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DOE OSTI2021

Bispectral analysis of broadband turbulence and geodesic acoustic modes in the T-10 tokamak

Local fluctuations of electrostatic potential, poloidal electric field, magnetic potential and electron density are simultaneously measured in the T-10 tokamak by a heavy ion beam probe (HIBP) having a five-slit energy analyser, which allows an estimate of the turbulent particle flux and $\boldsymbol {E}\times \boldsymbol {B}$ rotation velocity in the off-minor-axis gradient zone of the toroidal plasma column. The high spatial and temporal resolution of the modern multichannel HIBP makes it an effective tool to study plasma oscillations. Motivated by previous work that has documented time-resolved interactions between measured plasma parameters using correlation analysis (coherence of $E_{\textrm {pol}}$ and density $n_e$ , and cross-phase), a new result from bicorrelation analysis (bicoherence of magnetic potential $A_\zeta$ and density $n_e$ , and biphase) is reported for documenting the evidence of wave–wave coupling and energy transfer associated with the interaction between geodesic acoustic modes (GAM) and broadband, quasi-coherent modes.

Physics↗

DOE OSTI2024

Erosion resistance test of SiC mirror sample for ITER divertor VUV spectrometer

A series of tests have been performed to validate the resistance of the Silicon Carbide (SiC) mirror as the first mirror material of ITER VUV spectrometers to all ITER environmental conditions. Here we focused on the erosion (and deposition) of the SiC mirror sample caused by high-energy neutral particles. The flux of neutral particles reaching the first mirror was calculated using the ZEMAX software with a simplified entrance duct model. In the calculation, the particle flux reaching the first wall is necessary and the previously reported values derived from the SOLPS calculations were used. Based on this estimated particle flux, the erosion resistance tests were performed to check the erosion effect due to the high-energy neutral particles on the first mirror of ITER divertor VUV spectrometer. In the experiment, erosion was induced by exposing the SiC mirror sample to hydrogen and deuterium plasmas (and ions). The target fluence of incident ions in experiment is based on the estimation of the flux of neutral particles in ITER. The surface shape, composition, erosion depth, and surface roughness were measured to check the damage of the mirror surface after erosion test. Based on simulations and erosion resistance tests, it was concluded that the SiC mirror can be used as the first mirror of ITER divertor VUV spectrometer.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

On the role of ion temperature gradient turbulence in driving ion thermal transport in neutral beam injection-heated L-mode plasmas in a superconducting tokamak

In this paper, we report a detailed experimental study of the role of ion-temperature-gradient driven turbulence in driving ion thermal transport in neutral beam injection (NBI)-heated L-mode plasmas in Experimental Advanced Superconducting Tokamak (EAST) (Wan et al 2000 Nucl. Fusion 40 1057) coupled with linear and nonlinear gyrokinetic simulations. Significant ion-scale turbulence k ⟂ < cm −1 (k ⟂ < 1.5, where k ⟂ where is the perpendicular wavenumber and ρ s is the ion gyroradius calculated using local electron temperature T e ), measured with a microwave reflectometer, is observed in the plasma core at for r/a ≈ 0.265 and 0.5 (where r is half the diameter of the closed flux surface at a given radial position, and a is half the diameter of the last closed flux surface). Local linear stability analysis with experimental equilibrium quantities at these two radial locations using the GS2 gyrokinetic code shows that the most unstable ion-scale micro-instability is the ion temperature gradient (ITG) mode. Since the computed maximum ion-scale linear growth rates are larger than the local Waltz-Miller E × B shearing rate (Waltz and Miller 1999 Phys. Plasmas 6 4265), the E × B shear is unable to suppress ion-scale turbulence, consistent with the experimental observation of ion-scale turbulence. Experimental ion and electron thermal transports, calculated with power balance analysis, are both anomalous with ion thermal transport being the dominant channel. Nonlinear gyrokinetic simulations using the electrostatic global particle-in-cell gyrokinetic tokamak simulation (GTS) code (Weixing Wang et al 2010 Phys. Plasmas 17 072511) show good quantitative agreement between predicted ion thermal transport and that from the power balance analysis. These results demonstrate for the first time the important role of ITG turbulence in driving thermal transport in NBI-heated L-mode plasmas in EAST, showing the effectiveness of GTS code in simulating electrostatic turbulence in EAST and the need of suppressing ITG turbulence to improve energy confinement in EAST.

