A. Kirschner

Simulation of the plasma-wall interaction in a tokamak with the Monte Carlo code ERO-TEXTOR

The interaction of plasma with the walls has been one of the critical issues in the development of fusion energy research. On the one hand, plasma induced erosion can seriously limit the lifetime of the wall components, while, on the other hand, eroded particles can be transported into the core plasma where they lead to dilution of the fusion plasma and to energy losses due to radiation. Low-Z wall materials induce only small radiation losses in the plasma core but suffer from large physical sputtering rates. Carbon based materials in addition suffer from chemically induced erosion. High-Z wall materials show significantly smaller erosion but lead to large radiation losses. One of the main goals of present plasma-wall studies is to find a special choice of wall materials for steady state plasma scenarios that will provide an optimum with respect to fuel dilution, radiation losses, wall lifetime and fuel inventory in the walls. To obtain a better understanding of the processes and to estimate the plasma-wall interaction behaviour in future fusion devices the 3-D Monte Carlo code ERO-TEXTOR, based originally on the ERO code, has been developed. It models the plasma-wall interaction and transport processes in the vicinity of a surface positioned in the boundary layer of TEXTOR. The main aim is to simulate the erosion and redeposition behaviour of different wall materials under various plasma conditions and to compare this with experimental results. This contribution describes the main features of the ERO-TEXTOR code and gives some examples of simulation calculations to illustrate the application of the code.

Beryllium Migration in JET ITER-like Wall Plasmas

JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.

Effect of surface roughness and substrate material on carbon deposition in tokamak TEXTOR

The technique of 13 CH4 tracer injection through test limiters was applied to study the influence of surface roughness and substrate material on the local 13 C deposition. Spherically shaped graphite limiters with integrated gas injection were prepared with two different grades of surface roughness, ~0.1 μm and ~1 μm. In addition, tungsten limiters with a roughness of ~0.1 μm were used to study the substrate material effect. The limiters were exposed to the SOL plasma for two discharge scenarios, Ohmic and neutral beam heated. The 13 C deposition efficiency - the ratio of the locally deposited to the injected amount of 13 C - and the deposition pattern were evaluated by post-mortem surface analysis. Surface roughness has a pronounced effect and increases the 13 C deposition efficiency on graphite with rougher surface by a factor of 3-5 compared to smoother graphite. On tungsten a factor of 2-4 less carbon is deposited than on graphite with similar surface roughness. A systematically higher amount of 13 C (by a factor 1.5-2.5) was deposited on limiters exposed to Ohmic compared to neutral beam heated plasmas.

Modelling of impurity deposition in gaps of castellated surfaces with the 3D-GAPS code

The Monte-Carlo neutral transport code 3D-GAPS is described. The code models impurity transport and deposition in remote areas, such as gaps between cells of castellated plasma-facing surfaces. A step-by-step investigation of the interplay of different processes that may influence the deposition inside gaps, namely particle reflection, elastic neutral collisions, different particle sources, chemical erosion and plasma penetration into gaps, is presented. Examples of modelling results in application to the TEXTOR experiment with a castellated test limiter are provided. It is shown that only with the assumption of the presence of species with different reflection probabilities, do simulated carbon deposition profiles agree with experimental observations for side surfaces of the gaps. These species can be attributed to different particle sources, e.g. carbon atoms and hydrocarbon radicals. Background carbon ions and atoms have low and moderate values of the reflection coefficient (R ≤ 0.6), while some of the hydrocarbon radicals produced by chemical erosion of redeposited carbon layers have high reflection probability (R ≥ 0.9). Deposition at the bottom of the gaps cannot be adequately reproduced unless extreme assumptions on particle sources and reflection properties are imposed. Elastic neutral collisions and ionization of neutrals escaping the gaps have no significant influence on the results. Nevertheless, particle-in-cell simulations of plasma penetration into gaps are essential for estimating the incoming ion flux and leading to a better quantitative agreement with experimental observations.

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.

