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The first neutral beam injector (NBI) experiments of the Wendelstein 7-X stellarator took place in summer 2018. The modelling of the fast ion production and slowing down processes predicts losses of the NBI fast ions to the first wall on the order of 15%. One location receiving a high load (possibly peaking at several M W/m2) is the immersion tube for optical and infrared monitoring of the divertor targets. The stainless steel face of the tube has three vacuum windows, which are sensitive to temperature gradients and overheating. To protect the windows from damage caused by the fast ions, different heat load mitigation techniques were investigated. Given the available time and resources until the first NBI experiments, a protective stainless steel collar mounted at the front of the immersion tubes was regarded the most realistic solution. This contribution describes the fast ion modelling of the loads, the new design, thermal modelling of the design, and finally experimental experience with the protective collar showing heat loads in excess of 1.5 M W/m2. The fast ion heat loads have been assessed computationally with the ASCOT code and experimentally with thermal imaging.

In the previous divertor campaign, the Wendelstein 7-X (W7-X) device injected 3.6 MW of neutral beam heating power allowing for the achievement of densities approaching 2 × 1020 m−3, and providing the first initial assessment of fast ion confinement in a drift optimized stellarator. The neutral beam injection (NBI) system on W7-X is comprised of two beam boxes with space for four radio frequency sources each. The 3.6 MW of heating reported in this work was achieved with two sources in the NI21 beam box. The effect of combined electron-cyclotron resonance heating (ECRH) and NBI was explored through a series of discharges varying both NBI and ECRH power. Discharges without ECRH saw a linear increase in the line-integrated plasma density, and strong peaking of the core density, over the discharge duration. The presence of 1 MW of ECRH power was found to be sufficient to control a continuous density rise during NBI operation. Simulations of fast ion wall loads were found to be consistent with experimental infrared camera images during operation. In general, NBI discharges were free from the presence of fast ion induced Alfvénic activity, consistent with low beam betas. These experiments provide data for future scenario development and initial assessment of fast-ion confinement in W7-X, a key topic of the project.

Neutron production rates in fusion devices are determined not only by the kinetic profiles but also the fast ion slowing-down distributions. In this work, we investigate the effect of magnetic configuration on neutron production rates in future deuterium plasmas in the Wendelstein 7-X (W7-X) stellarator. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from thermonuclear and beam-target fusion. The 14.1 MeV neutron production rates were found to be between $1.49 \times 10^{12}$ $\mathrm{s}^{-1}$ and $1.67 \times 10^{12}$ $\mathrm{s}^{-1}$, which is estimated to be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges., 17 pages, 10 figures. Version submitted to Plasma Physics and Controlled Fusion

The Collective Thomson Scattering (CTS) diagnostic measures the scattering spectrum of incident radiation off collective fluctuations in plasmas. In Wendelstein 7-X (W7-X) the diagnostic uses a 140 GHz heating gyrotron as a source of the probing radiation. At this frequency, the CTS spectra are heavily affected by the electron cyclotron emission, and the microwave beam propagation is restricted at typical W7-X plasma parameters. The diagnostic was successfully commissioned in the last experimental campaign and demonstrated ion temperature measurements. However, the signal-to-noise ratio was too low for measuring other quantities such as the fast-ion velocity distribution function or the fuel ion ratio. Currently, the W7-X CTS diagnostic is undergoing an upgrade to a frequency of 175 GHz. This will increase the sensitivity of the diagnostic, since the noise due to electron cyclotron emission will be reduced, and it will relax the constraints on microwave beam propagation in W7-X. Here we present the salient features of the upgraded CTS system and discuss its prospects for both thermal-ion and fast-ion measurements.

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

openaire: EC/H2020/633053/EU//EUROfusion Fast ion confinement is of major importance for the ignition of a burning fusion plasma. In future deuterium plasma campaigns of the Wendelstein 7-X stellarator, W7-X, the amount of triton burn-up is one possible measure for fast ion confinement. A well-established technique to observe triton burn-up is the 14 MeV neutron rate. In this paper, it is estimated whether an existing scintillating fibre neutron detector is also suited to measure triton burn-up in W7-X with sufficient accuracy. An estimation is presented, which can be applied to any tokamak or stellarator design and is one-dimensional in the minor radius. The inputs are profiles of density, temperature, and differential volume element as well as the triton slowing-down time. The estimation calculates the thermal deuteron fusion rate and the associated deuteron-triton fusion rate; thus, the triton burn-up generated 14 MeV neutron rate. It neither takes triton diffusion nor explicit losses into account. This thermally generated fusion rate is comparedto the neutral beam injection heating induced beam-plasma fusion rate.

