Browsing by Subject "RADIATION-DAMAGE"

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  • Dudarev, Sergei L.; Mason, Daniel R.; Tarleton, Edmund; Ma, Pui-Wai; Sand, Andrea E. (2018)
    Predicting strains, stresses and swelling in nuclear power plant components exposed to irradiation directly from the observed or computed defect and dislocation microstructure is a fundamental problem of fusion power plant design that has so far eluded a practical solution. We develop a model, free from parameters not accessible to direct evaluation or observation, that is able to provide estimates for irradiation-induced stresses and strains on a macroscopic scale, using information about the distribution of radiation defects produced by high-energy neutrons in the microstructure of materials. The model exploits the fact that elasticity equations involve no characteristic spatial scale, and hence admit a mathematical treatment that is an extension to that developed for the evaluation of elastic fields of defects on the nanoscale. In the analysis given below we use, as input, the radiation defect structure data derived from ab initio density functional calculations and large-scale molecular dynamics simulations of high-energy collision cascades. We show that strains, stresses and swelling can be evaluated using either integral equations, where the source function is given by the density of relaxation volumes of defects, or they can be computed from heterogeneous partial differential equations for the components of the stress tensor, where the density of body forces is proportional to the gradient of the density of relaxation volumes of defects. We perform a case study where strains and stresses are evaluated analytically and exactly, and develop a general finite element method implementation of the method, applicable to a broad range of predictive simulations of strains and stresses induced by irradiation in materials and components of any geometry in fission or fusion nuclear power plants.
  • Zhang, Shuo; Pakarinen, Olli Heikki; Backholm, Matilda; Djurabekova, Flyura; Nordlund, Kai; Keinonen, Juhani; Wang, T.S. (2018)
    In this work, we first simulated the amorphization of crystalline quartz under 50 keV Na-23 ion irradiation with classical molecular dynamics (MD). We then used binary collision approximation algorithms to simulate the Rutherford backscattering spectrometry in channeling conditions (RBS-C) from these irradiated MD cells, and compared the RBS-C spectra with experiments. The simulated RBS-C results show an agreement with experiments in the evolution of amorphization as a function of dose, showing what appears to be (by this measure) full amorphization at about 2.2 eV.atom(-1). We also applied other analysis methods, such as angular structure factor, Wigner-Seitz, coordination analysis and topological analysis, to analyze the structural evolution of the irradiated MD cells. The results show that the atomic-level structure of the sample keeps evolving after the RBS signal has saturated, until the dose of about 5 eV.atom(-1). The continued evolution of the SiO2 structure makes the definition of what is, on the atomic level, an amorphized quartz ambiguous.
  • Mason, D. R.; Sand, A. E.; Dudarev, S. L. (2019)
    We describe the development of a new object kinetic Monte Carlo (kMC) code where the elementary defect objects are off-lattice atomistic configurations. Atomic-level transitions are used to transform and translate objects, to split objects and to merge them together. This gradually constructs a database of atomic configurations-a set of relevant defect objects and their possible events generated on-the-fly. Elastic interactions are handled within objects with empirical potentials at short distances, and between spatially distinct objects using the dipole tensor formalism. The model is shown to evolve mobile interstitial clusters in tungsten faster than an equivalent molecular dynamics (MD) simulation, even at elevated temperatures. We apply the model to the evolution of complex defects generated using MD simulations of primary radiation damage in tungsten. We show that we can evolve defect structures formed in cascade simulations to experimentally observable timescales of seconds while retaining atomistic detail. We conclude that the first few nanoseconds of simulation following cascade initiation would be better performed using MD, as this will capture some of the near-temperature-independent evolution of small highly-mobile interstitial clusters. For the 20keV cascade annealing simulations considered here, we observe internal relaxations of sessile objects. These relaxations would be difficult to capture using conventional object kMC, yet are important as they establish the conditions for long timescale evolution.
  • Sand, A.E.; Byggmästar, J.; Zitting, A.; Nordlund, K. (2018)
    Most experimental work on radiation damage is performed to fairly high doses, where cascade overlap effects come into play, yet atomistic simulations of the primary radiation damage have mainly been performed in initially perfect lattice. Here, we investigate the primary damage produced by energetic ion or neutron impacts in bcc Fe and W. We model irradiation effects at high fluence through atomistic simulations of cascades in pre-damaged systems. The effects of overlap provide new insights into the processes governing the formation under irradiation of extended defects. We find that cascade overlap leads to an increase in the numbers of large clusters in Fe, while in W such an effect is not seen. A significant shift in the morphology of the primary damage is also observed, including the formation of complex defect structures that have not been previously reported in the literature. These defects are highly self-immobilized, shifting the damage away from the predominance of mobile 1/2〈111〉 loops towards more immobile initial configurations. In Fe, where cascade collapse is extremely rare in molecular dynamics simulations of individual cascades, we observe the formation of vacancy-type dislocation loops from cascade collapse as a result of cascade overlap.