Sun, P. J. [Chinese Academy of Sciences (CAS), Hef↗

DOE OSTI2023

Experimental observation of a field-aligned ion beam produced by magnetic reconnection of two flux ropes

An ion beam field-aligned to the background guide field ($B$ 0 = $330$ G) was observed in a reconnection experiment on the Large Plasma Device and, to the authors' knowledge, is the first experimental observation of its kind. Two kink-unstable flux ropes ( L = 11 m, d = 7.6 cm) were made to collide, which allows magnetic reconnection to occur. Sub-Alfvénic ion beams with energies of up to 15 eV were then observed from measurements of the local ion energy distribution function. The beam ions do not appear to be heated. Further, they were correlated with the collision of the ropes and appear to be energized by magnetic reconnection. The results and interpretation of the measurements are supported by three-dimensional gyrokinetic particle simulations of the merging flux ropes and electric field measurements from previous experiments [W. Gekelman et al. , Astrophys. J. 853, 33 (2018)]. The mechanism behind the acceleration appears to be non-local.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MWm -2 . The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (≥1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (≥0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Evidence of restricted heat transport in National Ignition Facility Hohlraums

We present experimental evidence of restricted electron thermal conduction in the high-Z coronal plasma regions of laser-driven Hohlraums on the National Ignition Facility. Four separate measurements, three of which are direct observations of Hohlraum dynamics, corroborate this finding. (1) The velocity of the coronal plasma ablated and heated by the outer-cone laser beams is determined by time-dependent imaging of the gold plasma plume, or “bubble.” The velocities of the incoming plume (perpendicular to the Hohlraum axis) are consistent with high-fidelity 2D radiation-hydrodynamic simulations using flux-limited thermal electron conduction with a flux multiplier f = 0.03. Simulations using f = 0.15, which is very nearly classical Spitzer–Härm transport, predict plume velocities slower than measured. (2) Specific features in time-resolved images of the Hohlraum wall at an angle of 19° are also more consistent with f = 0.03 simulations compared to f = 0.15. (3) Spectroscopic tracers were added to the Hohlraum wall in the outer-beam bubble region. The ratios of hydrogen-like to helium-like line emission are sensitive to the electron temperature of the bubble. The hydrogen-like to helium-like ratios extracted from the time-integrated spectra of manganese and cobalt tracers from two observation angles are consistent with f = 0.03 and not with f = 0.15. (4) The time of peak capsule emission, or “bang time,” an integrated measurement, is also more consistent with f = 0.03 than with f = 0.15. Overall, while these findings do not identify the causes of restricted thermal conduction in Hohlraums, they motivate future experiments to test specific hypotheses and focus on model development in the regions of the plasma exhibiting restricted transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Implementation of Pfirsch–Schlüter parallel flow in x-ray imaging crystal spectrometer inversion analysis

The x-ray imaging crystal spectrometer (XICS) tomographic inversion code for Wendelstein 7-X (W7-X) has been modified to consider the effects of parallel flows and has been applied to analyze measurements taken during recent experimental campaigns. Previous analysis neglected the effects of parallel flows due to the primarily perpendicular geometry of the sightlines and the small magnitude predicted by neoclassical theory. To reconsider these effects, the incompressibility condition for plasma flows is used to calculate the parallel Pfirsch–Schlüter flow component for the equilibrium configuration. By incorporating this condition along with the geometry of the sightlines—i.e. the fractional contributions of perpendicular and parallel flows—, an updated expression for the measured flow is used for the profile inversion. Application of this modified inversion code to data from W7-X shows that the magnitude of the radial electric field and the flux surface averaged perpendicular flow are reduced by approximately a factor of 2 and brought into better agreement with neoclassical predictions and charge exchange recombination spectroscopy measurements.