The first, of two planned, neutral beam injectors for the stellarator Wendelstein 7-X (W7-X) was commissioned for and participated in the experimental campaign (OP1.2b) from July to October 2018. The injector was equipped with two RF driven ion sources from which 90A of positive hydrogen ions were extracted at 55 kV. After neutralization, the two sources provided >3 MW of neutral beam heating power to the stellarator plasma. During the experimental campaign >300 shots were successfully performed for plasma heating or to allow for measurement of the ion temperature profile by charge exchange recombination spectroscopy (CXRS). The initial operation of the NBI system on W7-X was very successful, demonstrating both increased central plasma density and stored plasma energy, or allowing for the collection of the time resolved ion temperature over the bulk of the plasma. This paper presents, briefly, source conditioning and performance, focusing on some of the initial problems encountered. Described in more detail is the most significant challenge overcome during commissioning: failure to couple RF power into one of the sources.

In this work, a Serpent 2 neutronics model of the Wendelstein 7-X (W7-X) stellarator is prepared, and an response function for the Scintillating-Fibre neutron detector (SciFi) is calculated using the model. The neutronics model includes the simplified geometry for the key components of the stellarator itself as well as the torus hall. The objective of the model is to assess the 14.1 MeV neutron flux from deuteron-triton fusions in W7-X, where the neutrons are modelled only until they have slowed down to 1 MeV energy. The key messages of this article are: demonstration of unstructured mesh geometry usage for stellarators, W7-X in particular; technical documentation of the model and first insights in fast neutron behaviour in W7-X, especially related to the SciFi: the model indicates that the superconducting coils are the strongest scatterers and block neutrons from large parts of the plasma. The back-scattering from e.g. massive steel support structures is found to be small. The SciFi will detect neutrons from an extended plasma volume in contrast to having an effective line-of-sight., Comment: Proceedings of the SOFT2020 conference, to be sumbitted to Fusion Engineering and Design

openaire: EC/H2020/633053/EU//EUROfusion The neutral beam deposition model in the BEAMS3D code is validated against neutral beam attenuation data from Wendelstein 7-X (W7-X). A set of experimental discharges where the neutral beam injection system of W7-X was utilized were reconstructed. These discharges scanned the magnetic configurations and plasma densities of W7-X. The equilibrium reconstructions were performed using STELLOPT which calculates three-dimensional self-consistent ideal magnetohydrodynamic equilibria and kinetic profiles. These reconstructions leveraged new capabilities to incorporate electron cyclotron emission and X-ray imaging diagnostics in the STELLOPT code. The reconstructed equilibria and profiles served as inputs for BEAMS3D calculations of neutral beam deposition in W7-X. It is found that if reconstructed kinetic profiles are utilized, good agreement between measured and simulated beam attenuation is found. As deposition models provide initial conditions for fast-ion slowing down calculations, this work provides a first steptowards validating our ability to predict fast ion confinement in stellarators.

ASCOT Fusion Source Integrator AFSI, an efficient tool for calculating fusion reaction rates and characterizing the fusion products, based on arbitrary reactant distributions, has been developed and is reported in this paper. Calculation of reactor-relevant D-D, D-T and D-3He fusion reactions has been implemented based on the Bosch-Hale fusion cross sections. The reactions can be calculated between arbitrary particle populations, including Maxwellian thermal particles and minority energetic particles. Reaction rate profiles, energy spectra and full 4D phase space distributions can be calculated for the non-isotropic reaction products. The code is especially suitable for integrated modelling in self-consistent plasma physics simulations as well as in the Serpent neutronics calculation chain. Validation of the model has been performed for neutron measurements at the JET tokamak and the code has been applied to predictive simulations in ITER.

