A novel method for treating MAR in EMC3-Eirene, and first applications to W7-X

A system for studying the spatiotemporal dynamics of fluctuations in the boundary of the W7-X plasma using the "Gas-Puff Imaging" (GPI) technique has been designed, constructed, installed, and operated. This GPI system addresses a number of challenges specific to long-pulse superconducting devices, such as W7-X, including the long distance between the plasma and the vacuum vessel wall, the long distance between the plasma and diagnostic ports, the range of last closed flux surface (LCFS) locations for different magnetic configurations in W7-X, and management of heat loads on the system's plasma-facing components. The system features a pair of "converging-diverging" nozzles for partially collimating the gas puffed locally ≈135mm radially outboard of the plasma boundary, a pop-up turning mirror for viewing the gas puff emission from the side (which also acts as a shutter for the re-entrant vacuum window), and a high-throughput optical system that collects visible emission resulting from the interaction between the puffed gas and the plasma and directs it along a water-cooled re-entrant tube directly onto the 8 × 16 pixel detector array of the fast camera. The DEGAS 2 neutral code was used to simulate the Hα (656nm) and HeI (587nm) line emission expected from well-characterized gas-puffs of H2 and He and excited within typical edge plasma profiles in W7-X, thereby predicting line brightnesses used to reduce the risks associated with system sensitivity and placement of the field of view. Operation of GPI on W7-X shows excellent signal-to-noise ratios (>100 at 2 Mframes/s) over the field of view for minimally perturbing gas puffs. The GPI system provides detailed measurements of the two-dimensional (radial and poloidal) dynamics of plasma fluctuations in the W7-X edge and scrape-off layer and in and around the magnetic islands outside the LCFS that make up the island divertor configuration employed on W7-X.

Soft x-ray (SX) tomography is a useful diagnostic in fusion research, and a multi-channel SX diagnostic will be installed in JT-60SA, the largest elongated tokamak in the world. However, in the SX diagnostic of JT-60SA, plasmas will be only viewed from the low field side and the upper side of plasmas; the sight lines are limited, which would be common in future devices as well as JT-60SA. This kind of limited sight lines is not preferred for SX tomography to investigate the spatial structure of magnetohydrodynamics (MHD) modes because inadequate information of plasmas makes artifacts in the reconstructed SX profiles. One of the solutions to reduce the artifacts is to employ L1 regularization, which gives the essential and sparse contributions [Kaptanoglu etal., Phys. Plasmas 30, 033906 (2023)]. In this study, as a first topic, the applicability of L1 regularization to reduce the artifacts in SX tomography with limited sight lines is investigated with traditional L2 regularization for a high beta scenario of JT-60SA where MHD modes would occur. Here, as a series of basis functions, the Fourier-Bessel series (FBS) is employed because FBS has the poloidal Fourier modes explicitly. A disadvantage of FBS is that the accurate equilibrium inside the last closed flux surface (LCFS) is needed; interior measurement such as the motional Stark effect measurement is required, which is not always available during a whole discharge. The second topic of this study is to investigate other appropriate basis functions to study the spatial structure of MHD modes in elongated tokamak plasmas. Here, we introduce Saito's Laplacian eigenfunction (LEF). Saito's LEF can be calculated if LCFS is given and the LEF is expected to show the explicit poloidal Fourier mode. Because the calculation of LCFS with magnetic measurements is a basic task of plasma operations, Saito's LEF may be used anytime. Our investigation showed that L1 regularization can strongly improve the SX tomography with the traditional L2 regularization having FBS/LEF and would be effective against other tomographic problems in fusion devices.

In the Large Helical Device (LHD), direct oblique launching of the fundamental extraordinary (X-) mode from the high magnetic field side (HFS) is available without installation of any additional launching equipment on the inner side wall of the torus. In the experiment, power absorption was observed in two separated regions by the X-mode launching. The central electron density was about 8% of the cutoff density. The result of numerical analysis with the ray-tracing calculation suggests that most power of the launched X-mode is damped out as the X-mode in the fundamental electron cyclotron resonance (ECR) layer before it reaches the upper hybrid resonance (UHR) layer where the electron Bernstein wave (EBW) occurs which is excited via the slow X (SX-) B mode conversion process. Only about 0.2% of the launched power is mode converted to the EBW and is then absorbed at a maximum. One of the two separated power absorption regions observed in the experiment agrees well with the power absorption region of the X-mode suggested by the ray tracing. The other one agrees well with that of the O-mode despite the setting of the X-mode launching. Supposedly mixed waves of the X- and the O-mode might be launched in the experiment. We assumed that the incident transverse electromagnetic waves in vacuum couple with the electromagnetic modes in the plasma at the last closed flux surface (LCFS). However, the coupling point was supposedly located outside the LCFS.For the few rays that can reach the UHR layer we have recognized that the parallel component of the refractive index N∥ becomes close to zero and power absorption as the X-mode weakens when the rays pass through the fundamental ECR layer. A numerical investigation assuming a higher central electron density, that is 23% of the cutoff density, suggests a scenario of effective EBW excitation. The ray that passes through the centre of the focused Gaussian beam launched from the HFS so that N∥ becomes close to zero near the ECR layer, Z can reach the UHR layer without being damped out and excite EBW. About 71% of power of the launched X-mode is mode converted to the EBW and absorbed in the Doppler-shifted ECR layer.Observation of the parametric decay waves suggests that the decay wave was excited in the ‘exterior’ UHR layer which is located outside the LCFS before the launched X-mode reaches the HFS. The combination of tunnelling, reflection and mode-coupling processes, the so-called ‘Budden's problem’, is suggested to occur in the evanescent region between the ‘exterior’ right handed cyclotron cutoff and the UHR layer outside the LCFS. The tunnelling rate should be considered for estimating the power that can penetrate inside the LCFS from the HFS, in particular in the higher density regime where the excitation of the EBW is expected.