Pfirsch–Schlüter flows↗

DOE OSTI2026

The effectiveness of D 2 pellet and gas injection in reducing intra-ELM tungsten erosion and heat flux in the DIII-D small angle slot divertor

Edge localized modes (ELMs) in H-mode plasmas can melt and erode plasma-facing components (PFCs) and lead to impurities in the core, reducing confinement. This study analyzes the use of D 2 pellet and gas injection for ELM mitigation on the DIII-D tokamak during the 2022 Small Angle Slot V-shaped Tungsten (W) (SAS-VW) divertor campaign, reducing W erosion and heat flux during ELMs. D α (656 nm) and WI (400.9 nm) filterscopes and Langmuir probes provide photon emission and electron density/temperature to estimate W atom erosion using the S/XB method, while surface eroding thermocouples measured ELM peak heat flux at the outer strike point (OSP). Thomson Scattering measurements of pedestal T e and n e provided input to predict W divertor erosion and heat flux during ELMs via the Free-Streaming plus Recycling Model (FSRM). While a greater D 2 mass injection rate decreased the ELM peak heat flux, the impact on W erosion was not monotonic. The average ‘large’ intra-ELM W erosion was lower for plasma shots with D 2 mass injection compared to the plasma shots without any D 2 mass injection in the SAS-VW divertor. However, plasma shots did not see significant changes in ‘large’ intra-ELM W divertor erosion as the D 2 mass injection rate increased. On average, the FSRM overestimated SAS-VW experimental intra-ELM W erosion by a factor of 6.9, but the overestimation is reduced to a factor of 3.6 with the implication of C deposition effects. Generally, experimentally measured and predicted quantities were worse for plasma shots with a lower D 2 mass injection rate and corresponding higher plasma carbon (C) impurity percentage (f C ). The discrepancies are postulated to be due to C/W material mixing, for which a simple analytic mixed-material model is presented. These results highlight the importance of incorporating and improving the robustness of a mixed-material layer model in the analysis of PFC erosion on present and future tokamak devices.

Carbon (C)↗

DOE OSTI2025

Radio Frequency Transients Correlated with Electron Flux Measured On-Board the STP-Sat6

We analyzed transient electromagnetic signatures, referred to as spacecraft environment discharges (SEDs) or on-board discharges (OBDs), and the in-situ particle environment simultaneously measured by the Radio Frequency Sensor (RFS) and the Space and Atmospheric Burst Reporting System Z-plasma spectrometer (SABRS-ZPS), respectively. These sensors are on-board the U.S. Department of Defense Space Test Program Satellite 6 (STP-Sat6) in geostationary orbit. Radio frequency signatures of SEDs on STP-Sat6 have relatively high peak amplitudes; they often caused amplitude-saturation in the so-called high band (120–140 MHz frequency range) of the RFS likely due to the proximity of their source to the RFS’s dipole antennas. We examined the occurrence rate of SEDs reported by the RFS and its temporal relationship to electron flux in the 7.9–12.2 keV range. Peaks in SED count rates were time-correlated with peaks in electron flux. The electron-flux peaks appeared to lead the SED count rate peaks by 24–45 min for the 7.9–12.2 keV energies. This indicates that incidence of electrons at this relatively low-energy range likely resulted in ramping up of the charging of the spacecraft leading to periods of high occurrence rates of discharges in and around the spacecraft.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

Mechanistic aspects of plasma-enhanced catalytic methane decomposition by time-resolved operando diffuse reflectance infrared Fourier transform spectroscopy

