C. Björkas

Chemical sputtering of Be due to D bombardment

While covalently bonded materials such as carbon are well known to be eroded by chemical sputtering when exposed to plasmas or low-energy ion irradiation, pure metals have been believed to sputter only physically. The erosion of Be when subject to D bombardment was in this work measured at the PISCES-B facility and modelled with molecular dynamics simulations. During the experiments, a chemical effect was observed, since a fraction of the eroded Be was in the form of BeD molecules. This fraction decreased with increasing ion energy. The same trend was seen in the simulations and was explained by the swift chemical sputtering mechanism, showing that pure metals can, indeed, be sputtered chemically. D ions of only 7eV can erode Be through this mechanism.

Interatomic potentials for the Be–C–H system

Analytical bond-order potentials for beryllium, beryllium carbide and beryllium hydride are presented. The reactive nature of the formalism makes the potentials suitable for simulations of non-equilibrium processes such as plasma-wall interactions in fusion reactors. The Be and Be-C potentials were fitted to ab initio calculations as well as to experimental data of several different atomic configurations and Be-H molecule and defect data were used in determining the Be-H parameter set. Among other tests, sputtering, melting and quenching simulations were performed in order to check the transferability of the potentials. The antifluorite Be(2)C structure is well described by the Be-C potential and the hydrocarbon interactions are modelled by the established Brenner potentials.

Atomistic simulations of stainless steels: a many-body potential for the Fe–Cr–C system

Stainless steels found in real-world applications usually have some C content in the base Fe-Cr alloy, resulting in hard and dislocation-pinning carbides-Fe3C (cementite) and Cr23C6-being present in the finished steel product. The higher complexity of the steel microstructure has implications, for example, for the elastic properties and the evolution of defects such as Frenkel pairs and dislocations. This makes it necessary to re-evaluate the effects of basic radiation phenomena and not simply to rely on results obtained from purely metallic Fe-Cr alloys. In this report, an analytical interatomic potential parameterization in the Abell-Brenner-Tersoff form for the entire Fe-Cr-C system is presented to enable such calculations. The potential reproduces, for example, the lattice parameter(s), formation energies and elastic properties of the principal Fe and Cr carbides (Fe3C, Fe5C2, Fe7C3, Cr3C2, Cr7C3, Cr23C6), the Fe-Cr mixing energy curve, formation energies of simple C point defects in Fe and Cr, and the martensite lattice anisotropy, with fair to excellent agreement with empirical results. Tests of the predictive power of the potential show, for example, that Fe-Cr nanowires and bulk samples become elastically stiffer with increasing Cr and C concentrations. High-concentration nanowires also fracture at shorter relative elongations than wires made of pure Fe. Also, tests with Fe3C inclusions show that these act as obstacles for edge dislocations moving through otherwise pure Fe.

Modelling of Radiation Damage in Fe-Cr Alloys

High-Cr ferritic/martensitic steels are being considered as structural materials for a large number of future nuclear applications, from fusion to accelerator-driven systems and GenIV reactors. Fe-Cr alloys can be used as model materials to investigate some of the mechanisms governing their microstructure evolution under irradiation and its correlation to changes in their macroscopic properties. Focusing on these alloys, we show an example of how the integration of computer simulation and theoretical models can provide keys for the interpretation of a host of relevant experimental observations. In particular we show that proper accounting for two basic features of these alloys, namely, the existence of a fairly strong attractive interaction between self-interstitials and Cr atoms and of a mixing enthalpy that changes sign from negative to positive around 8 to 10 % Cr, is a necessary and, to a certain extent, sufficient condition to rationalize and understand their behavior under irradiation. These features have been revealed by ab initio calculations, are supported by experimental evidence, and have been adequately transferred into advanced empirical interatomic potentials, which have been and are being used for the simulation of damage production, defect behavior, and phase transformation in these alloys. The results of the simulations have been and are being used to parameterize models capable of extending the description of radiation effects to scales beyond the reach of molecular dynamics. The present paper intends to highlight the most important achievements and results of this research activity.

