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Spatially resolved rotational temperature of ground state hydrogen molecules desorbed from plasma-facing surface was measured in QUEST, LTX-β, and DIII-D tokamaks, and the increases of the rotational temperature with the surface temperature and due to collisional-radiative processes in the plasmas were evaluated. The increase due to collisional-radiative processes was calculated by solving rate equations considering electron and proton collisional excitation and deexcitation and spontaneous emission. The calculation results suggest a high sensitivity for the rotational temperature to electron and proton densities, but a negligible sensitivity to the electron, proton, and surface temperatures. In the three tokamaks with different plasma parameters and plasma facing surface materials, the spatial profile of the rotational temperature was estimated using Fulcher-α emission lines (600-608 nm). In QUEST, the spatial profile of the rotational temperature was estimated from spatially resolved spectra. In the other tokamaks, the rotational temperature was evaluated assuming a single point emission with a location determined from the Fulcher-α emission profile as measured with a filtered camera. In metal-walled devices QUEST and LTX-β, the rotational temperature increased with the surface temperature, and the calculated collisional-radiative increase is consistent with measured increase assuming that the rotational temperature at the surface is approximately 450 K higher than the surface temperature. In DIII-D with carbon walls, a larger collisional-radiative increase than the other tokamaks was observed because of the higher density leading to a large difference from the calculated increase compared to the other smaller tokamaks. Measurement of the Fulcher-α emission profile with higher spatial resolution in DIII-D may reduce the difference and reveal the effect of the surface temperature on the rotational temperature. These results shows the increases in the rotational temperature with the surface temperature and due to the collisional-radiative processes.

One of the important aspects of the service life of the first and second optical mirrors (FM and SM) in ITER UWAVS diagnostics is to understand the influence of plasma cleaning on materials re-deposition and optical surface quality after multiple cleaning cycles. The capacitively coupled RF 30-60 MHz system is a candidate for the UWAVS mirror cleaning. It generates ion fluxes of tens of electron-volts, sputtering contaminants as well as construction materials. In the present report, we discuss materials re-deposition after cleaning with 30-60 MHz discharges inside the prototype of the front-end vacuum compartment. Typical ion energies were below 100 eV at the RF electrode and 20-30 eV at the grounded wall. Materials sputtered from the FM were deposited at various locations and on the SM. The deposition rates were estimated as 0.05 nm/hour or lower. The experiments showed advantages of Molybdenum as a plasma facing construction material due to lower sputtering. Traces of Fluorine were found, which may indicate cable damage at higher powers. The single-crystal Mo-mirror mock-up was etched at 0.5-1 nm/hr. rate. The mirror reflectivity showed satisfactory results after 60 h of exposure, with 50-60 nm removed. The overall effect on the reflectivity considering full service time is yet to be investigated.

First metallic mirrors used in optical diagnostics such as the Upper Wide Angle Viewing System (UWAVS) in ITER may undergo contamination with beryllium and tungsten. Contamination levels of 10 nm can significantly degrade mirror performance. A UWAVS first mirror cleaning system prototype uses radio-frequency (RF) gas discharge to remove contaminants. Although tungsten is not expected to be a main contaminant, removing it with ion sputtering may become challenging due to the high atomic mass. Cleaning tests were focused on 10–20 nm tungsten layers removal from various substrates using a tailored RF circuit in He and Ne. W-coatings were also produced on an existing single-crystal (SC) Mo-mirror 105 mm diameter. He and Ne were chosen for many tests due to their effectiveness and suitability for ITER operations. Sputtering conditions were determined through ion energy measurements. Ion energies of 100–150 eV and currents of 0.5–0.8 A/m2 were found to be optimum for 30–40 MHz RF discharges. The cleaning rate in He was 0.3–0.5 nm/hour for W-deposits. Etching rate of 1 nm/hour was observed for the Mo-mirror. No surface roughness degradation was determined with SEM and optical interferometry after mirror material was removed. Overall exposure time of the SC Mo-mirror was 100 h.