The trajectories and the ablation positions of boron dust particles dropped from an impurity powder dropper in the peripheral plasma in the Large Helical Device (LHD) were calculated using a three‐dimensional edge plasma simulation code (EMC3‐EIRENE) and a dust transport simulation code (DUSTT). The simulation shows that the trajectory of the boron dust particles is deflected at the upper divertor leg due to the effect of the hydrogen plasma flow, and the ablation positions of the dust particles in an ergodic layer change toward the outboard side of the torus for higher plasma densities. The effect of the boron ion flow in the divertor leg on the deflection is investigated by coupling the two codes self‐consistently. The simulation predicts that the boron ions in the divertor leg, which are produced by sputtering on the divertor plates, which do not affect the change in the ablation positions. It also shows that the ablation positions move toward the inboard side and approach the Last Closed Flux Surface (LCFS) in case of increased boron dust drop rates, which is caused by the lowered plasma flow in the upper divertor leg due to the lowered electron temperature by radiation cooling by the dropped dust particles.

Measurements of electrostatic fluctuations on the edge of the Advanced Toroidal Facility (ATF) torsatron are used to study the role of the edge turbulence in the particle transport in this current-free magnetic configuration. Spatial profiles of the plasma electron density n{sub e}, temperature T{sub e}, and fluctuations in density ({tilde n}{sub e}) and in the plasma floating potential ({tilde {phi}}{sub f}) are measured at the edge in electron cyclotron heated plasmas using a Langmuir probe array. At the last closed flux surface (LCFS), r/a {approx} 1, T{sub e} {approx} 20--40 eV, and n{sub e} {approx} 10{sup 12} cm{sup {minus}3} for a line-averaged electron density {bar n}{sub e} = (3--6) {times} 10{sup 12} cm{sup {minus}3}. The relative fluctuation levels decrease as the probe is moved into the core plasma. For T{sub e} > 10 eV, {tilde n}{sub e}/n{sub e} {approx} 5%, and e{tilde {phi}}{sub f}/T{sub e} {approx} 2{tilde n}{sub e}/n{sub e} at r/a = 0.95. The measured fluctuations spectra are broadband (40--300 kHz) with {bar k}{rho}{sub s} {approx} 0.05--0.1, where {bar k} is the average wave number of the fluctuations and {rho}{sub s} is the ion Larmor radius at the sound speed. Near the LCFS, the density fluctuations can be approximated by {tilde n}{sub e}/n{sub e} {approx} 0.4/{bar k}L{sub n}, where L{sub n} is the gradient scale length of n{sub e}. The propagation direction of the fluctuations reverses to the electron diamagnetic direction at r/a < 1. The phase velocity of the fluctuations and the electron drift velocity are comparable. The particles flux estimated from the fluctuations is consistent with fluxes obtained from the particle balance using the H{sub {alpha}} spectroscopic measurements. Many of the features seen in the ATF edge fluctuations resemble those of ohmically heated plasmas in the Texas Experimental Tokamak.

Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

It is the first time that fast negative-ion loss signals are observed by Fast-ion Loss Detector (FILD) and analyzed in EAST neutral beam injection experiment. Velocity-space distributions of lost fast negative and positive ions reconstructed by the FILDSIM code are basically consistent. Orbit analysis backtracked by the ASCOT code shows that the fast negative ions are generated at the low-field side (LFS) scrape-off layer, whereas the fast positive ions are generated near the last closed flux surface (LCFS) at both the high- and LFSs. The existence of fast negative-ion loss indirectly suggests a very high neutral particle density near both the main limiter and antenna limiter. The differing birth locations of fast positive and negative ions result in distinct loss behaviors during ELM H-mode and n = 3 rotating RMP experiments. The asymmetry of neutral particle density in tokamak suggest that previous studies may have underestimated the fraction of charge exchange (CX) losses due to the assumption of a uniform neutral particle density outside the LCFS. Therefore, simulations evaluating the CX loss fraction in the future should account for a three-dimensional distribution of neutral particle density, including local gas puffing and the neutral particle density around various limiters.

