Showing papers in "Nuclear Fusion in 2012"

Heuristic drift-based model of the power scrape-off width in low-gas-puff H-mode tokamaks

Abstract: A heuristic model for the plasma scrape-off width in low-gas-puff tokamak H-mode plasmas is introduced. Grad B and curv B drifts into the scrape-off layer (SOL) are balanced against near-sonic parallel flows out of the SOL, to the divertor plates. The overall particle flow pattern posited is a modification for open field lines of Pfirsch‐ Schl¨ uter flows to include order-unity sinks to the divertors. These assumptions result in an estimated SOL width of ∼2aρp/R. They also result in a first-principles calculation of the particle confinement time of H-mode plasmas, qualitatively consistent with experimental observations. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, derived above, with heat from the main plasma. The separatrix temperature is calculated based on a two-point model balancing power input to the SOL with Spitzer‐H¨ arm parallel thermal conduction losses to the divertor. This results in a heuristic closed-form prediction for the power scrape-off width that is in reasonable quantitative agreement both in absolute magnitude and in scaling with recent experimental data. Further work should include full numerical calculations, including all magnetic and electric drifts, as well as more thorough comparison with experimental data. (Some figures may appear in colour only in the online journal)

An integrated model for materials in a fusion power plant: transmutation, gas production, and helium embrittlement under neutron irradiation

Abstract: The high-energy, high-intensity neutron fluxes produced by the fusion plasma will have a significant life-limiting impact on reactor components in both experimental and commercial fusion devices. As well as producing defects, the neutrons bombarding the materials initiate nuclear reactions, leading to transmutation of the elemental atoms. Products of many of these reactions are gases, particularly helium, which can cause swelling and embrittlement of materials.This paper integrates several different computational techniques to produce a comprehensive picture of the response of materials to neutron irradiation, enabling the assessment of structural integrity of components in a fusion power plant. Neutron-transport calculations for a model of the next-step fusion device DEMO reveal the variation in exposure conditions in different components of the vessel, while inventory calculations quantify the associated implications for transmutation and gas production. The helium production rates are then used, in conjunction with a simple model for He-induced grain-boundary embrittlement based on electronic-structure density functional theory calculations, to estimate the timescales for susceptibility to grain-boundary failure in different fusion-relevant materials. There is wide variation in the predicted grain-boundary-failure lifetimes as a function of both microstructure and chemical composition, with some conservative predictions indicating much less than the required lifetime for components in a fusion power plant.

Overview of the physics and engineering design of NSTX upgrade

Abstract: The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5 s to 5–8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described.

Tungsten nano-tendril growth in the Alcator C-Mod divertor

Abstract: Growth of tungsten nano-tendrils (?fuzz?) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50?100?nm) and the depth of the layer (600???150?nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices.

Off-diagonal particle and toroidal momentum transport: a survey of experimental, theoretical and modelling aspects

Abstract: In tokamaks, turbulent particle and toroidal momentum transport are both characterized by the presence of off-diagonal contributions which play an essential role in establishing the profile shapes of the density and the toroidal rotation under most conditions. In this paper similarities and differences between the two turbulent transport channels are pointed out and, thereby, interesting physical aspects which connect the two channels are identified. The main contributions to off-diagonal particle and toroidal momentum transport are reviewed by means of a rather simplified description, which aims at providing, when possible, a direct connection between theoretical, modelling and experimental research.

Screening of resonant magnetic perturbations by flows in tokamaks

Abstract: The non-linear reduced four-field RMHD model in cylindrical geometry was extended to include plasma rotation, neoclassical poloidal viscosity and two fluid diamagnetic effects. Interaction of the static resonant magnetic perturbations (RMPs) with the rotating plasmas in tokamaks was studied. The self-consistent evolution of equilibrium electric field due to RMP penetration is taken into account in the model. It is demonstrated that in the pedestal region with steep pressure gradients, mean flows perpendicular to the magnetic field, which includes and electron diamagnetic components plays an essential role in RMP screening by plasma. Generally, the screening effect increases for lower resistivity, stronger rotation and smaller RMP amplitude. Strong screening of central islands was observed limiting RMP penetration to the narrow region near the separatrix. However, at certain plasma parameters and due to the non-linear evolution of the radial electric field produced by RMPs, the rotation can be compensated by electron diamagnetic rotation locally. In this case, RMPs can penetrate and form magnetic islands. Typical plasma parameters and RMPs spectra on DIII-D, JET and ITER were used in modelling examples presented in the paper.

