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Osteoporotic vertebral fractures signify an increased risk of future fractures and mortality and can manifest the diagnosis of osteoporosis. We investigated the prevalence of vertebral fractures in routine CT of patients with long-term hospital records. Three out of ten patients showed osteoporotic vertebral fractures (VFs) corresponding to the highest rates reported in European population-based studies.VFs are a common manifestation of osteoporosis, which influences future fracture risk. Their epidemiology has been investigated in population-based studies. However, few studies report the prevalence of osteoporotic VF in patients seen in clinical routine and include all common fracture levels of the thoracolumbar spine. The purpose of this study was to investigate the prevalence of osteoporotic VF in patients with CT scans and long-term hospital records and identify clinical factors associated with prevalent VFs.All patients aged 45 years and older with a CT scan and prior hospital record of at least 5 years that were seen in the study period between September 2008 and May 2017 were reviewed. Imaging requirements were a CT scan with sagittal reformations including at least T6-L4. Patients with multiple myeloma were excluded. Fracture reading was performed using the Genant semi-quantitative method. Medical notes were reviewed for established diagnoses of osteoporosis and clinical information. Clinical factors (e.g. drug intake, chemotherapy, and mobility level) associated with prevalent VF were identified in logistic regression.The study population consisted of 718 patients (228 women and 490 men; mean age 69.3 ± 10.1 years) with mainly cancer staging and angiography CT imaging. The overall prevalence of VFs was 30.5%, with non-significantly more men showing a fracture (32.5%) compared to women (26.3%; p 0.05). Intake of metamizole for ≥ 3 months was significantly associated with a prevalent VF. Medical records did not include information about bone health in 90% of all patients. CT reports did mention a VF in only 24.7% of patients with a prevalent VF on CT review.Approximately 30% of elderly patients with CT imaging and long-term hospital records showed VFs. Only one-quarter of these patients had VFs mentioned in CT reports. Osteoporosis management could be improved by consequent reporting of VFs in CT, opportunistic bone density measurements, and early involvement of fracture liaison services.

A Fast Fourier Transform (FFT) based method has been developed, which improves the frequency response of the Motional Stark Effect (MSE) system by about a factor of 10 over the conventional analog lock-in method. The method uses fits to rigorously derived analytic expressions for the FFT spectral components of the MSE signal to accurately obtain the amplitudes and phases of the 2f1 and 2f2 photo-elastic modulator (PEM) frequencies that encode the polarization angle. Since no frequency filtering is used in the FFT method, the frequency response is limited by fundamental measurement properties: the frequency response of the detector, photon statistics, sample rate, and the ability to resolve the spectral components. In contrast, the frequency response of the analog lock-in is limited by a low pass filter with a cutoff of around 500 Hz. In the case of the DIII-D MSE system, the output of the photo-multiplier tube detector was sampled at 500 kHz and FFTs with as few as 100 points were used to obtain the amplitudes of the 2f1 and 2f2 PEM frequency components. This corresponds to a frequency response of 5 kHz, about ten times faster than the analog lock-in amplifier system. Details of the FFT method will be presented and compared to those of the analog lock-in system.

The bone marrow proton density fat fraction (PDFF) assessed with MRI enables the differentiation between osteoporotic/osteopenic patients with and without vertebral fractures. Therefore, PDFF may be a potentially useful biomarker for bone fragility assessment.To evaluate whether magnetic resonance imaging (MRI)-based proton density fat fraction (PDFF) of vertebral bone marrow can differentiate between osteoporotic/osteopenic patients with and without vertebral fractures.Of the 52 study patients, 32 presented with vertebral fractures of the lumbar spine (66.4 ± 14.4 years, 62.5% women; acute low-energy osteoporotic/osteopenic vertebral fractures, N = 25; acute high-energy traumatic vertebral fractures, N = 7). These patients were frequency matched for age and sex to patients without vertebral fractures (N = 20, 69.3 ± 10.1 years, 70.0% women). Trabecular bone mineral density (BMD) values were derived from quantitative computed tomography. Chemical shift encoding-based water-fat MRI of the lumbar spine was performed, and PDFF maps were calculated. Associations between fracture status and PDFF were assessed using multivariable linear regression models.Over all patients, mean PDFF and trabecular BMD correlated significantly (r = - 0.51, P 0.001). In the osteoporotic/osteopenic group, those patients with osteoporotic/osteopenic fractures had a significantly higher PDFF than those without osteoporotic fractures after adjusting for age, sex, weight, height, and trabecular BMD (adjusted mean difference [95% confidence interval], 20.8% [10.4%, 30.7%]; P 0.001), although trabecular BMD values showed no significant difference between the subgroups (P = 0.63). For the differentiation of patients with and without vertebral fractures in the osteoporotic/osteopenic subgroup using mean PDFF, an area under the receiver operating characteristic (ROC) curve (AUC) of 0.88 (P = 0.006) was assessed. When evaluating all patients with vertebral fractures, those with high-energy traumatic fractures had a significantly lower PDFF than those with low-energy osteoporotic/osteopenic vertebral fractures (P 0.001).MR-based PDFF enables the differentiation between osteoporotic/osteopenic patients with and without vertebral fractures, suggesting the use of PDFF as a potential biomarker for bone fragility.

Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P) ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core. PSI-22 keywords: DIII-D, Divertor geometry, Gas injection and fueling, Impurity sources, Power deposition

The effects of drifts and associated flows and currents on the width of the parallel heat flux channel (λq) in the tokamak scrape-off layer (SOL) are analyzed using the SOLPS-ITER 2D fluid transport code. Motivation is supplied by Goldston’s heuristic drift (HD) model for λq, which yields the same approximately inverse poloidal magnetic field dependence seen in multi-machine regression. The analysis, focusing on a DIII-D H-mode discharge, reveals HD-like features, including comparable density and temperature fall-off lengths in the SOL, and up-down ion pressure asymmetry that allows net cross-separatrix ion magnetic drift flux to exceed net anomalous ion flux. In experimentally relevant high-recycling cases, scans of both toroidal and poloidal magnetic field (Btor and Bpol) are conducted, showing minimal λq dependence on either component of the field. Insensitivity to Btor is expected, and suggests that SOLPS-ITER is effectively capturing some aspects of HD physics. Absence of λq dependence on Bpol, however, is inconsistent with both the HD model and experimental results. The inconsistency is attributed to strong variation in the parallel Mach number, which violates one of the premises of the HD model. Keywords: Power scrape-off width, SOLPS-ITER, Drift effects, SOL flows

Detailed measurements of the ne, Te, and Ti profiles in the vicinity of the separatrix of ELMing H-mode discharges have been used to examine plasma stability at the extreme edge of the plasma and assess stability dependent models of the heat flux width. The results are strongly contrary to the critical gradient model, which posits that a ballooning instability determines a gradient scale length related to the heat flux width. The results of this analysis are not sensitive to the choice of location to evaluate stability. Significantly, it is also found that the results are completely consistent with the heuristic drift model for the heat flux width. Here the edge pressure gradient scales with plasma density and is proportional to the pressure gradient inferred from the equilibrium in accordance with the predictions of that theory.

As High-Z materials will likely be used as plasma-facing components (PFCs) in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13 C deposition resulting from local 13 CH 4 injection also provides information on radial transport due to E × B drifts and cross field diffusion. Finally, D 2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

The contributions of deuterium molecular emission to the total deuterium radiation was assessed in DIII-D ohmically-confined plasmas in high-recycling divertor conditions. Radial profiles of the deuterium Ly-α line intensity across the low-field side divertor leg were obtained with the recently installed divertor Survey Poor Resolution, Extended Spectrometer [1]. A high-resolution spectrometer was used to measure the poloidal profiles of the deuterium Balmer-α and the deuterium Fulcher-α band intensity in the visible wavelength range. The scrape-off layer plasma and neutral distributions were simulated using the edge fluid EDGE2D-EIRENE [2], and the numerical solutions constrained utilizing Thomson scattering and Langmuir probe measurements at the low-field side midplane and the divertor target plate. The studies show that for these conditions molecular emission plays a negligible role in the total radiative power balance of the low-field side divertor, but molecular processes are important when evaluating deuterium Balmer-α line intensity for code-experiment validation. Keywords: Deuterium atomic and molecular emission, Lyman-Werner and Fulcher band, DIII-D, Divertor edge fluid simulations, EDGE2D-EIRENE, EIRENE

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower $n_{e}$ , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative $D_{2}$ -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative $D_{2}$ -seeded SF divertor at $P_{\text {SOL}}=3-4$ MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected $P_{\text {SOL}} \simeq 9$ MW case. The radiative SF divertor with carbon impurity provides a wider $n_{e}$ operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar $q_{\text {peak}}$ reduction factors (see standard divertor).