ERO code benchmarking of ITER first wall beryllium erosion/re-deposition against LIM predictions

Previous studies (Carpentier et al 2011 J. Nucl. Mater. 415 S165–S169) carried out with the LIM code of the ITER first wall (FW) on beryllium (Be) erosion, re-deposition and tritium retention by co-deposition under steady-state burning plasma conditions have shown that, depending on input plasma parameter assumptions and sputtering yields, the erosion lifetime and fuel retention on some parts of the FW can be a serious concern. The importance of the issue is such that a benchmark of this previous work is sought and has been provided by the ERO code (Pitts et al 2011 J. Nucl. Mater. 415 S957–S964) simulations described in this paper. Provided that inputs to the codes are carefully matched, excellent agreement is found between the erosion/deposition profiles from both codes for a given ITER-shaped FW panel. Issues regarding the difficult problem of the correct treatment of Be sputtering are discussed in relation to the simulations. The possible influence of intrinsic Be impurity is investigated.

The effect of plasma impurities on the sputtering of tungsten carbide

Understanding of sputtering by ion bombardment is needed in a wide range of applications. In fusion reactors, ion impacts originating from a hydrogen-isotope-rich plasma will lead, among other effects, to sputtering of the wall material. To study the effect of plasma impurities on the sputtering of the wall mixed material tungsten carbide molecular dynamics simulations were carried out. Simulations of cumulative D cobombardment with C, W, He, Ne or Ar impurities on crystalline tungsten carbide were performed in the energy range 100-300 eV. The sputtering yields obtained at low fluences were compared to steady state SDTrimSP yields. During bombardment single C atom sputtering was preferentially observed. We also detected significant W(x)C(y) molecule sputtering. We found that this molecule sputtering mechanism is of physical origin.

Damage production in GaAs and GaAsN induced by light and heavy ions

Ion irradiation causes damage in semiconductor crystal structures and affects charge carrier dynamics. We have studied the damage production by high-energy (100keV–10MeV) H, He, Ne, and Ni ions in GaAs and GaAs90N10 using molecular dynamics computer simulations. We find that the heavier Ne and Ni ions produce a larger fraction of damage in large clusters than H and He. These large clusters are either in the form of amorphous zones or (after room-temperature aging or high-temperature annealing) in the form of vacancy and antisite clusters. The total damage production in GaAs and GaAs90N10 is found to be practically the same for all the ions. A clearly smaller fraction of the damage in GaAs90N10 compared to GaAs is in large clusters, however. Our results indicate that experimentally observed differences in charge carrier lifetimes between light and heavy ion irradiations, and before and after annealing, can be understood in terms of the large defect clusters. An increasing amount of damage in large clusters...

Multiscale modelling of plasma-wall interactions in fusion reactor conditions

The interaction of fusion reactor plasma with the material of the first wall involves a complex multitude of interlinked physical and chemical effects. Hence, modern theoretical treatment of it relies to a large extent on multiscale modelling, i.e. using different kinds of simulation approaches suitable for different length and time scales in connection with each other. In this review article, we overview briefly the physics and chemistry of plasma–wall interactions in tokamak-like fusion reactors, and present some of the most commonly used material simulation approaches relevant for the topic. We also give summaries of recent multiscale modelling studies of the effects of fusion plasma on the modification of the materials of the first wall, especially on swift chemical sputtering, mixed material formation and hydrogen isotope retention in tungsten.

Molecules can be sputtered also from pure metals: sputtering of beryllium hydride by fusion plasma–wall interactions

The low-energy erosion mechanisms of first-wall materials subject to a fusion plasma are poorly known theoretically, even though they pose a critical problem for the development of tokamak-like fusion reactors. Using molecular dynamics computer simulations and analytical theory, we have examined the fundamental mechanisms of the erosion of first-wall materials, focusing on molecular release from beryllium surfaces. We show that the observed sputtering of BeD molecules from beryllium when exposed to a D plasma can be explained by the swift chemical sputtering mechanism, and that it also can happen in BeW alloys. This demonstrates that pure metals can, in contrast to conventional wisdom, be sputtered chemically. We also link the simulations of BeD sputtering to the plasma impurity transport code ERO, in order to follow the behavior of sputtered BeD species in a plasma. This multi-scale approach enables direct comparisons with experimental observations of BeD sputtering in the PISCES-B facility.

A Be?W interatomic potential

In this work, an interatomic potential for the beryllium-tungsten system is derived. It is the final piece of a potential puzzle, now containing all possible interactions between the fusion reactor materials beryllium, tungsten and carbon as well as the plasma hydrogen isotopes. The potential is suitable for plasma-wall interaction simulations and can describe the intermetallic Be(2)W and Be(12)W phases. The interaction energy between a Be surface and a W atom, and vice versa, agrees qualitatively with ab initio calculations. The potential can also reasonably describe Be(x)W(y) molecules with x, y = 1, 2, 3, 4.