The effect of progressively increasing lithium coatings on plasma discharge characteristics, transport, edge profiles and ELM stability in the National Spherical Torus Experiment

Abstract: Lithium wall coatings have been shown to reduce recycling, suppress edge-localized modes (ELMs), and improve energy confinement in the National Spherical Torus Experiment (NSTX). Here we document the effect of gradually increasing lithium wall coatings on the discharge characteristics, with the reference ELMy discharges obtained in boronized, i.e. non-lithiated conditions. We observed a continuous but not quite monotonic reduction in recycling and improvement in energy confinement, a gradual alteration of edge plasma profiles, and slowly increasing periods of ELM quiescence. The measured edge plasma profiles during the lithium-coating scan were simulated with the SOLPS code, which quantified the reduction in divertor recycling coefficient from ?98% to ?90%. The reduction in recycling and fuelling, coupled with a drop in the edge particle transport rate, reduced the average edge density profile gradient, and shifted it radially inwards from the separatrix location. In contrast, the edge electron temperature (Te) profile was unaffected in the H-mode pedestal steep gradient region within the last 5% of normalized poloidal flux, ?N ; however, the Te gradient became steeper at the top of the H-mode pedestal for 0.8?

Optimized tokamak power exhaust with double radiative feedback in ASDEX Upgrade

Abstract: A double radiative feedback technique has been developed on the ASDEX Upgrade tokamak for optimization of power exhaust with a standard vertical target divertor. The main chamber radiation is measured in real time by a subset of three foil bolometer channels and controlled by argon injection in the outer midplane. The target heat flux is in addition controlled by nitrogen injection in the divertor private flux region using either a thermoelectric sensor or the scaled divertor radiation obtained by a bolometer channel in the outer divertor. No negative interference of the two radiation controllers has been observed so far. The combination of main chamber and divertor radiative cooling extends the operational space of a standard divertor configuration towards high values of P/R. Pheat/R = 14 MW m−1 has been achieved so far with nitrogen seeding alone as well as with combined N + Ar injection, with the time-averaged divertor peak heat flux below 5 MW m−2. Good plasma performance can be maintained under these conditions, namely H98(y,2) = 1 and βN = 3.

L- to H-mode transitions at low density in ASDEX Upgrade

Abstract: The results from ASDEX Upgrade discharges dedicated specifically to the investigation of low-density L-to-H transitions are presented. The plasmas were heated by electron cyclotron resonance heating to achieve a separation of electron and ion heat channels. Under such conditions, the ratio of electron to ion temperature at the plasma edge increases with decreasing density at the L–H transition and can be as high as 3.5. Our results strongly support the essential role of the ion channel in the L–H transition, via the diamagnetic Er provided by the ion pressure gradient.

Saturation of deuterium retention in self-damaged tungsten exposed to high-flux plasmas

Abstract: Polycrystalline, annealed tungsten targets were bombarded with 12.3 MeV W4+ ions to various damage levels. Deuterium was implanted by high-flux plasmas in Pilot-PSI (>1024 m−2 s−1) at a surface temperature below 525 K. Deuterium retention has been studied by nuclear reaction analysis and by thermal desorption spectroscopy. We found that deuterium retention is strongly enhanced by the tungsten bombardment and that saturation occurs at a W4+ fluence of about 3 × 1017 m−2. The maximum deuterium concentration in the damaged region was measured to be 1.4 at.%. This is in accordance with other experiments that were carried out at much lower fluxes. We therefore conclude that the saturation behaviour and the maximum retention are not affected by the high fluxes used in our experiments.A simple geometric model is presented that assumes that the saturation solely originates in the tungsten irradiation and that explains it in terms of overlapping saturated volumes. The saturated volume per incident MeV ion amounts to 3 × 104 nm3. From our results, we are able to obtain an approximate value for the average occupation number of the vacancies.