The magnetic pre-sheath (MPS) length, L MPS, is a critical parameter to define the sheath potential, which controls the ion trajectory of low-Z species (D, T, He, and C), as well as the prompt re-deposition of high-Z species. To determine L MPS, we fabricated micro-trenches (30 × 30 × 4 μm) via focused ion beam milling on a silicon surface and exposed them to L-mode deuterium plasmas in DIII-D via the divertor material evaluation system (DiMES) removable sample exposure probe. The areal distribution of impurity depositions, mainly consisting of carbon, was measured by energy-dispersive x-ray spectroscopy (EDS) to reveal the deuterium ion shadowing effect on the trench floors. The carbon deposition profiles showed that the erosion was maximized for the azimuthal direction of φ = −40° (referenced to the toroidal magnetic field direction) as well as the polar angle of θ = 80°. A Monte Carlo equation-of-motion (EOM) model, based on a collisionless MPS, was used to calculate the azimuthal and polar deuterium ion angle distributions (IADs) at the surface for a range of L MPS = k × ρ i, where ρ i is the ion gyro radius and k = 0.5–4. Then, gross erosion profiles were calculated by a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the calculated azimuthal and polar IADs for each value of k. Good agreement with the experimental C deposition profiles was obtained for the case k = 2.5–3.5. This result is consistent with a previous kinetic modeling prediction of k ∼ 3, as well as previous analytical investigations that predicted the L MPS to be several ion gyro radii. A validation of theoretical sheath models supports its applicability to ITER and pilot plant divertors to successfully predict plasma–materials interactions.

The structure of the non-axisymmetric heat load distribution at the divertor plates is determined not only by the toroidal but also from the poloidal spectrum of non-axisymmetric field perturbations. Whether they are intrinsic, like error fields, or they are applied through 3D coils, the non-axisymmetric fields produce complex 3D edge magnetic topologies (footprints) that alter the properties of the heat and particle flux distributions on the divertor target plates. In this manuscript, a study of the impact of applied 3D field poloidal spectrum on the footprint size and structure is done for the DIII-D tokamak using the resistive MHD code M3D-C1 coupled with the field line tracing code TRIP3D. To resolve the impact of the poloidal spectrum of the magnetic perturbation, the relative phase of the two rows of in-vessel 3D coils used to produce both a n = 2 and a n = 3 perturbation is varied, where n is the toroidal harmonic of the magnetic perturbation. This shows that the largest footprint is predicted when the relative phase of the two rows is close to zero, which is also where the resonant coupling with the plasma is maximized. These results suggest that it will be challenging to decouple the footprint size from the requisite resonant coupling for RMP–ELM control. The correlation between the measured heat load and particle flux distributions at the outer divertor plates in DIII-D and the magnetic measurements is in good agreement with the predicted dependence of the magnetic footprint size on the amplitude of the resonant component of the plasma response.

A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt), is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te) and density (ne) across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point) having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

The Upper port Wide Angle Viewing System (UWAVS) is diagnostic system that provides visible light and InfraRed images of the divertor area in the ITER tokamak. The main functions of the UWAVS are to provide divertor wall and plasma luminance and temperature distributions at high spatial resolution (3–5 mm) and frame rates (500–1000 Hz) for machine protection as well as to enable species influx monitoring (Be, C, O, Ne, Kr, Ar) to the divertor. In addition the UWAVS supports visual inspection as well as various other measurements. For this purpose an optical system is designed to image the divertor area onto four different cameras in the port cell area. Altogether five of these systems are foreseen to provide sufficient coverage of the divertor area (150m2). Each of these systems is equipped with active alignment to compensate for vessel motion, internal transmission monitoring sources to assess mirror reflectivity degradation, as well as calibration sources in the port cell to monitor the camera pixel response, all to guarantee the overall systems performance during operation and over lifetime. All these aspects are combined into an optical design targeting diffraction limited imaging, taking in account the complexities of the harsh ITER environment. This design and its performance are supported by manufacturability analysis, tolerancing analysis and alignment strategy and Structural Thermal Optical Performance (STOP) analysis. The latter is an TNO in-house developed method that provides a direct and reproducible link between the structural thermal analysis for a combined load case comprising nuclear heating, cooling water pressure and temperature, gravity and mounting stresses (in Ansys) and the associated changes of the optical performance, including imaging quality and alignment (in Zemax OpticStudio).