Modelling of 13CH4 injection experiments with graphite and tungsten test limiters in TEXTOR using the coupled code ERO-SDTrimSP

The 3D Monte-Carlo code ERO, which calculates erosion processes, impurity transport and deposition, has been coupled to the Monte-Carlo code SDTrimSP to simulate material mixing processes in wall components more precisely. SDTrimSP calculates the transport of ions in solids by means of the binary collision approximation. It keeps track of the depth dependent material concentration caused by implantation of projectiles in the solid. Modelling with the coupled code ERO-SDTrimSP is compared with dedicated TEXTOR experiments, in which the formation of mixed surface layers has been studied. In these experiments, methane 13CH4 was injected through graphite and tungsten spherical limiters during plasma exposure and the local redeposition probability was measured post mortem by surface analysis. A significant difference in the carbon 13C deposition efficiency, i.e. the ratio of the locally deposited to the injected amount of 13C, between graphite and tungsten was found, 4% for graphite and 0.3% for tungsten. Modelling of these experiments with ERO-SDTrimSP reproduces the clear substrate dependence with about 2% deposition efficiency on graphite and less than 0.5% on tungsten in good agreement with the experiment. The reason for the substrate dependence is partly explained by the higher physical sputtering yield of a thin carbon film on top of a tungsten substrate compared with a graphite substrate. Surface roughness of the materials has been identified to be another important parameter for the interpretation of the results.

Study of physical and chemical assisted physical sputtering of beryllium in the JET ITER-like wall

The effective sputtering yield of Be was determined in situ by emission spectroscopy of low ionizing Be as function of the deuteron impact energy (Ein = 25–175 eV) and Be surface temperature (Tsurf = 200 °C–520 °C) in limiter discharges carried out in the JET tokamak. Be self sputtering dominates the erosion at high impact energies (Ein > 150 eV) and causes far beyond 1. drops to low values, below 4.5%, at the accessible lowest impact energy (Ein 25 eV) achievable in limiter configuration. At medium impact energies, Ein = 75 eV, two contributors to the measured of 9% were identified: two third of the eroded Be originates from bare physical sputtering and one third from chemical assisted physical sputtering . The later mechanism has been clearly identified by the appearance of BeD A–X emission and quantified in cause of a temperature dependence at which the BeD practically vanishes at highest observed Be limiter temperatures. The recorded Tsurf dependence, obtained in a series of 34 identical discharges with ratch-up of Tsurf by plasma impact and inertial cooling after the discharge, revealed that the reduction of BeD is correlated with an increase of D2 emission. The release mechanism of deuterium in the Be interaction layer is exchanged under otherwise constant recycling flux conditions at the limiter.The reduction of with Tsurf is also correlated to the reduction of the Be content in the core plasma providing information on the total source strength and Be screening. The chemical assisted physical sputtering, always present at the nominal limiter pre-heating temperature of Tsurf = 200 °C, is associated with an additional sputtering channel with respect to ordinary physical sputtering which is surface temperature independent. These JET experiments in limiter configuration are used to benchmark the ERO code and verify ITER first wall erosion prediction. The ERO code overestimates the observed Be sputtering in JET by a factor of about 2.5 which can be transferred to ITER predictions and prolong the expected lifetime of first wall elements.

Determination of rate coefficients for fusion-relevant atoms and molecules by modelling and measurement in the boundary layer of TEXTOR

Rate coefficients for the excitation and ionization of neutral as well as singly ionized particles and-–predominantly-–their ratios S(D)/XB, which are important for the conversion of photon into particle fluxes in ionizing fusion boundary plasma, have been modelled and experimentally determined in boundary plasmas for fusion-relevant species such as He I, Li I, C I, B I&II, O I&II, Si I&II, Mo I, W I&II, H2, CH(D), C2.

Modelling of erosion and deposition at limiter surfaces and divertor target plates

Abstract Better understanding the processes of wall erosion and deposition remains an important issue also for future devices. The three-dimensional Monte-Carlo code ERO-TEXTOR has been developed in order to simulate plasma wall interaction and transport of eroded particles in the vicinity of wall components and to allow comparisons with experimental observations. This paper presents quantitative modelling of local transport and deposition of 13 CH 4 which was injected into the edge plasma of TEXTOR-94. In addition, chemical erosion at a graphite limiter in TEXTOR is simulated. Especially, the observed CD photon emission will be compared with measurements. Finally, first results of the modelling of erosion and deposition at an ITER-like divertor target plate are presented.

Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

Results from three-dimensional modeling of plasma edge transport and plasma–wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q = 10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q 95 = 4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP coil current yield a reduction of the width of the divertor flux spreading to about 20–25 cm and cause increased peak heat fluxes back to values similar to those in the axisymmetric case. The dependencies of these features on the divertor recycling regime and the perpendicular transport assumptions, as well as toroidal averaged effects mimicking rotation of the RMP field, are discussed in the paper.