Surface Of Plasma Facing Components Research Articles

Wall heating by Subcritical Energetic Electrons generated by the runaway electron avalanche source

Abstract Subcritical energetic electrons (SEEs) produced by the runaway electron (RE) avalanche source at energies below the runaway threshold are found to be the primary contributor to surface heating of plasma-facing components (PFCs) during final loss events. This finding is supported by theoretical analysis, computational modeling with the Kinetic Orbit Runaway electrons Code (KORC), and qualitative agreement with DIII-D experimental observations. The avalanche source generates significantly more secondary electrons below the runaway threshold, which thermalize rapidly when well-confined. However, during a final loss event, the RE beam impacts the first wall, and SEEs are deconfined before they can thermalize. Additionally, because the energy deposition length decreases faster than energy, the deposited energy density, and thus the maximum PFC surface temperature change, is larger for SEEs than REs. KORC simulations employ an analytic first wall to model particle deconfinement onto a non-axisymmetric wall composed of individual tiles. PFC surface heating is calculated using a 1D model extended to include an energy-dependent deposition length scale. Simulations of DIII-D qualitatively agree with infrared (IR) imaging only when SEEs from the avalanche source are included. These results demonstrate that SEEs are the dominant contributor to PFC surface heating and indicate that the avalanche source plays a critical role in the PFC damage caused during final loss events. The prominence of SEEs also has important implications for interpreting IR imaging, one of the primary diagnostics for RE-wall interaction diagnosis, despite REs dominating the energy and current density. This result improves predictions of wall damage due to post-disruption REs to estimate material lifetime and design RE mitigation systems for ITER and future reactors.

Assessing global beryllium erosion via tomographic reconstruction of 3D beryllium emission profiles in ITER

A new method for estimating the global erosion of beryllium (Be) in ITER is proposed. The method uses ray tracing-aided tomography to reconstruct the three-dimensional (3D) profile of beryllium visible-light emissivity in boundary plasma from images captured with filtered cameras of VIS/IR wide angle viewing system, H-alpha (and Visible) Spectroscopy diagnostics and signals collected with divertor impurity monitor. The light reflected into the detectors from metallic plasma-facing components (PFCs) is filtered out in the process. The reconstructed Be emissivity is then used to assess the Be influx density distribution along all Be PFCs by integrating the product of the emissivity and the S/XB coefficient along the normal to the PFC surface. The accuracy of this method is evaluated by a comparison with synthetic emissivity data produced by recent simulation of global Be erosion and migration in ITER using the ERO2.0 code. The impact of the uncertainty of PFC light reflection properties on the error in reconstructing the 3D Be emissivity profile and Be influx density is analyzed. The method allows to recover with good accuracy the Be influx density in plasma-wetted areas under the conditions of H-mode fusion power operation with high plasma density in far scrape-off layer (SOL). Under the conditions of lower far-SOL plasma density and L-mode operation, only the total Be influx integrated over the area of the first wall panels with relatively high Be erosion can be reconstructed with a high accuracy. It is shown that neglecting the effects of light reflection may lead to a twofold overestimation of the total Be influx.

Development of fibered LIBS device for tokamak Plasma Facing Components characterization

The LIBS technique enables to measure the deposits composition in fusion machines, particularly the tritium concentration on tokamak Plasma Facing Components (PFC), thus allowing the estimation of the deposited tritium quantity for safety control and the composition of the PFC surface in light impurities (oxides) which can pollute the plasma when evaporated. This paper presents the development of a fibered LIBS device, mounted on an articulated arm and able to operate in nanosecond and picosecond laser regimes for PFC characterizations. The different components of the LIBS device are presented as well as the first LIBS spectra on tungsten targets.