  • Mason, D. R.; Sand, A. E.; Yi, X.; Dudarev, S. L. (2018)
    Recently we have presented direct experimental evidence for large defect clusters being formed in primary damage cascades in self-ion irradiated tungsten [Yi et al., EPL 110:36001 (2015)]. This large size is significant, as it implies that strong elastic interaction between the defects will affect their subsequent evolution, especially if defects are formed close together. Here we present a direct experimental observation of the separation between visible defects in self-ion irradiated tungsten, in the form of a 2d pairwise radial distribution function extracted from transmission electron micrographs (TEM). We also present a detailed analysis of the observed radial distribution function, and infer the probable size and shape of individual cascades. We propose and validate a simple exponential form for the spatial distribution of defects within a single cascade. The cascade statistics necessary have been acquired by developing an automated procedure for analysing black-dot damage in TEM micrographs. We confirm that the same model also produces a high-quality fit to the separation between larger defects observed in MD simulations. For the first time we present experimental evidence for the sub-nanometre-scale spatial distribution of defect clusters within individual cascades. (C) 2017 EURATOM. Published by Elsevier Ltd on behalf of Acta Materialia Inc. All rights reserved.
  • Bonny, G.; Castin, N.; Bakaev, A.; Sand, A. E.; Terentyev, D. (2020)
    In recent years, a number of systematic investigations of high-energy collision cascades in tungsten employing advanced defect analysis tools have shown that interstitial clusters can form complex non-planar dislocation structures. These structures are sessile in nature and may potentially have a strong impact on the long-term evolution of the radiation microstructure. To clarify this aspect, we selected several representative primary damage states of cascades debris and performed annealing simulations using molecular dynamics (MD). We found that immobile complexes of non-planar dislocation structures (CDS) evolve into glissile and planar shaped 1/2 <1 1 1 > loops with an activation energy of similar to 1.5 eV. The CDS objects were implemented in an object kinetic Monte Carlo (OKMC) model accounting for the event of transformation into 1-D migrating loops, following the MD data. OKMC was then used to investigate the impact of the transformation event (and the associated activation energy) on the evolution of the microstructure.
  • Byggmästar, J.; Granberg, F.; Nordlund, K. (2018)
    Recent work has shown that the repulsive part of the interatomic potential at intermediate atomic separations strongly affects the extent and morphology of the damage produced by collision cascades in molecular dynamics simulations. Here, we modify an existing embedded atom method interatomic potential for iron to more accurately reproduce the threshold displacement energy surface as well as the many-body repulsion at intermediate and short interatomic distances. Using the modified potential, we explore the effects of an improved repulsive potential on the primary damage production and the cumulative damage accumulation in iron. We find that the extent of the damage produced by single cascades, in terms of surviving Frenkel pairs, directly correlates with the change in threshold displacement energies. On the other hand, the damage evolution at higher doses is more dependent on the formation and stability of different defect clusters, defined by the near-equilibrium part of the interatomic potential.
  • Lu, Eryang; Zhao, Junlei; Makkonen, Ilja; Mizohata, Kenichiro; Li, Zhiming; Hua, Mengyuan; Djurabekova, Flyura; Tuomisto, Filip (2021)
    We present evidence of homogenization of atomic diffusion properties caused by C and N interstitials in an equiatomic single-phase high entropy alloy (FeMnNiCoCr). This phenomenon is manifested by an unexpected interstitial-induced reduction and narrowing of the directly experimentally determined migration barrier distribution of mono-vacancy defects introduced by particle irradiation. Our observation by positron annihilation spectroscopy is explained by state-of-the-art theoretical calculations that predict preferential localization of C/N interstitials in regions rich in Mn and Cr, leading to a narrowing and reduction of the mono-vacancy size distribution in the random alloy. This phenomenon is likely to have a significant impact on the mechanical behavior under irradiation, as the local variations in elemental motion have a profound effect on the solute strengthening in high entropy alloys. (C) 2021 The Authors. Published by Elsevier Ltd on behalf of Acta Materialia Inc.