To study mechanistic aspects of plasma-enhanced catalysis, methane is decomposed by a supported Ni catalyst assisted by an Ar/O 2 atmospheric pressure plasma jet (APPJ). Here, the time-resolved surface response of the Ni catalyst is investigated by operando diffuse reflectance infrared Fourier transform spectroscopy (DRIFTS) using various experimental settings. Catalyst temperatures of room temperature (RT) and 500 °C and nozzle-catalyst surface distances of 3, 5 and 8 mm were examined, and the amount of O 2 of the Ar/O 2 gas mixture flowing through the APPJ was either 0 or 0.5%. A synergistic effect of surface bonded C-O was observed during the exposure of the Ni catalyst to the APPJ for low oxygen operating conditions (pure Ar jet). Surface bonded C-O formed only when there was plasma present and the C-O signal was enhanced for higher catalyst temperature. When the supported Ni catalyst was subjected to the plasma-generated particle fluxes using highly oxidizing conditions, the presence of surface bonded C-O was suppressed. The plasma-catalytic CO and CO 2 production in the gas phase measured downstream mirrored the surface behavior of C-O bonds when the plasma source operating condition was changed from a low oxygen portion to a high oxygen portion at high catalyst temperature (500 °C). CH n (n =1, 2, ) species on the catalyst surface were also studied by DRIFTS, and CHn destruction was found to correlate with C-O formation. In particular, the time-resolved CH n response showed a possible conversion process of CH n to C-O when the Ni catalyst was exposed to the plasma source. This finding may indicate a plasma-mediated regeneration of the catalyst by plasma-catalyst surface interactions.

42 ENGINEERING↗

DOE OSTI2024

Higher-order nonequilibrium term: Effective power density quantifying evolution towards or away from local thermodynamic equilibrium

A common approach to assess the nature of energy conversion in a classical fluid or plasma is to compare power densities of the various possible energy conversion mechanisms. A leading research area is quantifying energy conversion for systems that are not in local thermodynamic equilibrium (LTE), as is common in a number of fluid and plasma systems. Here we introduce the “higher-order nonequilibrium term” (HORNET) effective power density, which quantifies the rate of change of departure of a phase space density from LTE. It has dimensions of power density, which allows for quantitative comparisons with standard power densities. We employ particle-in-cell simulations to calculate HORNET during two processes, magnetic reconnection and decaying kinetic turbulence in collisionless magnetized plasmas, that inherently produce non-LTE effects. We investigate the spatial variation of HORNET and the time evolution of its spatial average. By comparing HORNET with power densities describing changes to the internal energy (pressure dilatation, Pi − D , and divergence of the vector heat flux density), we find that HORNET can be a significant fraction of these other measures (8% and 35% for electrons and ions, respectively, for reconnection; up to 67% for both electrons and ions for turbulence), meaning evolution of the system towards or away from LTE can be dynamically important. Applications to numerous plasma phenomena are discussed. Published by the American Physical Society 2024

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Measurements and modeling of type-I and type-II ELMs heat flux to the DIII-D divertor

Type-I and type-II edge-localized-modes (ELMs) heat flux profiles measured at the DIII-D divertor feature a peak in the vicinity of the strike-point and a plateau in the scrape-off-layer (SOL), which extends to the first wall. The plateau is present in attached and detached divertors and it is found to originate with plasma bursts upstream in the SOL. The integrated ELM heat flux is distributed at ~65% in the peak and ~35% in this plateau. The parallel loss model, currently used at ITER to predict power loads to the walls, is benchmarked using these results in the primary and secondary divertors with unprecedented constraints using experimental input data for ELM size, radial velocity, energy, electron temperature and density, heat flux footprints and number of filaments. The model can reproduce the experimental near-SOL peak within ~20%, but cannot match the SOL plateau. Employing a two-component approach for the ELM radial velocity, as guided by intermittent data, the full radial heat flux profile can be well matched. The ELM-averaged radial velocity at the separatrix, which explains profile widening, increases from ~0.2 km s –1 in attached to ~0.8 km s –1 in detached scenarios, as the ELM filaments' path becomes electrically disconnected from the sheath at the target. The results presented here indicate filaments fragmentation as a possible mechanism for ELM transport to the far-SOL and provide evidence on the beneficial role of detachment to mitigate ELM flux in the divertor far-SOL. However, these findings imply that wall regions far from the strike points in future machines should be designed to withstand significant heat flux, even for small-ELM regimes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Understanding the L-H isotope effect at the DIII-D tokamak and advancements in synthetic turbulence diagnostics