Since high density operation is advantageous for building an efficient fusion reactor, understanding the density limit in tokamaks has been seen as one of the most important issues. This paper reports a series of measurements around the last-closed flux surface (LCFS) in L-mode plasmas by using a thermal helium beam diagnostic. Fluctuation analysis has been employed to characterize the poloidal flow and the turbulence structure. A reversal of the poloidal flow in the scrape-off layer and concomitant cooling of the outer divertor plasma are observed as the density is raised. While, in the confined region, the change in the density barely affects the poloidal flow, a higher density shifts the fluctuation power spectral densities toward lower frequencies and wave numbers. The eddy tilting of this region is consistent with what is expected from the magnetic shear effect. A radially coherent low frequency mode appears in the case of the highest density investigated in this study (n¯e/ne,GW = 0.51), and higher frequencies near the LCFS are modulated by this mode.

Effect of N2, Ne and Ar seeding on Alcator C-Mod H-mode confinement

Radial emission line intensity profiles of hydrogen atoms have been measured in the edge region of NBI-heated plasma. The ionization flow and the inward particle flux of the atoms are analyzed with the help of collisional-radiative (CR) model calculations. The results indicate that most of atoms are ionized before reaching last closed flux surface (LCFS). The global particle confinement time is also analyzed and estimated to be 10 ms. This value is in good agreement with the ion flux on divertor plates measured by a Langmuir probe array.

The power load to the divertor surfaces is a key concern for future devices such as ITER, due to the thermal limits on the material surface. One factor that characterizes the heat flux to the divertor is the fall off length in the scrape-off layer (SOL), which recent empirical scalings have shown could be as small as 1 mm. These predictions are based on a multi-machine scaling of the heat flux width fitted using an expression for the divertor heat flux profile which includes a term for the exponential decay in the SOL and diffusion about the last closed flux surface (LCFS) in the private flux region. This expression has been used to fit a database of inter-ELM H mode profiles at the upper divertor and extract the fall off length, λq, for a range of different plasma parameters in double null plasmas. The midplane separation between the primary and secondary LCFS (δrsep) of the double null plasma used in the study are in the range 2 ⩽ δrsep ⩽ 7 mm and no correlation is seen between the fall off length and the δrsep. The MAST data shows good agreement with the formula, with the fitted fall off length spanning a range of 5–11 mm in the data base generated. Regression of this data has shown that the fall off length has the strongest dependence on the plasma current (or equivalently, the poloidal magnetic field at the outboard midplane) to the power −0.71. The scaling with the smallest χ2 error utilizes the poloidal magnetic field at the outboard midplane (Bpol,omp) and the power crossing the SOL in the relation with χ2 = 3.46 and R2 = 0.56 as a goodness of fit. The equivalent scaling with plasma current is with χ2 = 3.84 and R2 = 0.55. The moderate goodness of fit suggests that additional plasma parameters are required to accurately reproduce the observed variation in λq.

Development of an ICRH antenna system at W7-X for plasma heating and wall conditioning

A fast video camera is verified to be a powerful tool for observation of filaments/blobs near the last closed flux surface (LCFS). In order to extract the fluctuation component from the raw data of the fast camera, a pre-processing technique, sliding time window averaging subtraction (STWAS) has been developed to remove the background of slowly varying emission from the bulk plasma. By using this pre-processing technique, the fast camera data are analyzed. A method to identify the filaments in the pre-processed image is also discussed.

Dedicated experiments in the HT-7 tokamak were performed to investigate the influence of erosion and deposition on the mirror samples. The first mirror (FM) samples made of polycrystalline (PC) stainless steel (SS), molybdenum (Mo) and tungsten (W) were fixed on a holder at an angle of 45° with respect to the horizontal plane and set at different locations with different connection lengths along the magnetic field. The optical reflectivity of the first mirror was measured by a spectrophotometer before and after plasma exposure. It was found that the surface morphology and specular reflectivity of the mirror samples after the exposure were different with respect to the different distances from the mirror surface to the last closed flux surface (LCFS) of the plasma in the tokamak. It was also found that shortening the connection length before the mirror surface would weaken the influence of the plasma erosion and impurity deposition on the mirror surface. In order to maintain the optical characteristics of the mirror surface, it is necessary to adopt the in-situ cleaning and mirror protection techniques.

Nonneutral plasmas confined in the helical magnetic surface (HMS) region can be produced by injecting electrons from outside the last closed flux surface (LCFS). Recently, we numerically calculated outward electron orbits that extend to the inner HMS region. Once it penetrates the HMS region, the injected electron is never lost to the chamber wall, because the negative self-electric potential φs in the stochastic magnetic region (SMR) acts as a potential barrier. Remarkably, during the reflection process at the potential barrier, the electron is still trapped in the foot of the negative φs region, although it completely overcomes the SMR. The electron then resumes inward movement across the HMS.