Neutral particle analysis (NPA) is one of the few methods for diagnosing fast ions inside a plasma by measuring neutral atom fluxes emitted due to charge exchange reactions. The JET tokamak features an NPA diagnostic which measures neutral atom fluxes and energy spectra simultaneously for hydrogen, deuterium and tritium species. A synthetic NPA diagnostic has been developed and used to interpret these measurements to diagnose energetic particles in JET plasmas with neutral beam injection (NBI) heating. The synthetic NPA diagnostic performs a Monte Carlo calculation of the neutral atom fluxes in a realistic geometry. The 4D fast ion distributions, representing NBI ions, were simulated using the Monte Carlo orbit-following code ASCOT. Neutral atom density profiles were calculated using the FRANTIC neutral code in the JINTRAC modelling suite. Additionally, for rapid analysis, a scan of neutral profiles was precalculated with FRANTIC for a range of typical plasma parameters. These were taken from the JETPEAK database, which includes a comprehensive set of data from the flat-top phases of nearly all discharges in recent JET campaigns. The synthetic diagnostic was applied to various JET plasmas in the recent hydrogen campaign where different hydrogen/deuterium mixtures and NBI configurations were used. The simulated neutral fluxes from the fast ion distributions were found to agree with the measured fluxes, reproducing the slowing-down profiles for different beam isotopes and energies and quantitatively estimating the fraction of hydrogen and deuterium fast ions.

A study of the resonant magnetic perturbation (RMP) effect on transit beam ion behavior is performed using the total neutron emission rate (Sn) measurement of the deuterium plasma in the Large Helical Device. We conducted no RMP field, one-half RMP field, and full RMP field discharges and compared Sn that reflects the global beam ion confinement information. It is determined that owing to the RMP field, Sn decreased by approximately 15 - 30%. Numerical calculations based on the classical confinement of beam ions were performed to investigate the bulk plasma parameter effect on Sn. The calculated Sn shows that the degradation of Sn by RMP is mainly caused by the degradation of the electron temperature owing to island formation which results in a shorter slowing down time of beam ions.

With the objective of understanding the energetic-particle loss mechanism in three-dimensional plasmas, a Faraday-cup-based fast-ion loss detector (FILD) was developed and installed in OP1.2b in Wendelstein 7-X (W7-X) as a collaboration between the National Institute for Fusion Science and Max-Planck-Institute for Plasma Physics. The FILD, which consists of double apertures and thin aluminum foils, was based on a magnetic spectrometer using the magnetic field of the fusion device. The double aperture limits the thermal ion, but allows the energetic ion to enter the FILD box. The thin aluminum foils serve as the ion collector. Orbit-following calculations were performed in order to find a suitable position for the FILD. The results indicated that barely co- and counter-going transit ions reached the FILD position mounted on the multi-purpose manipulator installed on W7-X. Moreover, because the injection angle of neutral beams injector installed on W7-X was relatively perpendicular, the target range of pitch angle was set to from 91 degrees to 150 degrees. An energy-and-pitch-angle map was created using a grid calculation code in order to decide on the position/size of aperture and aluminum foil. The grid calculation code indicates the position where the energetic ion would strike on the aluminum foil. Here, the map was used as the basis for designing an aluminum foil pattern with two energies and four pitch angle ranges. Also, note that the lower energy range was designed so that it accommodates a neutral beam injection energy of 55 kV. Measurements of beam ion losses were performed in neutral beam (NB) blip experiments, where concurrent increases and decreases of barely co-going transit beam ion losses due to NB injections were observed. Furthermore, this study validated that the FILD installed on a midplane manipulator probe is capable of probing a range of radii spanning 1.0–1.5 cm, over which the beam ion loss current would vary significantly.

We model the transport of a beryllium (10Be) marker during a sequence of an inner-wall limited and a diverted Ohmic plasma phase in JET with the objective of identifying principal migration pathways. The 3D orbit-following code ASCOT is used for predictive analysis of an experiment during the 2011–2012 campaign on JET where three central pieces of a wall tile enriched with 10Be were installed to an inner wall guard limiter (IWGL) of the tokamak. Assuming erosion during the inner-wall limited plasma, the simulations indicate that 10Be is deposited along the IWGLs during the limiter phase which, when assuming further erosion, can lead to high deposition on the inner (high-field side) divertor during the diverted phase. In contrast, beryllium confined in the core plasma during the limiter phase is seen to be predominantly uniformly deposited during the diverted phase on the outer (low-field side) wall limiters and divertor tiles.