The Chodura sheath for angles of a few degrees between the magnetic field and the surface of divertor targets and limiters

Abstract: To achieve low deposited power flux density to solid surfaces in magnetic fusion devices, very small values of α are required, where α is the angle between B and the surface tangent. For an oblique magnetic field, there exists in front of the solid surface a Chodura sheath (CS) (also known as the 'magnetic pre-sheath') of thickness several ρi, the ion Larmor radius. The standard assumption is that the CS is additional to the Debye sheath (DS) of thickness several λD, the Debye length. Simple fluid modelling for collisionless CS conditions gives the drop in normalized electrostatic potential across the CS as eΔCS/kTe = ln(sin α). For an electrically floating wall there is the separate constraint of ambipolar flow to the wall eΔfloating/kTe = 0.5 ln[(2πme/mi)(1 + Ti/Te)], where Δfloating = ΔCS + ΔDS. For the case of a deuterium plasma and Ti = Te, |eΔfloating/kTe| = 2.84. For α < 3.35°, |eΔCS/kTe| exceeds 2.84 which evidently implies that the DS ceases to exist for such values of α and the entire potential drop would then occur across the CS. New analysis of the CS provides solutions for a number of quantities of practical importance, which improve on the solutions presently in use in models and edge impurity codes. Compared with the latter, the results of the present analysis indicate that (i) the E-field directed towards the solid surface is stronger and (ii) the plasma density drops more rapidly approaching the solid surface. The effect of (i) is to increase the probability of prompt local deposition of sputtered particles, while (ii) has the opposite effect.

Role of singular layers in the plasma response to resonant magnetic perturbations

Abstract: The response of an H-mode plasma to magnetic perturbations that are resonant in the edge is evaluated using a fluid model. With two exceptions, the plasma rotation suppresses the formation of magnetic islands, holding their widths to less than a tenth of those predicted by the vacuum approximation. The two exceptions are at the foot of the pedestal, where the plasma becomes more resistive, and at the surface where the perpendicular component of the electron velocity reverses. The perturbations exert a force on the plasma so as to brake the perpendicular component of the electron rotation. In the pedestal, the corresponding Maxwell stress drives the radial electric field in such a way as to accelerate ion rotation. Despite the suppression of the islands, the perturbations give rise to particle fluxes caused by magnetic flutter, with a negligible contribution from E × B convection. In the pedestal, the fluxes are such as to reduce the density.

Control of neoclassical tearing modes

Abstract: Neoclassically driven tearing modes (NTMs) are a major problem for tokamaks operating in a conventional ELMy H-mode scenario. Depending on the mode numbers these pressure-driven perturbations cause a mild reduction in the maximum achievable βN = βt/(Ip/aBt) before the onset of the NTM, or can even lead to disruptions at a low edge safety factor, q95. A control of these types of modes in high βN plasmas is therefore of vital interest for magnetically confined fusion plasmas. The control consists of two major approaches, namely the control of the excitation of these modes and the removal, or at least mitigation, of these modes, once an excitation could not be avoided. For both routes examples will be given and the applicability of these approaches to ITER will be discussed.

Effect of externally applied resonant magnetic perturbations on resistive tearing modes

Abstract: Static resonant magnetic perturbations (RMPs) generated by saddle coil current have been applied in J-TEXT tokamak experiments in order to study their effects on tearing mode instabilities. With increasing RMP amplitude in time during the discharge, the mode stabilization is first observed, but a large locked mode follows if the RMP amplitude is increased to a very large value, indicating that the RMP amplitude is important in determining the plasma response and the tearing mode behaviour. By careful adjustment of the RMP amplitude, the (partial) stabilization of the m/n?=?2/1 tearing mode by RMPs of moderate amplitude has been achieved without causing mode locking (m and n are the poloidal and toroidal mode numbers). To compare with experimental results, nonlinear numerical modelling based on reduced MHD equations has been carried out. With experimental parameters as input, both the mode locking and mode stabilization by RMPs are also obtained from numerical modelling. Further calculations have been carried out to study the plasma parameters affecting the mode stabilization by RMPs, including the plasma rotation frequency, viscosity, Alfv?n velocity and the RMP amplitude. It is found that the suppression of the tearing mode by RMPs of moderate amplitude is possible for a sufficiently high ratio of plasma rotation velocity to the Alfv?n speed. A larger plasma viscosity enhances the mode stabilization.

