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In this work, an interatomic potential for the W-Re system is fitted and benchmarked against experimental and density functional theory (DFT) data, of which part are generated in this work. Having in mind studies related to the plasticity of W-Re alloys under irradiation, emphasis is put on fitting point-defect properties, elastic constants, and dislocation properties. The developed potential can reproduce the mechanisms responsible for the experimentally observed softening, i.e., decreasing shear moduli, decreasing Peierls barrier, and asymmetric screw dislocation core structure with increasing Re content in W-Re solid solutions. In addition, the potential predicts elastic constants in reasonable agreement with DFT data for the phases forming non-coherent precipitates (r- and v-phases) in W-Re alloys. In addition, the mechanical stability of the different experimentally observed phases is verified in the temperature range of interest (700-1500 K). As a conclusion, the presented potential provides an excellent tool to study plasticity in W-Re alloys at the atomic level. [ABSTRACT FROM AUTHOR]

• Several commercial W grades are compared. • Application of cross-rolling step offers some advantageous properties. • All studied grades exhibit similar embrittlement under neutron irradiation. Tungsten is plasma-facing material (PFM) for the divertor and test blanket module to be deployed in ITER. During operation, sub-components made of tungsten will be subjected to cyclic heat exposure, plasma particles loads and bombardment by fast neutrons. The optimization of tungsten grades, in terms of mechanical properties, hopefully offering enhanced resistance against high heat flux and irradiation damage is ongoing, and recently A.L.M.T. has proposed an extra cross-rolling step for the industrial fabrication of thick tungsten plates. Here, we present the results of recent experiments aimed at characterizing the impact of high flux plasma load, thermal shock and neutron irradiation on the damage induced in newly released grades and conventional ITER specification tungsten. The assessment included the tensile mechanical properties before and after the irradiation as well as microstructural investigation of surface modification due to the high heat flux testing. The neutron irradiation and high heat flux tests were performed in a wide range of temperatures (400 - 1200 °C) and covered typical exposure conditions that will be met by PFM in the ITER divertor. [ABSTRACT FROM AUTHOR]

• Fusion-specific irradiation station in MYRRHA facilities is proposed. • It is a versatile irradiation facility with a fast spectrum of neutrons. • Facility is in line with the objective of increasing the TRL of fusion materials and components. One of the main objectives of the EUROfusion Roadmap to Fusion Energy during Horizon Europe is to finalize the conceptual design (thus preparing the engineering design phase) and initiate the construction of a fusion spectrum neutron source (IFMIF-DONES). The Belgian Nuclear Research Center is currently developing the MYRRHA research facilities, which include a subcritical accelerator-driven system ignited by a 600 MeV superconducting linear proton accelerator coupled to a 100 MeV proton injector. The implementation of the multipurpose fast spectrum reactor MYRRHA will proceed in phases, with phase I, covering 2019–2027, including the deployment of the 100 MeV proton accelerator and associated fusion material research facilities. The research on fusion materials on the basis of the 100 MeV proton beam and with hot cell facilities will focus on innovative materials and composites and will be fully complementary to IFMIF-DONES as well to the planned research program on using currently available nuclearized facilities such as BR2 (Belgian Research Reactor 2) and Hot Cells for mechanical, microstructural and chemical analysis. A bespoke fusion irradiation module is also foreseen on the MYRRHA reactor, the construction of which is planned in the third phase of the implementation of the project (i.e. after 2035). Provisions have been made in the core of the MYRRHA reactor to include so-called In-Pile Sections around the central core or at the central core position below the spallation target when operating in subcritical mode. In this contribution, we review the role and technical capabilities that MYRRHA facilities may provide in the framework of the European Fusion Electricity Roadmap and specifically for DEMO Engineering Design, Material's Development and enhancement of Technology Readiness Level. Aspects of education and training, being an integral part of the European nuclear fusion program, are also touched upon. Finally, the synchronization of the timelines for the implementation of MYRRHA and the EU fusion facilities are discussed. [ABSTRACT FROM AUTHOR]