An approach for measurement of the magnetic field near the surface of plasma-facing components in tokamak by laser induced breakdown spectroscopy based on Zeeman effect

Laser-induced breakdown spectroscopy (LIBS) has been demonstrated to diagnose the surface element distribution of plasma-facing components (PFCS) in fusion devices. However, the magnetic field intensity and polarity near the wall surface change frequently and cannot be measured in real time, which will affect the analytical accuracy of LIBS results. The determination of the strength and polarity direction of the magnetic field near the surface will be conducive to the quantitative analysis of LIBS and provide valuable information for the distribution of the magnetic field near the surface of PFCs. In this work, the effects of magnetic fields on the characteristics of laser-induced Mo plasma, such as the ablation depth and mass, the spatio-temporal evolution behavior of spectrum and the plasma parameters, were first systematically investigated. The different phenomenon of continuum radiation, Mo I line and Mo II line under the magnetic field were mutually verified by optical emission spectroscopy and fast plasma imaging methods. Meanwhile, an approach for measurement of the magnetic field intensity and polarity near the surface of PFCS by LIBS method has been proposed and investigated. When the magnetic field polarity is changed, the spatial distribution profile of Mo spectrum will drift in the opposite direction affected by of the Lorentz force, which realizes the measurement of magnetic field polarity. Moreover, the splitting distance of Mo I line caused by Zeeman effect with the variable magnetic fields were preliminarily calculated. The experimental values of splitting distance agree well with the theoretical values that calculated by the known magnetic field strength. These results confirm the feasibility of measuring the magnetic field intensity and polarity of PFCs surface by LIBS approach.

Helium bubble size effects on the surface morphological response of plasma-facing tungsten

Abstract We report a simulation study on the effects of helium (He) bubble size on the morphological evolution and pattern formation on the surface of tungsten used as a plasma-facing component (PFC) in nuclear fusion devices. We have carried out a systematic investigation based on self-consistent dynamical simulations of surface morphological evolution according to an atomistically-informed, 3D continuum-scale model that captures well the relevant length and time scales of surface nanostructure formation in PFC tungsten. The model accounts for PFC surface diffusion, driven by the biaxial compressive stress originating from the over-pressurized He bubbles in the near-surface region of PFC tungsten as a result of He plasma exposure, combined with the formation of self-interstitial atoms in tungsten that diffuse toward the PFC surface and the flux of surface adatoms generated as a result of surface vacancy-adatom pair formation upon He implantation; this transport of surface adatoms contributes to the anisotropic growth of surface nanostructural features due to the different rates of adatom diffusion along and across step edges of islands on the tungsten surface. Our detailed analysis reveals that varying the average He bubble size plays an important role in the PFC surface growth kinetics as well as the resulting surface topography. Specifically, we find that an increase in the He bubble size leads to a deceleration in the growth rate of the tungsten nanotendrils that emanate from the PFC surface. We also find that the separation distance between the resulting surface features increases with increasing He bubble size, as well as over time. This coarsening effect is a thermally activated process resulting in an accurate description of the temperature dependence of the average surface feature separation by an Arrhenius relation.

Effects of surface vacancy-adatom pair formation on PFC tungsten surface morphological response

We report a simulation study of the effect of He-irradiation-induced surface vacancy-adatom pair formation on the surface morphological evolution of plasma-facing component (PFC) tungsten and examine a number of factors that impact such evolution. Our analysis is based on self-consistent dynamical simulations according to an atomistically-informed, continuum-scale surface evolution model that has been developed following a hierarchical multiscale modeling strategy and can access the spatiotemporal scales of relevance to fuzz formation. The model accounts for the flux of surface adatoms generated as a result of the surface vacancy-adatom pair formation effect upon He implantation, which contributes to the anisotropic growth of surface nanostructural features due to the different rates of adatom diffusion along and across step edges of islands on the tungsten surface. We have carried out atomic-scale computations of optimal diffusion pathways along and across island step edges on the W(110) surface and calculated Ehrlich–Schwoebel (ES) barriers in adatom diffusion along and across such step edges. This aspect of surface adatom diffusion contributes to anisotropic surface atomic fluxes, terrace and step diffusive currents, and has been incorporated into our PFC surface evolution model, which predicts the formation of preferentially aligned nanoridge stripe patterns on the PFC surface. We establish that these anisotropic diffusive currents accelerate nanotendril growth on the PFC surface and the onset of surface nanostructure pattern formation. We also explore systematically the dependence of the PFC surface morphological response on the surface temperature and He ion incident flux, characterize in detail the resulting surface topographies and growth kinetics, and compare the predicted surface morphologies with experimental observations. Our simulation predictions for the emerging surface nanostructure patterns under certain plasma exposure conditions are consistent with experimental findings in the literature.