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (~30 kPa) and stored energy (3.2 MJ) with H 98y2 ≈ 1.6–2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric independent of current between q 95 = 2.8–3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses ≈0 injected torque and the operating space is more ITER-relevant. Finally, the high- (⩽3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving ~40% divertor heat flux reduction using either argon or neon with P tot up to 15 MW.

The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, Te ⩽ 2 eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D 2 D 2 seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( I p =1.2 I p = 1.2 MA, P NBI =4–5 P NBI = 4 – 5 MW, and B×∇B B × ∇ B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary B p B p null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. Keywords: SOL width, Radio-frequency heating, Neutral beam heating, Low collisionality

A cross machine analysis of heat flux profile measurements has produced a particularly simple scaling law for the scrape-off-layer (SOL) width, going as B p , mp e with e ∼ −1.0 where B p , mp is the poloidal magnetic field at the outer midplane separatrix. This result is in very good agreement with a model based on drift-induced transport. We are also developing a model that extends the kinetic ballooning mode pedestal paradigm to the separatrix and into the SOL. We have calculated the critical pressure gradient at the separatrix using the BALOO code and found that it scales proportionately with the measured pressure gradient. Using an improved high rep-rate and higher edge resolution Thomson scattering system on DIII-D, it is now possible to make detailed comparisons of SOL profile characteristics with the heat flux profile. We find that a two-point flux-limited model is in better agreement with measurements than a conduction-limited model.

Spectrometer measurements and filter upgrades to a motional Stark effect polarimeter measuring the outer half-radius of the DIII-D tokamak helped to identify asymmetries in the polarization angle of Stark-split emission. The measured polarization angle of the π components differs and is not orthogonal to the σ component. These differences persist over a range of densities and with low levels of background light. It is suggested that the difference in the polarization angle between components is from a change in the ellipticity of the emitted light across the Stark components coupled with imperfect polarization preservation from an in-vessel mirror.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

Fast visible imaging of the lower divertor from above is used to study the structure and dynamics of lobes induced by resonant magnetic perturbations (RMPs) in Edge-Localized Mode (ELM) suppression experiments in DIII-D. The best compromise between the amount of light and sharp imaging was obtained using emission at 601 nm from Fulcher band molecular deuterium. Multiple spatially resolved peaks in the D2 emission, taken as a proxy for the particle flux, are readily resolved during RMPs, in contrast to the heat flux measured by infrared cameras, which shows little spatial structure in ITER-like conditions. The 25 mm objective lens provides high spatial resolution (2–4 mm/pixel) from the centerpost to the outer shelf over 40° toroidally that overlaps the field of view of the IRTV that measures the divertor heat flux, allowing direct comparison in non-axisymmetric discharges. The image is coupled to a Phantom 7.3 camera using a Schott wound fiber bundle, providing high temporal resolution that allows the lobe dynamics to be resolved between ELMs and across ELM suppression onset. These measurements are used to study the heat and particle flux in 3D magnetic fields and to validate models for the plasma response to RMPs.

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness ...

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5× reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

Recent experiments on the DIII-D tokamak have examined the effect of particle transport on the electron cyclotron current drive (ECCD) profile using measurements of the magnetic field pitch angles by motional Stark effect polarimetry. While previous ECCD studies on DIII-D did not observe any clear effects of transport, these new experiments at high ECCD power, low density, and radiation temperatures above 20 keV clearly demonstrate that the ECCD profile can be reduced and broadened compared to the Fokker-Planck code CQL3D predictions assuming no radial transport. A diffusion coefficient of [approximate]0.4 m2 /s is required in CQL3D to reproduce the experimental ECCD profile at high relative power densities, while smaller diffusion coefficients are needed at low relative power densities. This level of transport is comparable to the effective particle transport rate needed to maintain the density profile but an order of magnitude less than the electron thermal diffusivity. While radial transport of...