PLASMA-SURFACE INTERACTIONS 21 — Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 The triggering of rapid small edge localized modes (ELMs) by high frequency pellet injection has been proposed as a method to prevent large naturally occurring ELMs that can erode the ITER plasma facing components (PFCs). Deuterium pellet injection has been used to successfully demonstrate the ondemand triggering of edge localized modes (ELMs) at much higher rates and with much smaller intensity than natural ELMs. The proposed hypothesis for the triggering mechanism of ELMs by pellets is the local pressure perturbation resulting from reheating of the pellet cloud that can exceed the local high-n ballooning mode threshold where the pellet is injected. Nonlinear MHD simulations of the pellet ELM triggering show destabilization of high-n ballooning modes by such a local pressure perturbation. A review of the recent pellet ELM triggering results from ASDEX Upgrade (AUG), DIII-D, and JET reveals that a number of uncertainties about this ELM mitigation technique still remain. These include the heat flux impact pattern on the divertor and wall from pellet triggered and natural ELMs, the necessary pellet size and injection location to reliably trigger ELMs, and the level of fueling to be expected from ELM triggering pellets and synergy with larger fueling pellets. The implications of these issues for pellet ELM mitigation in ITER and its impact on the PFCs are presented along with the design features of the pellet injection system for ITER. 2014 Elsevier B.V. All rights reserved. 1. Introduction The H-mode confinement regime of plasma operation is planned for ITER in order to achieve high fusion performance. It is characterized by a steep pressure gradient and ‘‘pedestal’’ at the plasma edge that is expected to lead to the quasi-periodic instability of edgelocalized modes (ELMs) [1]. ELMs expel periodic bursts of particles and energy from the plasma, which if large enough can pose a serious threat to the PFCs by erosion and melting from the high heat fluxes and

Controlled runaway electron (RE) plateau-wall strikes with different initial impurity levels are used to study the effect of background plasma ion charge Z (resistivity) on RE-wall loss dynamics. It is found that Joule heating (magnetic to kinetic energy conversion) during the final loss does not go up monotonically with increasing Z but peaks at intermediate Z similar to 6. Joule heating and overall time scales of the RE final loss are found to be reasonably well-described by a basic 0D coupled-circuit model, with only the loss time as a free parameter. This loss time is found to be fairly well correlated with the avalanche time, possibly suggesting that the RE final loss rate is limited by the avalanche rate. First attempts at measuring total energy deposition to the vessel walls by REs during the final loss are made. At higher plasma impurity levels Z > 5, energy deposition to the wall appears to be consistent with modeling, at least within the large uncertainties of the measurement. At low impurity levels Z < 5, however, local energy deposition appears around 5-20x less than expected, suggesting that the RE energy dissipation at low Z is not fully understood. Published by AIP Publishing. This work was supported in part by the U.S. Department of Energy under Nos. DE-FG02-07ER54917, DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC52-07NA27344, and DE-AC05-06OR23100 and in part by the Spanish Direccion General de Investigacion Cientifica y Tecnica under Projects ENE2012–31753 and ENE2015-66444R (MINECO/FEDERE, UE). DIII-D data shown in this paper can be obtained in digital format by following the links at https://fusion.gat.com/global/D3D_DMP Publicado

Far Scrape-Off Layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma contact with the low field side (LFS) main chamber wall. In H-mode between edge localized modes (ELMs) the plasma?wall contact is weaker than in L-mode. During ELM fluxes of particles and heat to the LFS wall increase transiently above the L-mode values. Depending on the discharge conditions, ELMs are responsible for 30?90% of the net ion flux to the outboard chamber wall. ELMs in high density discharges feature intermittent transport events similar to those observed in L-mode and attributed to blobs of dense hot plasma formed inside the separatrix and propagating radially outwards. Though the blobs decay with radius, some of them survive long enough to reach the outer wall and possibly cause sputtering. In lower density H-modes, ELMs can feature blobs of pedestal density propagating all the way to the outer wall.

High-pressure gas jets of neon and argon are used to mitigate the three principal damaging effects of tokamak disruptions: thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall. The gas jet penetrates as a neutral species through to the central plasma at its sonic velocity. The injected gas atoms increase up to 500 times the total electron inventory in the plasma volume, resulting in a relatively benign radiative dissipation of >95% of the plasma stored energy. The rapid cooling and the slow movement of the plasma to the wall reduce poloidal halo currents during the current decay. The thermally collapsed plasma is very cold (∼1–2 eV) and the impurity charge distribution can include >50% fraction neutral species. If a sufficient quantity of gas is injected, the neutrals inhibit runaway electrons. A physical model of radiative cooling is developed and validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage will be more severe. Initial results of real-time disruption detection triggering gas jet injection for mitigation are shown.