Abstract It is determined that while heat flux differences between hydrogen and deuterium isotope experiments result from natural differences in carbon impurity content at DIII-D, it is not the origin of the low to high confinement mode (L-H) transition isotope effect. More specifically, a two times larger edge radial electric field in hydrogen compared to deuterium is uncovered and believed to play an important role. The origin of this radial electric field difference is determined to have two possible origins: differences in poloidal rotation and turbulent Reynolds stress in the closed field line region, and increased outer strike point temperatures and space potentials on open field lines. Experimental observations from both profile and turbulence diagnostics are supported by nonlinear gyrokinetic simulations using the code CGYRO. Simulations illustrated heat transport isotope effects in the plasma edge and shear layer resulting from differences in impurity content, electron non-adiabaticity, and main ion mass dependent E × B shear stabilization. Turbulence prediction comparisons from flux-matched CGYRO simulations to experimental measurements including electron temperature, density and velocity fluctuations are found to be in good agreement with available data. A dedicated DIII-D experiment in hydrogen was performed to seed more carbon than naturally occurring, to match deuterium experiments, and possibly reduce the L-H power threshold based on gyro-kinetic predictions. To our surprise, while ion temperature gradient (ITG) turbulence was stabilized, nodiscernible change in L-H power threshold were observed in these special hydrogen experiments. In particular, it is noticed that the edge radial electric field and Reynolds stress were observed as nearly unchanging in the presence of ITG stabilization. These experimental data have enabled a more comprehensive picture of the multitude of isotope effects at play in fusion experiments, and the important potential connection between the confined and unconfined plasma regions in regulating L-H transition dynamics.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Relevance of E × B drifts for particle and heat transport in divertors

Radial electric fields up to ~4 kV m –1 are observed in the boundary between the private flux region (PFR) and the scrape-off layer (SOL) driving E × B drifts between the inner and outer targets at speeds up to 2.8 km s –1 in the Tokamak à configuration variable divertor. The resulting E × B fluxes, located in a narrow region ($\Delta {\rho _\Psi } < 0.012$ in normalized radius or $\Delta $ R – R sep <4 mm mapped to the outer midplane) are equivalent to around 20% of the total heat and particle flux to the divertor targets (inner + outer). At the peak E r , the E × B poloidal transport is equivalent to parallel flows with M ∥ ~ 3. In the snowflake divertor with a second X-point in the outer SOL, the drifts in the PFR-SOL boundary were equivalent to around 30% of the total heat and particle flux to the divertor targets and cover a region ~50% wider than in the single null ($\Delta {\rho _\Psi }$ ~ 0.018, $\Delta $R – R sep ~ 6 mm). Furthermore, the location of the PFR-SOL boundary drift shifts radially in the E ∥ × B direction when reversing the toroidal field direction. Peaks in density and electron pressure have been identified near the primary X-point along with large gradients in density, temperature, and potential, the latter resulting in a local electric field ~2.7 kV m –1 which drives a drift (1.9 km s –1 ) upwards towards the closed flux surfaces. Floating potential (V f ) magnitudes up to 75 V (~2 kT e ) were measured, indicating that V f and parallel currents should not be neglected when estimating plasma potential.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Effects of radial transport on divertor power and particle flux widths under different operational regimes in EAST