Momentum confinement was investigated on DIII-D as a function of applied neutral beam torque at constant normalized beta βN, by varying the mix of co (parallel to the plasma current) and counter neutral beams. Under balanced neutral beam injection (i.e. zero total torque to the plasma), the plasma maintains a significant rotation in the co-direction. This 'intrinsic' rotation can be modeled as being due to an offset in the applied torque (i.e. an 'anomalous torque'). This anomalous torque appears to have a magnitude comparable to one co neutral beam source. The presence of such an anomalous torque source must be taken into account to obtain meaningful quantities describing momentum transport, such as the global momentum confinement time and local diffusivities.Studies of the mechanical angular momentum in ELMing H-mode plasmas with elevated qmin show that the momentum confinement time improves as the torque is reduced. In hybrid plasmas, the opposite effect is observed, namely that momentum confinement improves at high torque/rotation. GLF23 modeling suggests that the role of E × B shearing is quite different between the two plasmas, which may help to explain the different dependence of the momentum confinement on torque.

The response of plasma parameters and broad wavenumber turbulence (1–39 cm−1, kρs = 0.1–10, relevant to ion temperature gradient, trapped electron mode and electron temperature gradient mode turbulence, here ρs = ion gyroradius) to auxiliary electron cyclotron heating (ECH) is reported on. One fluid thermal fluxes and diffusivities increase appreciably with ECH. Significant changes to the density fluctuations over the full range of measured wavenumbers are observed, with an increase for lower wavenumbers and a more spatially complicated response at high k. Spatially resolved high k measurements (k = 39 cm−1, kρs = 4–10) show a varying response to ECH, with decreasing at r/a = 0.35 and increasing at r/a = 0.6 and 1. These variations were found to have a positive correlation with ∇Te evaluated at nearby locations, consistent with a ∇Te drive. Comparison of the changes in high k fluctuation levels with linear gyrokinetic growth rates show qualitative agreement at the innermost location, r/a = 0.35 and disagreement at r/a = 0.6.

Alternatives to the usual picture of advanced tokamak (AT) discharges are those that form when anomalous thermal conductivity and/or resistivity alter plasma current and pressure profiles to achieve stationary characteristics through self-organizing mechanisms where a measure of desired AT features is maintained without external current-profile control. Regimes exhibiting these characteristics are those where the safety factor (q) evolves to a stationary profile with the on-axis and minimum q ∼ 1. Operating scenarios with fusion performance exceeding H-mode at the same plasma current and where the inductively driven current density achieves a stationary configuration with either small or nonexisting sawteeth should enhance the performance of ITER and future burning plasma experiments. We present simulation results of anomalous current-profile formation and evolution using theory-based hyper-resistive models. These simulations are stimulated by experimental observations with which we compare and contrast the simulated evolution. We find that the hyper-resistivity is sufficiently strong to modify the current-profile evolution to achieve conditions consistent with experimental observations. Modelling these anomalous effects is important for developing a capability to scale current experiments to future burning plasmas.

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration (D.D. Ryutov, Phys. Plasmas, 14, 064502, 2007) may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 – 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. The radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).

Recent upgrades to many of the diagnostic systems on DIII-D (Luxon J.L. 2002 Nucl. Fusion 42 614) such as the CO2 interferometer, far-infrared scattering, beam-emission spectroscopy (BES), and quadrature reflectometer have significantly extended their capabilities and made possible the experimental study of Alfven eigenmodes (AEs) through observation of the AE induced density perturbation. Measurements have revealed the presence of several different classes of AEs in DIII-D discharges including the toroidal Alfven eigenmode (TAE), reverse shear AE (RSAE or Alfven cascade) and ellipticity induced Alfven eigenmode. Based on a simple model for the RSAE frequency, a sensitive diagnostic for the evolution of the minimum magnetic safety factor (qmin) is presented and results are compared with motional Stark effect (MSE) measurements. Strong localization of high toroidal mode number RSAEs to regions near the minimum of the magnetic safety factor is exhibited on the CO2 interferometer and BES measurements. Based on this observation, a method for providing constraints on the radial location of qmin is demonstrated and a favourable comparison to MSE measurements is made. Detailed measurements of TAEs using a new all-digital large bandwidth two-colour CO2 interferometer system show a strong asymmetry between vertical and radial viewing interferometer chords confirming previously reported results. Additionally, effects related to line-integrated observations are clearly illustrated by comparison to local BES measurements and potential issues related to this are discussed.