The Belgian Nuclear Research Centre is currently developing MYRRHA - a subcritical accelerator driven system ignited by a 600 MeV superconducting linear proton accelerator, cooled with an eutectic lead bismuth. The implementation of MYRRHA will take place in phases, where phase I, covering 2019–2027, includes the deployment of the 100 MeV proton accelerator and beam-added research facilities. Development of the dedicated facilities to enable irradiation of fusion materials and components is in the portfolio of this R&D block. The baseline concept of the 100 MeV fusion-dedicated facilities has been presented for the first time in 2019 and in this contribution, we extend its description and provide main elements of the conceptual design of the irradiation installations, preliminary configuration of the hot cell facilities (required to load/unload the irradiation modules), information on the irradiation envelope. In addition, we summarize the results of the computational study performed to investigate the equivalence (in terms of damage rate, nuclear transmutation, temperatures and expected irradiation-induced microstructure) of the irradiation conditions at MYRRHA facilities in comparison with other material test facilities such as IFMIF and ESS, and in DEMO as well as in the mixed-spectrum material test reactors. The aspects related to miniaturization of the sample geometries are also discussed to substantiate the feasibility and relevance of this facility for fusion material's R&D programme. [Display omitted] The Belgian Nuclear Research Centre is currently developing MYRRHA - a subcritical accelerator driven system ignited by a 600 MeV superconducting linear proton accelerator, cooled with an eutectic lead bismuth. The implementation will take place in phases, where phase I, covering 2019–2026, includes the deployment of the 100 MeV proton accelerator. Development of the dedicated facilities to enable irradiation of fusion materials and components is in the portfolio of this R&D block. In this contribution, we provide main elements of the conceptual design of the irradiation installations, preliminary configuration of the hot cell facilities (required to load/unload the irradiation modules), information on the irradiation envelope and specific features of irradiation conditions. [ABSTRACT FROM AUTHOR]

Σ 3 { 112 } grain boundary (GB) is known to exhibit plastic slip in bcc metals, which overall contributes to the good ductility and toughness of such metals as iron, matrix element for the steels applied in nuclear industry. During nuclear operation, the neutron irradiation causes formation of point defects and their clusters which lead to the non-equilibrium mass transport and agglomeration of nanoscale defects inside grains and near the grain boundaries. Due to the effect of chemical segregation, GBs can encounter sessile obstacles preventing the GB plastic slip. Understanding the influence of the irradiation defects segregated at the grain boundaries is important for the development of structural steels for nuclear fusion and fission applications. The present paper studies the influence of the irradiation defects, such as Cr precipitates, He bubbles and voids on the plastic slip of the GB interface. The studied defects are found to act as strong obstacles to the glide of the elementary disconnections responsible for the shear-coupled GB migration. The interaction between a single elementary disconnection and a series of them provided by a source has been studied. As a result of the interaction with impenetrable defects the formation of residual defects such as loops is observed. [Display omitted] • Irradiation defects affect the propagation of the {112} GB disconnections. • The irradiation defects at the GB should lead to the hardening of the material. • Soft obstacles are sheared by the interaction with disconnections. • Interaction between disconnections and hard obstacles leaves behind residual defects. • The disconnections interaction mechanisms are similar to the ones for dislocations. [ABSTRACT FROM AUTHOR]

The degradation of austenitic stainless steels in a radiation environment is a known problem for the in-core components of nuclear light water reactors. For a better understanding of the prevailing mechanisms responsible for the materials' degradation, large-scale atomistic simulations are desirable. In this framework and as a follow-up on Bonny et al (2011 Modelling Simul. Mater. Sci. Eng.19 085008), we developed an embedded atom method type interatomic potential for the ternary FeNiCr system to model the production and evolution of radiation defects. Special attention has been drawn to the Fe10Ni20Cr alloy, whose properties were ensured to be close to those of 316L austenitic stainless steels. The potential is extensively benchmarked against density functional theory calculations and the potential developed in our earlier work. As a first validation, the potential is used in AKMC simulations to simulate thermal annealing experiments in order to determine the self-diffusion coefficients of the components in FeNiCr alloys around the Fe10Ni20Cr composition. The results from these simulations are consistent with experiments, i.e., DCr > DNi > DFe. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Interstitial dislocation loops with Burgers vector of [image omitted] type are observed in α-iron irradiated by neutrons or heavy ions, and their population increases with increasing temperature. Their effect on motion of a [image omitted] edge dislocation was reported earlier 1. Results are presented of a molecular dynamics study of interactions between a [image omitted] screw dislocation and [image omitted] loops in iron at temperature in the range 100 to 600 K. A variety of reaction mechanisms and outcomes are observed and classified in terms of the resulting dislocation configuration and the maximum stress required for the dislocation to break away. The highest obstacle resistance arises when the loop is absorbed to form a helical turn on the screw dislocation line, for the dislocation cannot glide away until the turn closes and a loop is released with the same Burgers vector as the line. Other than one situation found, in which no dislocation-loop reaction occurs, the weakest obstacle strength is found when the original [image omitted] loop is restored at the end of the reaction. The important role of the cross-slip and the influence of model boundary conditions are emphasised and demonstrated by examples. [ABSTRACT FROM AUTHOR]