International audience; Recent developments and achievements of the EUROfusion Code Development for Integrated Modelling project (WPCD), which aim is to provide a validated integrated modelling suite for the simulation and prediction of complete plasma discharges in any tokamak, are presented. WPCD develops generic complex integrated simulations, workflows, for physics applications, using the standardized European Integrated Modelling (EU-IM) framework. Selected physics applications of EU-IM workflows are illustrated in this paper.

Mitigating edge localized modes (ELMs) with resonant magnetic perturbations (RMPs) can increase energetic particle losses and resulting wall loads, which have previously been studied in the vacuum approximation. This paper presents recent results of fusion alpha and NBI ion losses in the ITER baseline scenario modelled with the Monte Carlo orbit following code ASCOT in a realistic magnetic field including the effect of the plasma response. The response was found to reduce alpha particle losses but increase NBI losses, with up to 4.2% of the injected power being lost. Additionally, some of the load in the divertor was found to be shifted away from the target plates toward the divertor dome.

We assess the effect of the European design of the pebble-bed helium-cooled test blanket modules (TBM) on fast ion power loads on ITER material surfaces. For this purpose, the effect of not only the TBMs but also the ferritic inserts (FI), used for mitigating the toroidal field ripple, were included in unprecedented detail in the reconstruction of the 3-dimensional magnetic field. This is important because, due to their low collisionality, fast ions follow the magnetic geometry much more faithfully than the thermal plasma. The Monte Carlo orbit-following code ASCOT was used to simulate all the foreseen operating scenarios of ITER: the baseline 15 MA standard H-mode operation, the 12.5 MA hybrid scenario, the 9 MA advanced scenario, and the half-field scenario with helium plasma that will be ITER’s initial operating scenario. The effect of TBMs was assessed by carrying out the simulations in pairs: one including only the effect of ferritic inserts, and the other including also the perturbation due to TBMs. Both thermonuclear fusion alphas and NBI ions from ITER heating beams were addressed. The TBMs are found to increase the power loads, but the absolute values remain small. Neither do they produce any additional hot spots.

The magnetic perturbation due to the ferromagnetic test blanket modules (TBMs) may deteriorate fast ion confinement in ITER. This effect must be quantified by numerical studies in 3D. We have implemented a combined finite element method (FEM) -- Biot-Savart law integrator method (BSLIM) to calculate the ITER 3D magnetic field and vector potential in detail. Unavoidable geometry simplifications changed the mass of the TBMs and ferritic inserts (FIs) up to 26%. This has been compensated for by modifying the nonlinear ferromagnetic material properties accordingly. Despite the simplifications, the computation geometry and the calculated fields are highly detailed. The combination of careful FEM mesh design and using BSLIM enables the use of the fields unsmoothed for particle orbit-following simulations. The magnetic field was found to agree with earlier calculations and revealed finer details. The vector potential is intended to serve as input for plasma shielding calculations., In proceedings of the 28th Symposium on Fusion Technology

This paper explains how to obtain the distribution function of minority ions in tokamak plasmas using the Monte Carlo method. Since the emphasis is on energetic ions, the guiding-center transformation is outlined, including also the transformation of the collision operator. Even within the guiding-center formalism, the fast particle simulations can still be very CPU intensive and, therefore, we introduce the reader also to the world of high-performance computing. The paper is concluded with a few examples where the presented method has been applied., ITER International School 2014

The effect of ASDEX Upgrade (AUG) edge localized mode (ELM)-mitigation coils on fast ion wall loads was studied with the fast particle following Monte Carlo code ASCOT. Neutral beam injected particles were simulated in two AUG discharges both in the presence and in the absence of the magnetic field perturbation induced by the eight newly installed in-vessel coils. In one of the discharges (#26476) beams were applied individually, making it a useful basis for investigating the effect of the coils on different beams. However, no ELM mitigation was observed in #26476, probably due to the low plasma density. Therefore, another discharge (#26895) demonstrating clear ELM mitigation was also studied. The magnetic perturbation due to the in-vessel coils has a significant effect on the fast particle confinement, but only when total magnetic field, B tot, is low. When B tot was high, the perturbation did not increase the losses, but merely resulted in redistribution of the wall power loads. Hence, it seems to be possible to achieve ELM mitigation using in-vessel coils, while still avoiding increased fast ion losses, by simply using a strong B tot. Preliminary comparisons between simulated and experimental fast ion lost detector signals show a reasonable correspondence.