Here, a study of the effects of radial transport on the outer divertor particle and power flux widths (λ js and λ q ) is carried out using plasmas under different operational regimes in the Experimental Advanced Superconducting Tokamak (EAST). In the EAST experiments, the λ js values measured during the grassy intra- and inter-ELM phases are similar, and the averaged value of λ js during grassy ELMy discharges is found to be smaller than that of the intra-ELM phase of type-I ELMy discharges but larger than that of the inter-ELM phase of type-I ELMy discharges. Simulations of scans of the radial particle and heat transport coefficients, D and χ e , performed using the BOUT++ transport code show that both λq and λjs increase with D and χe, especially when D or χ e is larger than a threshold value, indicating that background turbulence starts to matter in the determination of λ q and λ js when D or χ e surpasses the threshold value. A comparison between the simulation of the D scan and the experimental results shows that the different values of λ js obtained under different plasma operational regimes are probably due to the different intensities of background turbulence, which is beyond the scope of Goldston's heuristic drift-based model as reported by Goldston et al (2012 Nucl. Fusion 52 013009), since their model was derived by assuming that turbulent transport does not exceed a certain level. The ratio of λ js to λ q is greatly affected by the radial transport; however, simulations carried out using Scrape-off Layer Plasma Simulation (SOLPS) show that divertor geometry and plasma density have big influences on λ js , and thus could also affect the ratio of λ js to λ q .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Overview of the first Wendelstein 7-X long pulse campaign with fully water-cooled plasma facing components

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Spatially resolved absolute densities of reactive species and positive ion flux in He-O 2 RF-driven atmospheric pressure plasma jet: touching and non-touching with dielectric substrate

Atmospheric pressure plasma jets (APPJs) are operating in direct contact with dielectric substrates for many applications. Nonetheless, many studies have been focusing on free jets. In this letter, we report spatially resolved measurements of reactive species including neutral and ionic species in an He + 1% O 2 RF-driven APPJ by molecular beam mass spectrometry equipped with a dielectric sampling plate, enabling to detect species densities for APPJ operating remotely or in direct contact with a substrate. Here, the spatially resolved distribution of O, O 2 (a 1 Δ g ) and O 3 is dominated by convection and, remarkably, shows minimum differences between touching and non-touching conditions.

74 ATOMIC AND MOLECULAR PHYSICS↗

DOE OSTI2024

Multi-physics modeling of tungsten collector probe samples during the WEST C4 He campaign

Abstract We describe the results of a multi-scale, multi-physics modeling assessment of SOLPS-ITER, hPIC2, RustBCA and Xolotl, in which five single-crystal tungsten (W) samples were placed in a reciprocating collector probe and exposed to helium (He) plasma in the WEST fusion device. In our models, we considered a pure (100 %) He plasma, as well as one with oxygen (O) present (95% He 5% O) corresponding to the impurity concentration estimated during the C4 He campaign in WEST. Our SOLPS simulations approximately match experimental reciprocating Langmuir probe plasma measurements of plasma density and temperature. Using these plasma parameters as input, hPIC2 and RustBCA predict that the presence of oxygen impurities lead to a 15%–20% decrease in ion and heat fluxes to the surface, and an order of magnitude higher sputtering yields (compared with a pure He plasma). Xolotl predictions for the response of tungsten to plasma surface interactions (PSIs) agree with experimental LAMS analysis, and indicate large near-surface He concentrations, which quickly decay with depth. Our model also shows an increasing role of erosion—in removing the near-surface He—with time. Overall, slightly higher retention is predicted for tungsten exposed to a pure He plasma, with the largest differences in the near-surface gas content caused by the large oxygen-induced erosion. This highlights the important role that impurities play in PSI. Therefore, future work will focus on providing a fully self-consistent description of oxygen (and oxides, etc.) in our models, through multi-species implementation in GITR and inclusion of oxygen and tungsten oxide formation in Xolotl.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