Active feedback control for regulation of the safety factor (q) profile at the start of the high stored energy phase of an advanced tokamak discharge has been demonstrated in the DIII-D tokamak. The time evolution of the on-axis or minimum value of q is controlled during and just following the period of ramp-up of the plasma current using electron heating to modify the rate of relaxation of the current profile. In L-mode and H-mode discharges, feedback control of q is effective with the appropriate choice of either off-axis electron cyclotron heating or neutral beam heating as the actuator. The q profile is calculated in real time from a complete equilibrium reconstruction fitted to external magnetic field and flux measurements and internal poloidal field measurements from the motional Stark effect diagnostic.

The effect of plasma shape on sawtooth oscillations in the DIII-D tokamak plasmas is investigated by comparing discharges with cross-sections shaped like a bean and an oval. The two shapes are designed so that the Mercier instability threshold is reached when the axial safety factor is below unity for the bean and above unity for the oval cross-sections. This allows the role of interchange modes to be differentiated from that of the kink-tearing mode. The differences in the nature of the sawtooth oscillations in the bean and oval discharges are found to be determined primarily by extreme differences in the electron heat transport during the reheat. In both cases, the axial safety factor is found to be near unity following the crash. (Some figures in this article are in colour only in the electronic version)

Active control of plasma profiles is an essential requirement for operating within plasma stability limits, for steady-state operation and for optimization of the plasma performance. In DIII-D, plasma profiles have been actively controlled using various actuators in the following manner: (a) real time closed loop control of the q profile evolution using electron cyclotron heating and neutral beam injection as actuators; (b) active control of the density and pressure profiles in quiescent H-mode and quiescent double barrier plasmas using electron cyclotron current drive (ECCD) and pellet injection; (c) active control of the edge profiles to suppress edge localized modes using resonant magnetic perturbation with toroidal mode number n = 3, (d) real time control of the current density profile to suppress neoclassical tearing modes using localized deposition of co-ECCD.

Internal coils, 'I-Coils', were installed inside the vacuum vessel of the DIII-D device to generate non-axisymmetric magnetic fields to act directly on the plasma. These fields are predicted to stabilize the resistive wall mode (RWM) branch of the long-wavelength external kink mode with plasma beta close to the ideal wall limit. Feedback using these I-Coils was found to be more effective as compared to using external coils located outside the vacuum vessel. Locating the coils inside the vessel allows for a faster response and the coil geometry also allows for better coupling to the helical mode structure. Initial results were reported previously (Strait E.J. et al 2004 Phys. Plasmas 11 2505). This paper reports on results from extended feedback stabilization operations, achieving plasma parameters up to the regime of Cβ ≈ 1.0 and open loop growth rates of γopenτw ≳ 25 where the RWM was predicted to be unstable with only the 'rotational viscous stabilization mechanism'. Here Cβ ≈ (β - βno-wall.limit)/(βideal.wall.limit - βno-wall.limit) is a measure of the beta relative to the stability limits without a wall and with a perfectly conducting wall, and τw is the resistive flux penetration time of the wall. These feedback experimental results clarified the processes of dynamic error field correction and direct RWM stabilization, both of which took place simultaneously during RWM feedback stabilization operation. MARS-F modelling provides a critical rotation velocity in reasonable agreement with the experiment and predicts that the growth rate increases rapidly as rotation decreases below the critical. The MARS-F code also predicted that for successful RWM magnetic feedback, the characteristic time of the power supply should be limited to a fraction of the growth time of the targeted RWM. The possibility of further improvements in the presently achievable range of operation of feedback gain values is also discussed.

The DIII-D mission to explore the advanced tokamak (AT) regime places significant demands on the DIII-D plasma control system (PCS) [D.A. Humphreys, et al., Fusion Eng. Des. 66–68 (2003) 633], including simultaneous and accurate regulation of plasma shape, stored energy, and density, as well as coordinated suppression of magnetohydrodynamic (MHD) instabilities. The present work describes selected new solutions to key control problems in DIII-D AT operation. These include novel nonrigid, resistive linear plasma response models, nonlinear algorithms for avoidance of current limiting, and improved neoclassical tearing mode control algorithms.