This paper addresses the confinement of thermonuclear alpha particles and neutral beam injected deuterons in the 15 MA Q = 10 inductive scenario in the presence of the magnetic perturbation caused by the helium cooled pebble bed test blanket module using the vacuum approximation. Both the flat top phase and plasma ramp-up are studied. The transport of fast ions is calculated using the Monte Carlo guiding center orbit-following code ASCOT. A detailed three-dimensional wall, derived from the ITER blanket module CAD data, is used for evaluating the fast ion wall loads. The effect of the test blanket module is studied for both overall confinement and possible hot spots. The study indicates that the test blanket modules do not significantly deteriorate the fast ion confinement.

In this study, 3D orbit-following simulations with the ASCOT code of a tracer injection experiment on ASDEX Upgrade (AUG) are carried out to investigate the influence of toroidal field (TF) ripple and resonant magnetic perturbations (RMPs) on global 13 C transport in the scrape-off layer (SOL). The conducted ASCOT simulations predict ripple to induce toroidal periodicity into the 13 C deposition pattern. For resonant magnetic perturbations, the deposition pattern was observed to be locally influenced in the vicinity of the in-vessel coils creating the perturbations. In both cases, however, the global large-scale deposition of 13 C remained unchanged compared to the case of an axisymmetric magnetic field.

A comprehensive description of methods, suitable for solving the kinetic equation for fast ions and impurity species in tokamak plasmas using Monte Carlo approach, is presented. The described methods include Hamiltonian orbit-following in particle and guiding center phase space, test particle or guiding center solution of the kinetic equation applying stochastic differential equations in the presence of Coulomb collisions, neoclassical tearing modes and Alfv\'en eigenmodes as electromagnetic perturbations relevant to fast ions, together with plasma flow and atomic reactions relevant to impurity studies. Applying the methods, a complete reimplementation of the well-established minority species code ASCOT is carried out as a response both to the increase in computing power during the last twenty years and to the weakly structured growth of the code, which has made implementation of additional models impractical. Also, a benchmark between the previous code and the reimplementation is accomplished, showing good agreement between the codes., Comment: 13 pages, 9 figures, submitted to Computer Physics Communications

Phase-space time-resolved measurements of fast-ion losses induced by edge localized modes (ELMs) and ELM mitigation coils have been obtained in the ASDEX Upgrade tokamak by means of multiple fast-ion loss detectors (FILDs). Filament-like bursts of fast-ion losses are measured during ELMs by several FILDs at different toroidal and poloidal positions. Externally applied magnetic perturbations (MPs) have little effect on plasma profiles, including fast-ions, in high collisionality plasmas with mitigated ELMs. A strong impact on plasma density, rotation and fast-ions is observed, however, in low density/collisionality and q(95) plasmas with externally applied MPs. During the mitigation/suppression of type-I ELMs by externally applied MPs, the large fast-ion bursts observed during ELMs are replaced by a steady loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection (NBI) prompt loss signal without MPs. Multiple FILD measurements at different positions, indicate that the fast-ion losses due to static 3D fields are localized on certain parts of the first wall rather than being toroidally/poloidally homogeneously distributed. Measured fast-ion losses show a broad energy and pitch-angle range and are typically on banana orbits that explore the entire pedestal/scrape-off-layer (SOL). Infra-red measurements are used to estimate the heat load associated with the MP-induced fast-ion losses. The heat load on the FILD detector head and surrounding wall can be up to six times higher with MPs than without 3D fields. When 3D fields are applied and density pump-out is observed, an enhancement of the fast-ion content in the plasma is typically measured by fast-ion D-alpha (FIDA) spectroscopy. The lower density during the MP phase also leads to a deeper beam deposition with an inward radial displacement of approximate to 2 cm in the maximum of the beam emission. Orbit simulations are used to test different models for 3D field equilibrium reconstruction including vacuum representation, the free boundary NEMEC code and the two-fluid M3D-C1 code which account for the plasma response. Guiding center simulations predict the maximum level of losses, approximate to 2.6%, with NEMEC 3D equilibrium. Full orbit simulations overestimate the level of losses in 3D vacuum fields with approximate to 15% of lost NBI ions.