  • Slotte, Jonatan; Kilpeläinen, Simo; Segercrantz, Natalie; Mizohata, Kenichiro; Räisänen, Jyrki; Tuomisto, Filip (2021)
    A unique experimental setup at the Accelerator Laboratory of the University of Helsinki enables in situ positron annihilation spectroscopy (PAS) analysis on ion irradiated samples. In addition, the system enables temperature control (10-300 K) of the sample both during irradiation and during subsequent positron annihilation measurements. Using such a system for defect identification and annealing studies comes with a plethora of possibilities for elaborate studies. However, the system also poses some restrictions and challenges to these possibilities, both related to irradiation and to the PAS analysis. This review tries to address these issues.
  • Granberg, F.; Byggmastar, J.; Nordlund, K. (2021)
    Tungsten has been chosen as the plasma-facing wall material in fusion reactors, due to its high density and melting point. The wall material will not only be sputtered at the surface, but also damaged deep inside the material by energetic particles. We investigate the high-dose damage production and accumulation by computational means using molecular dynamics. We observe that the choice of interatomic potential drastically affects the evolution. The structure and stability of the obtained defect configurations are validated using a quantum-accurate Gaussian approximation potential. (c) 2021 The Author(s). Published by Elsevier B.V. This is an open access article under the CC BY license ( )
  • Phillips, N. W.; Yu, H.; Das, S.; Yang, D.; Mizohata, K.; Liu, W.; Xu, R.; Harder, R. J.; Hofmann, F. (2020)
    Developing a comprehensive understanding of the modification of material properties by neutron irradiation is important for the design of future fission and fusion power reactors. Self-ion implantation is commonly used to mimic neutron irradiation damage, however an interesting question concerns the effect of ion energy on the resulting damage structures. The reduction in the thickness of the implanted layer as the implantation energy is reduced results in the significant quandary: Does one attempt to match the primary knock-on atom energy produced during neutron irradiation or implant at a much higher energy, such that a thicker damage layer is produced? Here we address this question by measuring the full strain tensor for two ion implantation energies, 2 MeV and 20 MeV in self-ion implanted tungsten, a critical material for the first wall and divertor of fusion reactors. A comparison of 2 MeV and 20 MeV implanted samples is shown to result in similar lattice swelling. Multi-reflection Bragg coherent diffractive imaging (MBCDI) shows that implantation induced strain is in fact heterogeneous at the nanoscale, suggesting that there is a non-uniform distribution of defects, an observation that is not fully captured by micro-beam Laue diffraction. At the surface, MBCDI and high-resolution electron back-scattered diffraction (HR-EBSD) strain measurements agree quite well in terms of this clustering/non-uniformity of the strain distribution. However, MBCDI reveals that the heterogeneity at greater depths in the sample is much larger than at the surface. This combination of techniques provides a powerful method for detailed investigation of the microstructural damage caused by ion bombardment, and more generally of strain related phenomena in micro-volumes that are inaccessible via any other technique. (C) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.
  • Mason, Daniel R.; Granberg, Fredric; Boleininger, Max; Schwarz-Selinger, Thomas; Nordlund, Kai; Dudarev, Sergei L. (2021)
    Hydrogen isotopes are retained in plasma-facing fusion materials, triggering hydrogen embrittlement and changing tritium inventory as a function of exposure to neutron irradiation. But modeling highly damaged materials-exposed to over 0.1 displacements per atom (dpa)-where saturation of damage is often observed, is difficult because a microstructure containing high density of defects evolves nonlinearly as a function of dose. In this study we show how to determine the defect and hydrogen isotope content in tungsten exposed to high irradiation dose, using no adjustable or fitting parameters. First, we generate converged high dose (>1 dpa) microstructures, using a combination of the creation-relaxation algorithm and collision cascade simulations. Then we make robust estimates of vacancy and void regions using a modified Wigner-Seitz decomposition. The resulting estimates of the void surface area enable predicting the deuterium retention capacity of tungsten as a function of radiation exposure. The predictions are compared to 3He nuclear reaction analysis measurements of tungsten samples, self-irradiated at 290 K to different damage doses and exposed to low-energy deuterium plasma at 370 K. The theory gives an excellent match to the experimental data, with both model and experiment showing that 1.5-2.0 at.% deuterium is retained in irradiated tungsten in the limit of high dose.