Abstract: Atomistic simulations are used to study how thermal ageing or radiation-enhanced/induced microstructural changes affect the motion of an edge dislocation in Fe–Cr alloys. Monte Carlo (MC) techniques were used to obtain distributions of Cr atoms in Fe–Cr alloys containing from 5% to 12%Cr, with an ab initio-based interatomic potential correctly reproducing the thermodynamic properties of Fe–Cr alloys in the Fe-rich region. The edge dislocation motion was then modelled using molecular dynamics, by applying constant strain-rate and the corresponding flow-stress was monitored, in Fe and Fe–Cr defect-free matrices with random and MC-obtained Cr distributions, as well as in matrices containing Cr precipitates of different sizes and radiation defects (voids and dislocation loops). We show that the flow-stress in concentrated Fe–Cr alloys is much higher than in dilute ones. The presence of Cr precipitates significantly contributes to hardening and their strength increases linearly with size. The critical stress to shear voids and Cr precipitates is higher in Fe–Cr than in pure Fe. [Copyright &y& Elsevier]

• Neutron irradiation induced severe embrittlement in pure Cr and Cr-10W. • In pure Cr, the neutron irradiation does not reduce the fracture stress. • In Cr-10W, the neutron irradiation reduces fracture stress. Design of plasma-facing components (PFC) for DEMO divertor unravels new challenges to be met by the in-vessel materials. Embrittlement induced by 14 MeV neutrons in the baseline first wall material - tungsten (W) endangers structural integrity of PFCs. Chromium (Cr) and/or Cr-W alloy has been considered as a candidate material in the design of mid heat flux PFCs as structural body of the monoblock. In our preceding work, the potential of the vacuum arc melting for the fabrication of pure Cr and Cr-W solid solution (with 10 at.%) has been demonstrated. The pure Cr exhibits ductile deformation above 50°C, while the ductile to brittle transition in Cr-10%W is at ~300°C. The particular advantage of the alloying was seen in the strong increase of the work hardening capacity as well as much higher fracture strength compared to pure Cr. Moreover, the solid solution formation as a result of alloying might improve the resistance of the structural materials against irradiation, as was shown for the case of iron-chromium alloys/steels. In this work, the mechanical properties of two pure Cr grades and Cr-10at.%W alloy are assessed after neutron irradiation performed in the conditions relevant for the application of PFCs. The mechanical tests are performed over a wide temperature range to investigate the strength at the irradiation temperature as well as to clarify the role of limited defect annealing. The neutron irradiation resulted in remarkable degradation of mechanical properties making all the studied materials brittle at the applied irradiation temperature. In the case of Cr-10at.%W, the irradiation caused such severe modifications that the fracture strength has become lower than in non-irradiated state. The fracture surface analysis is made to demonstrate the nature of the brittle fracture and to discuss possible reasons for the observed brittleness. [ABSTRACT FROM AUTHOR]

• Annealing at 800 °C increases loop and void densities and moderately increases the mean sizes. • Annealing at 1000 °C reduces the loop and voids densities, while the mean sizes increase. • The applied annealing did not have any effect on the void denuded zones near grain boundaries. • High temperature irradiation is more efficient than post-irradiation annealing to reduce defect formation. In this paper, we present a transmission electron microscopy study of three tungsten grades subjected to neutron irradiation and post-irradiation annealing. The irradiation was performed at 600 °C up to 0.2 dpa and the microstructure was investigated after the irradiation, reported in a previous publication, as well as after the annealing at 800 °C and 1000 °C. The evolution of the radiation induced defects is characterized and the overall microstructural features are compared in three states: virgin-as irradiated – as annealed after irradiation. The annealing at 800 °C causes an increase in the loop and void densities and moderate increase in the mean sizes. An increase of the annealing temperature to 1000 °C causes a reduction of the loop density, while the loop size increases. Also the void density reduces, but not necessarily the void fraction. The peak of the void volume fraction is found around 800 °C, which coincides with the earlier reported experimental data on the void swelling peak under neutron irradiation. [ABSTRACT FROM AUTHOR]