ELM mitigation by supersonic molecular beam injection into the H-mode pedestal in the HL-2A tokamak

Abstract: Density profiles in the pedestal region (H-mode) are measured in HL-2A and the characteristics of the density pedestal are described. Cold particle deposition by supersonic molecular beam injection (SMBI) within the pedestal is verified. Edge-localized mode (ELM) mitigation by SMBI into the H-mode pedestal is demonstrated and the relevant physics is elucidated. The sensitivity of the effect to SMBI pressure and duration is studied. Following SMBI, the ELM frequency increases and the ELM amplitude decreases for a finite duration. Increases in ELM frequency of are achieved. This experiment argues that the ELM mitigation results from an increase in higher frequency fluctuations and transport events in the pedestal, which are caused by SMBI. These inhibit the occurrence of large transport events which span the entire pedestal width. The observed change in the density pedestal profiles and edge particle flux spectrum with and without SMBI supports this interpretation. An analysis of the experiment and a model shows that ELMs can be mitigated by SMBI with shallow particle penetration into the pedestal.

Integrated real-time control of MHD instabilities using multi-beam ECRH/ECCD systems on TCV

Abstract: Simultaneous real-time control of multiple MHD instabilities is experimentally demonstrated in the TCV tokamak. Multiple sources of EC heating and current drive, injected through real-time controlled launchers, are used to stabilize 3/2 and 2/1 neoclassical tearing modes (NTMs) rapidly after their appearance. Control of the sawtooth instability using a new sawtooth-pacing technique is demonstrated, providing precise control of the time of appearance of the sawtooth crash. Efficient NTM preemption can then be performed by applying pulsed power on the mode rational surface at the time of the seed-island generating sawtooth crash. These three elements are combined into one integrated control system which can simultaneously control the sawtooth period, preempt the formation of NTMs and suppress these if they appear.

Linear properties of reversed shear Alfvén eigenmodes in the DIII-D tokamak

Abstract: Linear properties of the reverse shear Alfven eigenmode (RSAE) in a well-diagnosed DIII-D tokamak experiment (discharge #142111) are studied in gyrokinetic particle simulations. Simulations find that a weakly damped RSAE exists due to toroidal coupling and other geometric effects. The mode is driven unstable by density gradients of fast ions from neutral beam injection. Various damping and driving mechanisms are identified and measured in the simulations. Accurate damping and growth rate calculation requires a non-perturbative, fully self-consistent simulation to calculate the true mode structure. The mode structure has no up–down symmetry mainly due to the radial symmetry breaking by the density gradients of the fast ions, as measured in the experiment by electron cyclotron emission imaging. The RSAE frequency up-sweeping and the mode transition from RSAE to TAE (toroidal Alfven eigenmode) are in good agreement with the experimental results when the values of the minimum safety factor are scanned in gyrokinetic simulations.

Quantitative comparison of experimental impurity transport with nonlinear gyrokinetic simulation in an Alcator C-Mod L-mode plasma

Abstract: Nonlinear gyrokinetic simulations of impurity transport are compared to experimental impurity transport for the first time. The GYRO code (Candy and Waltz 2003 J. Comput. Phys. 186 545) was used to perform global, nonlinear gyrokinetic simulations of impurity transport for a standard Alcator C-Mod, L-mode discharge. The laser blow-off technique was combined with soft x-ray measurements of a single charge state of calcium to provide time-evolving profiles of this non-intrinsic, non-recycling impurity over a radial range of 0.0 ≤ r/a ≤ 0.6. Experimental transport coefficient profiles and their uncertainties were extracted from the measurements using the impurity transport code STRAHL and rigorous Monte Carlo error analysis. To best assess the agreement of gyrokinetic simulations with the experimental profiles, the sensitivity of the GYRO predicted impurity transport to a wide range of turbulence-relevant plasma parameters was investigated. A direct comparison of nonlinear gyrokinetic simulation and experiment is presented with an in depth discussion of error sources and a new data analysis methodology.

Integration of a radiative divertor for heat load control into JET high triangularity ELMy H-mode plasmas

Abstract: Experiments on JET with a carbon-fibre composite wall have explored the reduction of steady-state power load in an ELMy H-mode scenario at high Greenwald fraction similar to 0.8, constant power and close to the L to H transition. This paper reports a systematic study of power load reduction due to the effect of fuelling in combination with seeding over a wide range of pedestal density ((4-8) x 10(19) m(-3)) with detailed documentation of divertor, pedestal and main plasma conditions, as well as a comparative study of two extrinsic impurity nitrogen and neon. It also reports the impact of steady-state power load reduction on the overall plasma behaviour, as well as possible control parameters to increase fuel purity. Conditions from attached to fully detached divertor were obtained during this study. These experiments provide reference plasmas for comparison with a future JET Be first wall and an all W divertor where the power load reduction is mandatory for operation.