Plasma facing components with capillary porous system and liquid metal coolant flow

Liquid metal can create a renewable protective surface on plasma facing components (PFC), with an additional advantage of deuterium pumping and the prospect of tritium extraction if liquid lithium (LL) is used and maintained below 450 °C, the temperature above which LL vapor pressure begins to contaminate the plasma. LM can also be utilized as an efficient coolant, driven by the Lorentz force created with the help of the magnetic field in fusion devices. Capillary porous systems can serve as a conduit of LM and simultaneously provide stabilization of the LM flow, protecting against spills into the plasma. Recently, a combination of a fast-flowing LM cooling system with a porous plasma facing wall (CPSF) was investigated [A. Khodak and R. Maingi, Nucl. Mater. Energy 26, 100935 (2021)]. The system takes an advantage of a magnetohydrodynamics velocity profile as well as attractive LM properties to promote efficient heat transfer from the plasma to the LL at low pumping energy cost, relative to the incident heat flux on the PFC. In the case of a disruption leading to excessive heat flux from the plasma to the LM PFCs, LL evaporation can stabilize the PFC surface temperature, due to high evaporation heat and apparent vapor shielding. The proposed CPSF was optimized analytically for the conditions of a fusion nuclear science facility [Kessel et al., Fusion Sci. Technol. 75, 886 (2019)]: 10 T toroidal field and 10 MW/m2 peak incident heat flux. Computational fluid dynamics analysis confirmed that a CPSF system with 2.5 mm square channels can pump enough LL so that no additional coolant is needed.

Qualification & testing of joining development for DEMO limiter component

A development plan for validation of functional principles is defined to support the challenges of mock-up manufacturing and testing. It is aimed to develop the process and infrastructure for qualifying fusion components for the limiters in the European DEMO. The limiters are components that define the plasma boundary by direct contact during normal and off normal transient events and thus they protect the First Wall of the Breeding Blanket System from extreme heat fluxes during these events. Within this framework, the joints play an important role for making feasible the combination of dissimilar materials required for the plasma facing components (PFCs) by providing a compliant interlayer. The main limiter PFC functionality is to act as thermal barrier, therefore the materials combination shall be able to support the thermal gradient between the high heat flux (HHF) from the plasma and the heat sink. The primary aim of the PFC is to ensure the structural integrity of the heat sink of the PFC (to prevent in-vessel loss of coolant accident), as it is foreseen that during the off-normal transients the PFC surface may melt or evaporate. Minimizing the large deformation and guaranteeing the strength and fatigue behaviour of the joints is required to achieve this.Therefore, a testing programme for joining development and qualification based on brazing technology is performed. It is focused on joint assessment between representative filler metals (OB1025TM, OB950TM, PB950TM, NBLMTM & H-BronzeTM) from the different families in the market and the chemical compatibility, capillarity flow and spreadability on the typical base materials used for PFC (laser powder bed fusion additive manufactured W-6%Ta, W, P91, OFHC Copper, CuCrZr). The main results show good wetting of the gold-copper alloy (OB1025TM) with all the base materials. It allows to progress with the integration of a PFC to create the process and infrastructure for optimizing the design of critical joints. NBLMTM seems to be an interesting filler for materials with high melting temperature as tungsten and P91 and OB950TM presents acceptable wetting condition with the base materials. PB950TM is rejected because the excess of wetting on all the base materials.