The impact of edge localized modes (ELMs) and externally applied resonant and non-resonant magnetic perturbations (MPs) on fast-ion confinement/transport have been investigated in the ASDEX Upgrade (AUG), DIII-D and KSTAR tokamaks. Two phases with respect to the ELM cycle can be clearly distinguished in ELM-induced fast-ion losses. Inter-ELM losses are characterized by a coherent modulation of the plasma density around the separatrix while intra-ELM losses appear as well-defined bursts. In high collisionality plasmas with mitigated ELMs, externally applied MPs have little effect on kinetic profiles, including fast-ions, while a strong impact on kinetic profiles is observed in low-collisionality, low q 95 plasmas with resonant and non-resonant MPs. In low-collisionality H-mode plasmas, the large fast-ion filaments observed during ELMs are replaced by a loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection prompt loss signal without MPs. A clear synergy in the overall fast-ion transport is observed between MPs and neoclassical tearing modes. Measured fast-ion losses are typically on banana orbits that explore the entire pedestal/scrape-off layer. The fast-ion response to externally applied MPs presented here may be of general interest for the community to better understand the MP field penetration and overall plasma response. Ministerio de Economía y Empresa ((RYC-2011-09152 y ENE2012-31087) Marie Curie (Grant PCIG11-GA-2012-321455) US Department of Energy (DE-FC02-04ER54698, SC-G903402, DE-FG02-04ER54761, DE-AC02-09CH11466 and DE-FG02- 08ER54984) NRF Korea contract 2009-0082012 MEST under the KSTAR project

The fusion alpha and beam ion with steady-state power loads in all four main operating scenarios of ITER have been evaluated by the ASCOT code. For this purpose, high-fidelity magnetic backgrounds were reconstructed, taking into account even the internal structure of the ferritic inserts and tritium breeding modules (TBM). The beam ions were found to be almost perfectly confined in all scenarios, and only the so-called hybrid scenario featured alpha loads reaching 0.5 MW due to its more triangular plasma. The TBMs were not found to jeopardize the alpha confinement, nor cause any hot spots. Including plasma response did not bring dramatic changes to the load. The ELM control coils (ECC) were simulated in the baseline scenario and found to seriously deteriorate even the beam confinement. However, the edge perturbation in this case is so large that the sources have to be re-evaluated with plasma profiles that take into account the ECC perturbation.

Computations in toroidal geometry are systematically performed for the plasma response to 3D magnetic perturbations produced by ferritic inserts (FIs) and test blanket modules (TBMs) for four ITER plasma scenarios: the 15 MA baseline, the 12.5 MA hybrid, the 9 MA steady state, and the 7.5 MA half-field helium plasma. Due to the broad toroidal spectrum of the FI and TBM fields, the plasma response for all the n = 1–6 field components are computed and compared. The plasma response is found to be weak for the high-n (n > 4) components. The response is not globally sensitive to the toroidal plasma flow speed, as long as the latter is not reduced by an order of magnitude. This is essentially due to the strong screening effect occurring at a finite flow, as predicted for ITER plasmas. The ITER error field correction coils (EFCC) are used to compensate the n = 1 field errors produced by FIs and TBMs for the baseline scenario for the purpose of avoiding mode locking. It is found that the middle row of the EFCC, with a suitable toroidal phase for the coil current, can provide the best correction of these field errors, according to various optimisation criteria. On the other hand, even without correction, it is predicted that these n = 1 field errors will not cause substantial flow damping for the 15 MA baseline scenario.