  • Marian, Jaime; Becquart, Charlotte S.; Domain, Christophe; Dudarev, Sergei L.; Gilbert, Mark R.; Kurtz, Richard J.; Mason, Daniel R.; Nordlund, Kai; Sand, Andrea E.; Snead, Lance L.; Suzudo, Tomoaki; Wirth, Brian D. (2017)
    Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural and plasma-facing materials will be exposed to unprecedented conditions of irradiation, heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements with carefully validated predictions. For plasma-facing components such as the first wall and the divertor, tungsten (W) has been selected as the leading candidate material due to its superior high-temperature and irradiation properties, as well as for its low retention of implanted tritium. In this paper we provide a review of recent efforts in computational modeling of W both as a plasma-facing material exposed to He deposition as well as a bulk material subjected to fast neutron irradiation. We use a multiscale modeling approach-commonly used as the materials modeling paradigm-to define the outline of the paper and highlight recent advances using several classes of techniques and their interconnection. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining challenges and future research directions.
  • Mason, Daniel R.; Duc Nguyen-Manh; Marinica, Mihai-Cosmin; Alexander, Rebecca; Sand, Andrea E.; Dudarev, Sergei L. (2019)
    The low-energy structures of irradiation-induced defects in materials have been studied extensively over several decades, as these determine the available modes by which a defect can diffuse or relax, and how the microstructure of an irradiated material evolves as a function of temperature and time. Consequently, many studies concern the relative energies of possible defect structures, and empirical potentials are commonly fitted to or evaluated with respect to these. But recently [S. L. Dudarev et al., Nucl. Fusion 58, 126002 (2018)], we have shown that other parameters of defects not directly related to defect energies, namely, their elastic dipole tensors and relaxation volumes, determine the stresses, strains, and swelling of reactor components under irradiation. These elastic properties of defects have received comparatively little attention. In this study, we compute relaxation volumes of irradiation-induced defects in tungsten using empirical potentials and compare to density functional theory results. Different empirical potentials give different results, but some clear potential-independent trends can be identified. We show that the relaxation volume of a small defect cluster can be predicted to within 10% from its point-defect count. For larger defect clusters, we provide empirical fits as a function of defect cluster size. We demonstrate that the relaxation volume associated with a single primary-damage cascade can be estimated from the primary knock-on atom energy. We conclude that while annihilation of defects invariably reduces the total relaxation volume of the cascade debris, there is still no conclusive verdict about whether coalescence of defects reduces or increases the total relaxation volume. Published under license by AIP Publishing.
  • Levo, Emil; Granberg, Fredric; Nordlund, Kai; Djurabekova, Flyura (2021)
    Multiprincipally designed concentrated solid solution alloys, such as high entropy alloys (HEA) and equiatomic multi-component alloys (EAMC-alloys) have shown much promise for use as structural components in future nuclear energy production concepts. The irradiation tolerance in these novel alloys has been shown to be superior to that in more conventional metals used in current nuclear reactors. However, studies involving irradiation of HEAs and EAMC-alloys have usually been performed at room temperature. Hence, in this study the irradiation damage is investigated computationally in two different Ni-based EAMC-alloys and pure Ni at four different temperatures, ranging from 138 K to 800 K. The irradiation damage was studied by analyzing point defects, defect cluster sizes and dislocation networks in the materials. Dislocation loop mobility calculations were performed to help understanding the formation of different dislocation networks in the irradiated materials. Utilizing the knowledge of the depth distribution of damage, and using simulations of Rutherford backscattering in channeling conditions (RBS/c), we can relate our results to experimental data. The main findings are that the alloys have superior irradiation tolerance at all temperatures as compared to pure Ni, and that the damage is reduced in all materials with an increase in temperature.
  • Castin, N.; Dubinko, A.; Bonny, G.; Bakaev, A.; Likonen, J.; De Backer, A.; Sand, A.E.; Heinola, K.; Terentyev, D. (2019)
    The microstructure changes taking place in W under irradiation are governed by many factors, amongst which C impurities and their interactions with self-interstitial atoms (SIA). In this work, we specifically study this effect by conducting a dedicated 2-MeV self-ions irradiation experiment, at room temperature. Samples were irradiated up to 0.02, 0.15 and 1.2 dpa. Transmission electron microscopy (TEM) expectedly revealed a large density of SIA loops at all these doses. Surprisingly, however, the loop number density increased in a non-monotonous manner with the received dose. Performing chemical analysis with secondary ion spectroscopy measurements (SIMS), we find that our samples were likely contaminated by C injection during the irradiation. Employing an object kinetic Monte Carlo (OKMC) model for microstructure evolution, we demonstrate that the C injection is the likely factor explaining the evolution of loops number density. Our findings highlight the importance of the well-known issue of C injection during ion irradiation experiments, and demonstrate how OKMC models can help to rationalize this effect.