The expected degradation of mechanical properties of structural materials under irradiation (i.e. "in operation") in nuclear components represents a significant challenge for the design to account the impact of long-term irradiation effects. Therefore, a development of scientific and engineering expertise to understand and possibly control the severe impact of harsh neutron irradiation on materials is one of the tasks in the current material's research agenda. As it is well known from long-standing experience with fission systems, radiation embrittlement is caused by nano-scale features that obstruct plasticity mediated by dislocations. The present contribution highlights recent research and development efforts addressed towards the assessment of the interrelation between nano-structural features and hardening induced by neutron irradiation in fusion structural high-Cr ferritic/martensitic steels. In this work, the impact of doping by three chemical elements (nickel, silicon and phosphorus) on the microstructural evolution under neutron irradiation at 300 °C and 450 °C is studied experimentally as well as by computer simulations. The evolution of microstructure is assessed by transmission electron microscopy and object kinetic Monte Carlo simulations. The latter method utilizes a new approach enabling to follow both nano-scale irradiation defects and micro-segregation zones causing the formation of solute rich clusters, detectable by atom probe tomography. The results of the present work clearly point out that Ni–Si segregation alters the spatial and size distribution of dislocations loops already at as low dose as 0.1 dpa. The physical mechanisms behind the impact of minor alloying elements are discussed. [ABSTRACT FROM AUTHOR]

We have performed proton irradiation of W and W-5wt.%Ta materials at 350 °C with a step-wise damage level increase up to 0.7 dpa and using two beam energies, namely 40 keV and 3 MeV, in order to probe the accumulation of radiation-induced lattice damage in these materials. Interstitial-type a /2 <111> dislocation loops are formed under irradiation, and their size increases in W-5Ta up to a loop width of 21 ± 4 nm at 0.3 dpa, where loop saturation takes place. In contrast, the loop length in W increases progressively up to 183 ± 50 nm at 0.7 dpa, whereas the loop width remains relatively constant at 29 ± 7 nm at >0.3 dpa, giving rise to dislocation strings. The dislocation loops and tangles are observed in both materials examined after a 3 MeV proton irradiation at 350 °C. Ta doping delays the evolution of radiation-induced dislocation structures in W, and can consequently impact the hydrogen isotope retention under plasma exposure. [ABSTRACT FROM AUTHOR]

A relative contribution to irradiation hardening caused by dislocation loops and solute-rich precipitates is established for RPV steels of WWER-440 and WWER-1000 reactors, based on TEM measurements and mechanical testing at reactor operating temperature of 563 K. The pinning strength factors evaluated for loops and precipitates are shown to be much lower than those obtained for model alloys based on the room temperature testing as well as those evaluated by means of atomistic simulations in the temperature range of 300–600 K. This discrepancy is explained in the framework of a model of thermally activated dislocation motion, which takes into account the difference in temperature and strain rate employed in atomistic simulations and in mechanical testing. [ABSTRACT FROM AUTHOR]

Abstract: Object Kinetic Monte Carlo models allow for the study of the evolution of the damage created by irradiation to time scales that are comparable to those achieved experimentally. Therefore, the essential Object Kinetic Monte Carlo parameters can be validated through comparison with experiments. However, this validation is not trivial since a large number of parameters is necessary, including migration energies of point defects and their clusters, binding energies of point defects in clusters, as well as the interaction radii. This is particularly cumbersome when describing an alloy, such as the Fe–Cr system, which is of interest for fusion energy applications. In this work we describe an Object Kinetic Monte Carlo model for Fe–Cr alloys in the dilute limit. The parameters used in the model come either from density functional theory calculations or from empirical interatomic potentials. This model is used to reproduce isochronal resistivity recovery experiments of electron irradiated dilute Fe–Cr alloys performed by Abe and Kuramoto. The comparison between the calculated results and the experiments reveal that an important parameter is the capture radius between substitutional Cr and self-interstitial Fe atoms. A parametric study is presented on the effect of the capture radius on the simulated recovery curves. [Copyright &y& Elsevier]

Abstract: The objective of the FP6 Perfect Project was to develop a first example of integrated multiscale computational models, capable of describing the effects of irradiation in nuclear reactor components, namely vessel and internals. The use of ab initio techniques to study, in the most reliable way currently possible, atomic-level interactions between species and defects, and the transfer of this knowledge to interatomic potentials, of use for large scale dynamic simulations, lie at the core of this effort. The target materials of the Project were bainitic steels (vessel) and austenitic steels (internals), i.e. iron alloys. In this article, the advances made within the Project in the understanding of defect properties in Fe alloys, by means of ab initio calculations, and in the development of interatomic potentials for Fe and Fe alloys are overviewed, thereby providing a reference basis for further progress in the field. Emphasis is put in showing how the produced data have enhanced our level of understanding of microstructural processes occurring under irradiation in model alloys and steels used in existing nuclear power plants. [ABSTRACT FROM AUTHOR]