Onset of fuzz formation in plasma-facing tungsten as a surface morphological instability

Based on simulation results and theoretical analysis according to an atomistically informed continuum-scale model for the surface morphological response of tungsten plasma-facing component (PFC) in nuclear fusion devices, we demonstrate that the onset of the well-known surface nanostructure (known as ``fuzz'') formation in PFC tungsten is the outcome of a stress-induced surface morphological instability. Specifically, we show that formation and growth of nanotendrils emanating from the exposed surface of PFC tungsten, a precursor to fuzz formation, is caused by a long-wavelength surface morphological instability triggered by compressive stress in the PFC near-surface layer due to over-pressurized helium (He) bubbles forming in this near-surface region through implantation of low-energy He ions from the plasma. Using linear stability theory (LST), we predict the onset of surface growth in response to low-amplitude perturbations from a planar PFC surface morphology and calculate the average spacing between nanotendrils growing from the PFC surface, in excellent agreement with self-consistent dynamical simulations of surface morphological response according to the fully nonlinear surface evolution model starting with random fluctuations from the planar surface morphology that result in nanoscale surface roughness. In addition, we examine the morphological response of the PFC surface to low-amplitude perturbations of very long wavelengths from its planar morphology and interpret fully the simulation results on the basis of a weakly nonlinear tip-splitting instability theory, which predicts a post-instability nanotendril pattern formation with nanotendril separation consistent with the LST predictions regardless of the initial surface perturbation. Finally, we compare our simulation predictions for surface nanotendril growth with experimental measurements of fuzz layer growth on the exposed surface of PFC tungsten. Our simulation results are in very good agreement with the surface growth measurements at the early stages of fuzz growth, further establishing the onset of fuzz formation in PFC tungsten as the outcome of a stress-driven PFC surface morphological instability.

Hole formation effect on surface morphological response of plasma-facing tungsten

We report simulation results on the effect of helium (He) bubble bursting-mediated surface hole formation on the surface morphological response of tungsten plasma-facing components (PFCs) in nuclear fusion devices. Our analysis is based on an atomistically informed, continuum-scale model, which is capable of accessing the spatiotemporal scales relevant to the fuzz nanostructure formation process on the surface of PFC tungsten. Our simulations account, in an empirical fashion, for two types of subsurface bubble dynamical phenomena in the nanobubble region of PFC tungsten during He plasma irradiation, involving bubble bursting and surface crater formation. We demonstrate that the hole formation effect on the PFC tungsten surface accelerates the growth rate of nanotendrils and the onset of fuzz formation. As a result, the predicted incubation time for surface nanotendril growth is shortened, in agreement with experimental data of incubation fluence at comparable plasma exposure conditions. We also explore systematically the dependence of the PFC surface morphological response on the areal density of holes introduced at regular time intervals onto the He-implanted tungsten surface, a parameter in our analysis that serves as a proxy for the rate of He bubble bursting. More importantly, our simulations capture fine surface features in the PFC tungsten surface morphology and predict that the average spacing between nanotendrils is on the order of 100 nm, consistent with the experimental findings.

Effects of elastic softening and helium accumulation kinetics on surface morphological evolution of plasma-facing tungsten

Based on a continuous-domain model, capable of accessing the spatiotemporal scales relevant to fuzz formation on the surface of plasma-facing component (PFC) tungsten, we report self-consistent simulation results that elucidate the effects of elastic softening and helium (He) accumulation kinetics on the surface morphological response of PFC tungsten. The model accounts for the softening of the elastic moduli in the near-surface region of PFC tungsten, including both thermal softening at high temperature and softening due to He accumulation upon He implantation. The dependence of the elastic moduli on the He content follows an exponential scaling relation predicted by molecular-dynamics simulations, while the He content in the near-surface region of PFC tungsten evolves according to a first-order saturation kinetics, consistent with experimental and simulation results reported in the literature. We establish that this elastic softening accelerates both nanotendril growth on the PFC surface and the onset of fuzz formation. We also explore the role of the rate of He accumulation to a saturation level in the near-surface region of irradiated tungsten in the onset of fuzz formation. For PFC tungsten surfaces such as W(110) where, under typical irradiation conditions, the characteristic time scale for stress-driven surface diffusion is comparable to the characteristic time scale for He accumulation, we find that accelerating the rate of He accumulation accelerates the growth rate of nanotendrils emanating from the surface. Additionally, we present a systematic parametric study of the PFC surface morphological response to explore its dependence on the He accumulation kinetics that is controlled by the irradiation conditions for low-energy implantation. Finally, we introduce an incubation time for nanotendril growth on the PFC surface, a concept equivalent to that of incubation fluence discussed in the literature, to predict and explain the minimum exposure time required to observe fuzz formation on PFC tungsten surfaces.

