lower neutron fluence. However, optical density measured after irradiation at 688 degrees C was as much as two orders of magnitude higher, indicating that the fundamental mechanism for radiation-induced attenuation changes at this irradiation temperature. Additional analysis and comparison with previous works suggest that the attenuation may result from void formation, leading to increased Rayleigh scattering losses in the material and increased swelling that would also result in drift of Bragg grating-based sensors in sapphire fibers. These results pose serious questions regarding the feasibility of sapphire fiber-based sensors for high-temperature nuclear applications. (C) 2021 Elsevier B.V. All rights reserved.
NSTL
Elsevier