Tomography of fast-ion velocity-space distributions from synthetic CTS and FIDA measurements

Abstract: We compute tomographies of 2D fast-ion velocity distribution functions from synthetic collective Thomson scattering (CTS) and fast-ion Dα (FIDA) 1D measurements using a new reconstruction prescription. Contradicting conventional wisdom we demonstrate that one single 1D CTS or FIDA view suffices to compute accurate tomographies of arbitrary 2D functions under idealized conditions. Under simulated experimental conditions, single-view tomographies do not resemble the original fast-ion velocity distribution functions but nevertheless show their coarsest features. For CTS or FIDA systems with many simultaneous views on the same measurement volume, the resemblance improves with the number of available views, even if the resolution in each view is varied inversely proportional to the number of views, so that the total number of measurements in all views is the same. With a realistic four-view system, tomographies of a beam ion velocity distribution function at ASDEX Upgrade reproduce the general shape of the function and the location of the maxima at full and half injection energy of the beam ions. By applying our method to real many-view CTS or FIDA measurements, one could determine tomographies of 2D fast-ion velocity distribution functions experimentally.

Divertor tungsten tile melting and its effect on core plasma performance

Abstract: For the 2007 and 2008 run campaigns, Alcator C-Mod operated with a full toroidal row of tungsten tiles in the high heat flux region of the outer divertor; tungsten levels in the core plasma were below measurement limits. An accidental creation of a tungsten leading edge in the 2009 campaign led to this study of a melting tungsten source: H-mode operation with strike point in the region of the melting tile was immediately impossible due to some fraction of tungsten droplets reaching the main plasma. Approximately 15 g of tungsten was lost from the tile over ~100 discharges. Less than 1% of the evaporated tungsten was found re-deposited on surfaces, the rest is assumed to have become dust. The strong discharge variability of the tungsten reaching the core implies that the melt layer topology is always varying. There is no evidence of healing of the surface with repeated melting. Forces on the melted tungsten tend to lead to prominences that extend further into the plasma. A discussion of the implications of melting a divertor tungsten monoblock on the ITER plasma is presented.

Exploration of the equilibrium operating space for NSTX-Upgrade

Abstract: This paper explores a range of high-performance equilibrium scenarios achievable with neutral beam heating in the NSTX-Upgrade device (Menard J.E. 2012 Nucl. Fusion 52 083015). NSTX-Upgrade is a substantial upgrade to the existing NSTX device (Ono M. et al 2000 Nucl. Fusion 40 557), with significantly higher toroidal field and solenoid capabilities, and three additional neutral beam sources with significantly larger current-drive efficiency. Equilibria are computed with free-boundary TRANSP, allowing a self-consistent calculation of the non-inductive current-drive sources, the plasma equilibrium and poloidal-field coil currents, using the realistic device geometry. The thermal profiles are taken from a variety of existing NSTX discharges, and different assumptions for the thermal confinement scalings are utilized. The no-wall and ideal-wall n = 1 stability limits are computed with the DCON code. The central and minimum safety factors are quite sensitive to many parameters: they generally increase with large outer plasma-wall gaps and higher density, but can have either trend with the confinement enhancement factor. In scenarios with strong central beam current drive, the inclusion of non-classical fast-ion diffusion raises qmin, decreases the pressure peaking, and generally improves the global stability, at the expense of a reduction in the non-inductive current-drive fraction; cases with less beam current drive are largely insensitive to additional fast-ion diffusion. The non-inductive current level is quite sensitive to the underlying confinement and profile assumptions. For instance, for BT = 1.0 T and Pinj = 12.6 MW, the non-inductive current level varies from 875 kA with ITER-98y,2 thermal confinement scaling and narrow thermal profiles to 1325 kA for an ST specific scaling expression and broad profiles. Scenarios are presented which can be sustained for 8–10 s, or (20–30) τCR, at βN = 3.8–4.5. The value of qmin can be controlled at either fixed non-inductive fraction of 100% or fixed plasma current, by varying which beam sources are used, opening the possibility for feedback control of the current profile. In terms of quantities like collisionality, neutron emission, non-inductive fraction, or stored energy, these scenarios represent a significant performance extension compared with NSTX and other present spherical torii.