Pulsed laser removal of tungsten nanoparticle aggregates: Surface analysis and visualization of particle ejection dynamics

In this study, we report on some of the aspects that describe the interaction of nanosecond UV laser pulses with solid surfaces covered by layers of tungsten nanoparticles aggregates, with an interest in applications for next-generation fusion reactors. The safe and efficient operation of such reactors is challenged by their unavoidable contamination with dust by-products due to interactions between plasma and the surface of plasma-facing components. We investigate the cleaning possibility of tungsten nanoparticle layers from solid surfaces using a KrF excimer laser source with tailored spot shape. The samples that emulate dust contaminated surfaces are synthesized by magnetron sputtering combined with gas aggregation technique. In order to shed light into the particle removal efficiency, the tungsten samples are investigated by scanning electron microscopy, Raman spectroscopy, and profilometry. Furthermore, in order to gain insight into the W nanoparticles and fragments mobilization dynamics, a plasma imaging setup is used to visualize the speed and directionality of the ejected material. We find that a very good surface cleaning can be achieved, even for the lowest considered laser fluence (0.5 J/cm2) following 8 pulses. Material ejection from the laser irradiated surfaces occurs in two regimes, over very diverse time scales: (i) first, a fast plasma plume travelling at 40,000 km/h reaches the substrate in a few microseconds, (ii) followed by the ejection of large particle conglomerates at speeds of the order of few hundred km/h. The particle ejection dynamics is shown to differ depending on the type of atmosphere (ambient pressure and vacuum, respectively), whereas the plasma plume evolution could only be evidenced under vacuum conditions.

Active Radiative Liquid Lithium Divertor for Handling Transient High Heat Flux Events

The extreme heat flux anticipated in fusion reactor divertor plasma facing components (PFCs) is perhaps the most challenging technology issue for fusion energy development. Most divertor PFCs are designed based on the maximum steady-state operational limits. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the passive and active radiative liquid lithium divertor concepts (RLLD and ARLLD) (Ono et al. in Nucl Fusion 53:113030, 2013; Fusion Eng Des 89:2838, 2014). This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal with relatively small amount of lithium ~ few moles/s to handle the expected steady-state high divertor heat load. However, in addition to the high steady-state heat flux, the fusion reactor divertor PFCs could also experience significant transient heat flux such as ELMs and/or other magnetic reconnection events which can deposit large transient heat flux onto the divertor PFCs. If unprotected, it could damage the divertor PFC surfaces which could lead to a highly undesirable unplanned shutdown for PFC repair and/or replacement. In this paper, we explore feasibility of LLD and ARLLD concepts in handling such transient heat flux to protect the divertor PFCs from the extreme transient heat flux while maintaining the normal plasma operations. We also suggest a possible implementation technique using inductive pellet injector for the reactor PFC protection from transient heat flux which can be tested on NSTX-U.

Plasma Facing Component Characterization and Correlation With Plasma Conditions in Lithium Tokamak Experiment-β