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma - wall interaction. In this paper we will describe a set of experiments on DIII-D designed to provide detailed two-dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

The authors characterize the divertor target plate heat and particle fluxes that occur due to Edge-Localized-Modes (ELMs) during H-mode in DIII-D and ASDEX-Upgrade. During steady-state ELMing H-mode the fraction of main plasma stored energy lost with each ELM varies from 6% to 2% as input power increases above the H-mode power threshold. The ELM energy is deposited near the strikepoints on the divertor target plates in a fast time scale of {le} 1 ms. The spatial profile of the ELM heat pulse is flatter and broader, up to about a factor of 2, than that of the heat flux between ELMs. On ASDEX-Upgrade the inboard strike-point receives the greatest fraction, {ge} 75%, of ELM divertor heat flux, while on DIII-D the in/out split is nearly equal. The toroidal asymmetry of the heat pulse has produced a peaking factor on DIII-D of no more than 1.5. The particle flux, as measured by Langmuir probes, has also been found to be localized near the divertor strike-points. The increased particle flux during ELMs is a significant fraction of the total time-integrated divertor plate particle flux.

A comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed B{sub T} operation ({gradient}B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D{sub 2} injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions.

We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation.

Simulations of massive gas injection for disruption mitigation in DIII-D are carried out to compare the toroidal peaking of radiated power for the cases of one and two gas jets. The radiation toroidal peaking factor (TPF) results from a combination of the distribution of impurities and the distribution of heat flux associated with the mode. When ignoring the effects of strong uni-directional neutral beam injection and rotation present in the experiment, the injected impurities are found to spread helically along field lines preferentially toward the high-field-side, which is explained in terms of a nozzle equation. Therefore when considering the plasma rest frame, reversing the current direction also reverses the toroidal direction of impurity spreading. During the pre-thermal quench phase of the disruption, the toroidal peaking of radiated power is reduced in a straightforward manner by increasing from one to two gas jets. However, during the thermal quench phase, reduction in the TPF is achieved only for a particular arrangement of the two gas valves with respect to the field line pitch. In particular, the relationship between the two valve locations and the 1/1 mode phase is critical, where gas valve spacing that is coherent with 1/1 symmetry effectively reduces TPF.

The authors examine the scaling of the peak divertor heat flux and total divertor plate power in partially detached divertor (PDD) discharges in DIII-D, as a function of input power and radiated power. The peak divertor heat flux in the attached plasma increases linearly with input power, but saturates in the detached cases. The total divertor plate power remains linear with input power in both the attached and detached plasmas. This is consistent with the fact that the heat flux peak is reduced from the attached case but other areas receive increased radiant heating from the detached plasma. The divertor plate radiant heating is linear with input power because the total radiated power from the entire plasma is a linear function of input power in both attached and detached plasma. In the private flux region, radiated heat flux absorbed on the target plate calculated from bolometer data is enough to account for the measured plate heating. Approximately half of the overall plate heating power in detached plasma is due to absorbed radiation. By mapping the divertor heat flux before and during the PDD to flux coordinates, and comparing with a flux mapping of inserted bolometer and tangential TV data, they have verified that the radiated power is emitted from the same flux surfaces on which heat flux is reduced.

In this report we present measurements of the diverter heat flux in DIII-D for ELMing H-mode and radiative diverter conditions. In previous work we have examined heat flux profiles in lower single-null diverted plasmas and measured the scaling of the peak heat flux with plasma current and beam power. One problem with those results was our lack of good power accounting. This situation has been improved to better than 80--90% accountability with the installation of new bolometer arrays, and the operation of the entire complement of 5 Infrared (IR) TV cameras using the DAPS (Digitizing Automated Processing System) video processing system for rapid inter-shot data analysis. We also have expanded the scope of our measurements to include a wider variety of plasma shapes (e.g., double-null diverters (DND), long and short single-null diverters (SND), and inside-limited plasmas), as well as more diverse discharge conditions. Double-null discharges are of particular interest because that shape has proven to yield the highest confinement (VH-mode) and beta of all DIII-D plasmas, so any future diverter modifications for DIII-D will have to support DND operation. In addition, the proposed TPX tokamak is being designed for double-null operation, and information on the magnitude and distribution of divertermore » heat flux is needed to support the engineering effort on that project. So far, we have measured the DND power sharing at the target plates and made preliminary tests of heat flux reduction by gas injection.« less