Particle and power deposition on divertor targets in EAST H-mode plasmas

Abstract: The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons were made between the H-mode plasmas with lower hybrid current drive (LHCD) and those with combined ion cyclotron resonance heating (ICRH). The particle and heat flux profiles between and during ELMs were obtained from Langmuir triple-probe arrays embedded in the divertor target plates. And isolated ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM-free period. It was demonstrated that ELM-induced radial transport predominantly originated from the low-field side region, in good agreement with the ballooning-like transport model and experimental results of other tokamaks. ELMs significantly enhanced the divertor particle and heat fluxes, without significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle recycling at the divertor target, hence facilitating long-pulse H-mode operation. The particle and heat flux profiles during ELMs appeared to exhibit multiple peak structures, and were analysed in terms of the behaviour of ELM filaments and the flux tubes induced by modified magnetic topology during ELMs.

Measurement of neoclassically predicted edge current density at ASDEX Upgrade

Abstract: Experimental confirmation of neoclassically predicted edge current density in an ELMy H-mode plasma is presented. Current density analysis using the CLISTE equilibrium code is outlined and the rationale for accuracy of the reconstructions is explained. Sample profiles and time traces from analysis of data at ASDEX Upgrade are presented. A high time resolution is possible due to the use of an ELM-synchronization technique. Additionally, the flux-surface-averaged current density is calculated using a neoclassical approach. Results from these two separate methods are then compared and are found to validate the theoretical formula. Finally, several discharges are compared as part of a fuelling study, showing that the size and width of the edge current density peak at the low-field side can be explained by the electron density and temperature drives and their respective collisionality modifications.

Contribution of $\vec {E}\times \vec {B}$ drifts and parallel currents to divertor asymmetries

Abstract: A systematical study of drift and parallel current effects is reported based on the analysis of the simulations by the B2SOLPS5.2 transport code. It is demonstrated that divertor asymmetry is caused or amplified by the poloidal drift and parallel thermal current.

Gyrokinetic simulation model for kinetic magnetohydrodynamic processes in magnetized plasmas

Abstract: A nonlinear gyrokinetic simulation model incorporating equilibrium current has been formulated for studying kinetic magnetohydrodynamic processes in magnetized plasmas. This complete formulation enables gyrokinetic simulation of both pressure-gradient-driven and current-driven instabilities as well as their nonlinear interactions in multi-scale simulations. The gyrokinetic simulation model recovers the ideal magnetohydrodynamic theory in the linear long wavelength regime including ideal and kinetic ballooning modes, kink modes and shear Alfven waves. The implementation of this model in the global gyrokinetic particle code has been verified for the simulation of the effects of equilibrium current on the reversed shear Alfven eigenmode in tokamaks.

First mirrors in ITER: material choice and deposition prevention/cleaning techniques

Abstract: We present here our recent results on the development and testing of the first mirrors for the divertor Thomson scattering diagnostics in ITER. The Thomson scattering system is based on several large-scale (tens of centimetres) mirrors that will be located in an area with extremely high (3?10%) concentration of contaminants (mainly hydrocarbons) and our main concern is to prevent deposition-induced loss of mirror reflectivity in the spectral range 1000?1064?nm. The suggested design of the mirrors?a high-reflective metal layer on a Si substrate with an oxide coating?combines highly stable optical characteristics under deposition-dominated conditions with excellent mechanical properties. For the mirror layer materials we consider Ag and Al allowing the possibility of sharing the Thomson scattering mirror collecting system with a laser-induced fluorescence system operating in the visible range. Neutron tests of the mirrors of this design are presented along with numerical simulation of radiation damage and transmutation of mirror materials. To provide active protection of the large-scale mirrors we use a number of deposition-mitigating techniques simultaneously. Two main techniques among them, plasma treatment and blowing-out, are considered in detail. The plasma conditions appropriate for mirror cleaning are determined from experiments using plasma-induced erosion/deposition in a CH4/H2 gas mixture. We also report data on the numerical simulation of plasma parameters of a capacitively-coupled discharge calculated using a commercial CFD-ACE code. A comparison of these data with the results for mirror testing under deuterium ion bombardment illustrates the possibility of using the capacitively-coupled discharge for in situ non-destructive deposition mitigation/cleaning.