Lithium coatings in the Lithium Tokamak eXperiment (LTX) led to flat temperature profiles. The flat temperature profiles were observed along with a hot, low density edge, implying a broad, collisionless scrape-off layer (SOL). Additionally, in vacuo X-ray photoelectron spectroscopy (XPS) measurements established that lithium coatings evaporatively deposited onto high-Z plasma facing components (PFCs) became oxidized while retaining the ability to achieve good plasma performance long after lithium was applied to the PFCs. Longstanding theory predicted flat temperature profiles with low recycling walls, which was presumed to be due to hydrogen binding with elemental lithium to form lithium hydride. The presence of oxidized lithium, however, raised questions regarding the exact mechanism of hydrogen retention in LTX. To investigate these questions, the upgraded facility LTX- $\beta $ includes a new sample exposure probe (SEP) for more detailed in vacuo analysis of PFC samples. The SEP is equipped with a vacuum suitcase capable of transporting samples representative of the LTX- $\beta $ outer midplane PFCs to a stand-alone XPS system while maintaining pressures lower than the LTX- $\beta $ base vacuum to limit the contamination between sample exposure and analysis. The low-energy resolution XPS system used in past experiments could only enable the determination of elemental percentages on the PFC sample surfaces. Because the new XPS system has higher energy resolution, it is more direct to assign chemical compounds to the measured binding energies. This capability has been confirmed by comparing XPS data from PFC test samples with measurements using a commercial high-resolution XPS system. Quartz crystal microbalances (QCMs) were used to quantify the thickness of the deposited lithium on the LTX- $\beta $ PFCs. This article describes the application of the SEP to characterize the PFC surfaces using XPS and their relationship to plasma conditions.

Plasma-facing components damage and its effects on plasma performance in EAST tokamak

Abstract During the tokamak operations, plasma-wall interactions (PWI) can severely damage the plasma-facing components (PFCs), which will lead to degradation of plasma performance due to radiation losses by the impurities released from the damaged PFCs surfaces. The PFCs damage and its effects on the plasma performance are studied experimentally for the EAST tokamak. Substantial dusts were generated from the hot spots at the guard limiter of LHCD antenna and the tungsten (W) melting processes from the W upper divertor, which results in disruption due to high radiation losses. Two methods can be effectively used to recover the plasma discharges after tungsten melting induced disruptions, reducing the auxiliary heating power or moving away the position of strike point from the melting site. Statistics analysis indicate that about 12.0 % and 38.3 % of disruptions were due to severe impurity burst from the strong PWI processes in 2017 and 2018 campaign, respectively. In 2018 campaign, the plasma disruption due to tungsten impurity burst is about 26.3 %, which is much higher than 4.8 % in 2017 campaign. This is due to the higher auxiliary heating power, larger misalignment on W PFCs and the new W guard limiter. The main features on the effects of PFC damage on plasma operation are shown.

Effect of annealing on microstructure, texture and hardness of ITER-specification tungsten analyzed by EBSD, vickers micro-hardness and nano-indentation techniques

Tungsten and its alloys have been extensively studied to be used as materials for plasma facing components (PFCs) in future nuclear fusion reactors. Under fusion operating conditions, the surface of PFCs will experience high heat flux exposure which will affect some microstructural features of the material. In the present work, the effect of annealing temperature on microstructure, texture and mechanical behavior of ITER-specification commercial pure tungsten was investigated by combining Electron Backscatter Diffraction (EBSD), Vickers micro-hardness and nano-indentation techniques. The samples were analyzed in the as-received and in the annealed conditions. Annealing was done for 1 h at three different temperatures: 1300 °C, 1500 °C and 1800 °C. Three stages in microstructural evolution were observed: recovery at 1300 °C, recrystallization at 1500 °C and grain growth at 1800 °C. As the annealing temperature increases and recrystallization takes place, α-fiber texture components are strengthened. Simultaneously, a reduction in both Vickers and nano-indentation hardness is found.

Validating fast-ion wall-load IR analysis-methods against W7-X NBI empty-torus experiment

The first neutral beam injection (NBI) experiments in Wendelstein 7-X (W7-X) stellarator were conducted in the summer of 2018. The NBI system is used to heat the magnetically confined plasma by neutralising an accelerated hydrogen ion beam and directing it into the plasma, where the resulting energetic ions release their energy to heat the plasma. The modelling of the NBI fast ion experiments has commenced, including estimation of the shine-through and the orbit-losses. The stellarator has a wide-angle infra red (IR) imaging system to monitor the machine plasma facing component surface temperatures. This work validates the NBI model “Beamlet Based NBI (BBNBI)” and the newly written synthetic IR camera model. The validation is accomplished by comparing the measured and the synthetic IR camera measurements of an experiment where the NBI was injected into the vacuum vessel without a plasma. A good qualitative and quantitative match was found. This agreement is further supported by spectroscopic and calibration measurements of the NBI and and IR camera systems.