The magnetic perturbations due to in-vessel coils, foreseen to mitigate edge-localized modes (ELMs) in ITER, could also compromise the confinement of energetic ions. We simulate the losses of fusion alpha particles and neutral beam injection-generated fast ions in ITER under the influence of the 3D perturbations caused by toroidal field coils, ferritic inserts, test blanket modules and ELM control coils (ECCs) with the ASCOT code. The ECCs are found to stochastize the magnetic field deep inside the pedestal in the 15 MA inductive reference operating scenario. Such a field is found insufficient to confine not only the fast but also the thermal ion population, leading to a strongly reduced fast ion source in the edge. Therefore, even with a stochastic edge, no high fast ion power loads are expected. However, the plasma response has not yet been included in the calculation of ITER magnetic background data, and it is probable that the perturbation is currently overestimated.

A multi-purpose overview video system, based on EDICAM cameras, was set up at Wendelstein 7-X stellarator, in order to fulfill both machine protection and scientific observation purposes. Places of strong plasma-wall interaction, which can easily evolve to hot-spots, were detected by the EDICAM operators during plasma operation, based on intense local light emission. The EDICAM system was successfully used to avoid hot-spot formation during the commissioning of magnetic configurations with plasma operation. Featuring non-destructive readout capability, smaller areas of the torus interior could be monitored ca. two orders of magnitude faster, in parallel to the normal full frame overview. These fast measurements could be used to show the presence of plasma turbulence (filaments), also detected by other diagnostic systems.

Équipe 107 : Physique des plasmas chauds; International audience; The European Union Integrated Tokamak Modelling Task Force (ITM-TF) has developed a standardized platform and an integrated modeling suite of codes for the simulation and prediction of a complete plasma discharge in any tokamak. The framework developed by ITM-TF allows for the development of sophisticated integrated simulations (workflows) for physics application, e.g., free-boundary equilibrium with feedback control, magnetohydrodynamic stability analysis, core/edge plasma transport, and heating and current drive. A significant effort is also under way to integrate synthetic diagnostic modules in the ITM-TF environment, namely, focusing on three-dimensional reflectometry, motional Stark effect, and neutron and neutral particle analyzer diagnostics. This paper gives an overview of the conceptual design of ITM-TF and preliminary results of the aforementioned synthetic diagnostic modules

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 × 1019 m−3 , central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre. EURATOM 633053

The neutral beam injection (NBI) system at Wendelstein 7-X (W7-X) was operated for the first time in 2018. Detailed calorimetric measurements were carried out to accurately determine the energy flow within the NBI system and the injected power into the plasma. Of the electric energy put into the system (89± 6) % can be accounted for. The time-averaged injected NBI power is assessed to be Pinj ≈ (3.1 ± 0.8) MW for hydrogen injection with two positive ion sources at the maximum acceleration voltage of 54 kV. This corresponds to a conversion efficiency of the initially generated ion beam power to the injected neutral beam power of about 33%. The ionization in a plasma with an ion and electron temperature of Ti ≈ Te ≈ 1.3 keV and an electron density of ne ≈ 4 ·1019 m−3 is calorimetrically determined to be about 90%. The injected power is validated against beam emission spectroscopy applying Fast-Ion Dα analysis (FIDASIM). The inferred injected power leads to an energy confinement time of around τE ≈ 150 ms for pure NBI heating plasma experiments at W7-X with these heating and plasma parameters.

The modular stellarator Wendelstein 7-X (W7-X) in Greifswald (Germany) started operation in 2015 with short pulse limiter plasmas and continued with pulsed divertor plasmas in 2017−2018. In 2021, the next operation phase (OP) OP2 will start after installation of 10 water-cooled CFC armored divertors, allowing for steady state operation. Since divertor heat loads are very sensitive to plasma parameters, each water-cooled divertor needs to be monitored to interrupt or adapt plasma operation once overload is detected. For that purpose ten endoscopes are planned: two in module 3 and eight more in a different type of port in the other modules. The infrared (IR) radiation from the plasma facing surface as well as the plasma edge radiation in the visible (VIS) range is captured through a pinhole in a water-cooled plasma facing head and transmitted to the rear side outside the vacuum where the light is split and captured by an IR and VIS camera. The design challenge is to reach a high-resolution image of the entire target while capturing a large field of view (FOV) of 120 degrees. In this paper, the design and assembly strategy is presented, including the assessment of the optical, thermo-mechanical and hydraulic performance.