Deuterium retention in tungsten and tungsten-tantalum alloys exposed to high-flux deuterium plasmas

Abstract: A direct comparison of deuterium retention in samples of tungsten and two grades of tungsten-tantalum alloys-W-1% Ta and W-5% Ta, exposed to deuterium plasmas (ion flux similar to 10(24) m(-2) s(-1), ion energy at the biased target similar to 50 eV) at the plasma generator Pilot-PSI was performed using thermal desorption spectroscopy (TDS). No systematic difference in terms of total retention in tungsten and tungsten-tantalum was identified. The measured retention value for each grade did not deviate by more than 24% from the value averaged over the three grades exposed to the same conditions. No additional desorption peaks appeared in the TDS spectra of the W-Ta samples as compared with the W target, indicating that no additional kinds of traps are introduced by the alloying of W with Ta. In the course of the experiment the same samples were exposed to the same plasma conditions several times, and it is demonstrated that samples with the history of prior exposures yield an increase in deuterium retention of up to 130% under the investigated conditions compared with the samples that were not exposed before. We consider this as evidence that exposure of the considered materials to ions with energy below the displacement threshold generates additional traps for deuterium. The positions of the release peaks caused by these traps are similar for W and W-Ta, which indicates that the corresponding traps are of the same kind.

Poloidal asymmetry of parallel rotation measured in ASDEX Upgrade

Abstract: The parallel flows in the H-mode edge of ASDEX Upgrade are investigated. Beam-based charge-exchange recombination spectroscopy (CXRS) provides the toroidal and poloidal impurity flow velocities at the outboard midplane, while a deuterium-puff based CXRS measurement provides the toroidal impurity flow velocities at the inboard midplane. In order to more easily compare these measurements to fundamental boundary conditions, a basic overview of flows on a flux surface is presented. The boundary conditions are given by the continuity equation and mean that the flow velocities on a flux surface must have a specific structure in order to provide zero divergence. At first, poloidal impurity density asymmetries and radial transport are neglected. Inside of the pedestal-top of the electron density profile the measurements agree with the postulated flow structure, while they do not agree at the pedestal itself. Here, an extension of the theoretical scheme, which allows for a poloidal impurity density asymmetry, suggests that the measured flow velocities could be explained by an excess impurity density at the inboard midplane. In detail, the inboard impurity density is postulated to be at the separatrix up to a factor of 6.5 higher than impurity density at the outboard midplane. Near the pedestal-top of the electron density, this asymmetry disappears. Radial transport is considered as an explanation for that asymmetry. A conclusive disentanglement of the driving mechanisms requires further investigation.

Progress in GYRO validation studies of DIII-D H-mode plasmas

Abstract: The need for a validated predictive capability of turbulent transport in ITER is now widely recognized. However, to date most validation studies of nonlinear codes such as GYRO (Candy and Waltz 2003 J. Comput. Phys. 186 545) have focused upon low power L-mode discharges, which have significant differences in key dimensionless parameters such as ρ* = ρs/a from more ITER-relevant H-mode discharges. In order to begin addressing this gap, comparisons of the turbulent transport and fluctuations predicted by nonlinear GYRO simulations for a number of DIII-D (Luxon 2002 Nucl. Fusion 42 614) H-mode discharges to power balance analyses and experimental measurements are presented. The results of two H-mode studies are presented in this paper, this first of which investigates the importance of nonlocality at typical DIII-D H-mode ρ* values. Electrostatic global GYRO simulations of H-mode discharges at low and high rotation are shown to predict turbulence and transport levels lower than corresponding local simulations, and which are consistent with or slightly above experimental measurements and power balance analyses, even at 'near-edge' radii where gyrofluid and gyrokinetic models systematically underpredict turbulence and transport levels. The second study addresses the stabilizing effect of a significant density of energetic particles on turbulent transport. The results of local GYRO simulations of low-density QH-mode plasmas are presented, which model the fast beam ion population as a separate, dynamic ion species. The simulations show a significant reduction of transport with this fast ion treatment, which helps to understand previously reported results (Holland et al 2011 Phys. Plasmas 18 056113) in which GYRO simulations without this treatment significantly overpredicted (by a factor of 10 or more) power balance calculations. These results are contrasted with simulations of a high-density, low fast ion fraction QH-mode discharge, which predict transport levels consistent with power balance, regardless of the fast ion treatment.