SMITER: A field-line tracing environment for ITER

Built around the SMARDDA modules for magnetic field-line tracing [IEEE Tr. Plasma Sc. 42 (2014) 1932], the SMITER code package (SMARDDA for ITER) is a new graphical user interface (GUI) framework for power deposition mapping on tokamak plasma-facing components (PFC) in the full 3-D CAD geometry of the machine, taking as input a user-defined specification for parallel heat flux in the scrape-off layer (SOL) and a description of the equilibrium magnetic flux. The software package provides CAD model import and integration with the ITER Integrated Modelling and Analysis Suite (IMAS), parametric CAD components catalogue and modelling, CAD de-featuring for PFC surface extraction, meshing, visualization (using an integrated ParaView module), Python scripting and batch processing, storage in hierarchical data files, with several simulation cases in one study running in parallel and using message passing interface (MPI) for code speed-up. An integrated ParaView module can combine CAD geometry, magnetic field equilibrium, meshes and results for detailed setup analysis and a module is under development for full finite element computation of surface temperatures resulting from the power deposition patterns on 3-D PFCs. The code package has been developed for ITER, but can be deployed for similar modelling of any tokamak. This paper presents and discusses key features of this field-line tracing environment, demonstrates benchmarking against existing field-line tracing code and provides specific examples of power deposition mapping in ITER for different plasma configurations.

WEST regular in-vessel Inspections with the Articulated Inspection Arm robot

Currently on fusion devices, diagnostics are mainly aiming at plasma analysis and control. However, operational and programmatic needs have appeared for regular in-vessel components monitoring during plasma campaign. Light robotics systems could meet this requirement and may be a way as well to replace human interventions to fix damaged in vessel components. To minimize the impact on machine operation, the robotic system has to be mini-invasive and compatible with operating conditions (vacuum, temperature…).To fulfill this goal, CEA has developed a multipurpose carrier able to be operated inside WEST vessel between plasma pulses. A prototype of this robot, called Articulated Inspection Arm (AIA), was tested in 2008 in Tore Supra vacuum vessel. A major upgrade was performed in 2014–2015 with the aim of converting this prototype into a reliable tool in support to WEST operation. During the WEST components manufacturing and installation (2014–2016), the robot was integrated and tested in the EAST Tokamak.Since 2017, the AIA has been regularly used during the WEST plasma campaigns. Movies provided by the embedded camera allow to assess the evolution of Plasma Facing Components surface state and the effects of plasma loads, runaways and disruptions.The robot operation was also very helpful to assess the needs for maintenance, to assist mechanical assembly without man entry and to perform diagnostics calibration under relevant conditions.

Effect of boronization on plasma-facing graphite surfaces and its correlation with the plasma behavior in NSTX-U

Boronization is a Plasma Facing Component (PFC) conditioning technique widely used in tokamak machines. The National Spherical Torus Experiment-Upgrade (NSTX-U) applied this conditioning, using a plasma glow with a deuterated Trimethyl-boron (d-TMB) and He mixture. The use of boronization during the campaign improved the plasma performance, allowing longer plasma discharges and H-mode access. The chemical state of an ATJ graphite sample, used as a proxy for the NSTX-U PFCs, was monitored in-situ using the Materials Analysis Particle Probe (MAPP) diagnostic and X-ray Photoelectron Spectroscopy (XPS). The XPS data showed a progressive rise (from < 5% to 23%) in the oxygen concentration of the boronized ATJ sample as the D+ fluence increased. Filterscopes were used to measure the light emitted by oxygen impurities in the plasma near the surface of the PFC. An increase in the registered magnitude of the OII line, normalized to the Dγ intensity, was observed as the concentration of O on the ATJ surface increased. The plasma performance was found to be strongly correlated to oxygen impurity concentrations at the plasma edge and on the PFC surface, as measured by the discharge length and access to the H-mode regime. In this work, we present a quantitative analysis of the evolution of the chemistry of the ATJ surface, and the oxygen presence in the plasma-material interface, and report relevant plasma parameters observed during the same period of time.