The advanced tokamak programme on DIII-D is to develop the scientific basis for steady state, high performance operation in future devices. We report on experiments attempting to demonstrate sustainment of 100% noninductive current for several seconds at high beta, using up to 2.5 MW of off-axis electron cyclotron current drive and up to 15 MW of neutral beam injection with q95 ≈ 5. A 100% noninductive condition was achieved with βT = 3.6%, βN = 3.5, H89 = 2.4 and improved current drive alignment. However, the duration of this phase was limited by the pressure profile evolution leading to magnetohydrodynamic instabilities after about 0.7 s. In a separate discharge, a nearly (~90%) noninductive, stationary condition was maintained for one current relaxation time (1.8 s) only limited by the duration of the hardware system. These experiments have achieved normalized fusion performance with bootstrap current fraction fBS ≈ 60%, consistent with requirements for the ITER Q = 5 steady-state scenarios. The modelling tools that were successfully employed to devise experiments in DIII-D are applied to ITER, indicating that full noninductive operation is plausible for an ITER steady-state scenario with Q ≈ 5.

The measurement of the plasma current profile is crucial to many operating regimes and investigations on the DIII-D tokamak. The measurement is required to obtain accurate equilibria and to accurately calculate stability and transport characteristics of the plasma. The measurement of the profile is also required to obtain the different components of the current to guide efforts on the control of the current profile and experiments toward obtaining steady-state operating regimes. The edge current profile measurement is necessary to understand the formation of edge pedestal and edge-localized modes. The DIII-D tokamak has a three-array, 45-channel motional Stark effect (MSE) diagnostic to measure the plasma current density and radial electric field. A 32-channel lithium-beam (Li-beam) diagnostic has recently been installed on the DIII-D tokamak for the measurement of edge current density. Both diagnostics measure the current profile from the measurement of the pitch angle of the magnetic field that, in turn, is derived from the orientation angle of polarization of the appropriate neutral beam spectral line. The MSE and the Li-beam diagnostics are described, and some examples of measurements are shown.

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with ?N ? 3.5 and ?T ? 3.6% sustained for up to 1?s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by results of these and other recent experiment and simulation efforts.

A local analysis technique is presented for the analysis of MSE data to deduce the safety factor q in tokamak discharges. The technique preserves as much as possible the individuality of every single measure by a simple rule of translation of magnetic field pitch angle measurements into q-values. Based on a geometric approach, and the observation that the flux surfaces shapes are strongly constrained by that of the last closed flux surface (LCFS), by the position of the magnetic axis and by a few more global parameters, it provides a robust, nonsubjective, accurate technique that is useful for the experimental study of q-profiles and for the evaluation of its uncertainties. It also provides a useful tool for plasma control experiments as it does not submit data to a preliminary search of minima in a multi-parametric domain, a procedure that may lead to jumps in the time behavior of the produced results.

The local oscillating component of the poloidal magnetic field in plasma associated with magnetohydrodynamic (MHD) instabilities has been measured using the motional Stark effects (MSE) diagnostic on the DIII–D tokamark. The magnetic field perturbations associated with a resistive wall mode rotated by internal coils at 20 Hz was measured using the conventional MSE operation mode. These first observations of perturbations due to a MHD mode were obtained on multiple MSE channels covering a significant portion of the plasma and the radial profile of the amplitude of the perturbed field oscillations was obtained. The measured profile is similar to the profile of the amplitude of the electron temperature oscillation measured by electron cyclotron emission (ECE) measurements. In a new mode of measurement, the amplitude of a tearing mode rotating at a high frequency (∼7 kHz) was observed using the spectral analysis of high frequency MSE data on one channel. The spectrum consists of the harmonics of the light mod...

A set of twelve coils for stability control has recently been installed inside the DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] vacuum vessel, offering faster time response and a wider range of applied mode spectra than the previous external coils. Stabilization of the n=1 ideal kink mode is crucial to many high beta, steady-state tokamak scenarios. A resistive wall converts the kink to a slowly growing resistive wall mode (RWM). With feedback-controlled error field correction, rotational stabilization of the RWM has been sustained for more than 2.5 s. Using the internal coils, the required correction field is smaller than with the external coils, consistent with a better match to the mode spectrum of the error field. Initial experiments in direct feedback control have stabilized the RWMs at higher beta and lower rotation than could be achieved by the external coils in similar plasmas, in qualitative agreement with numerical modeling. The new coils have also allowed wall stabilization in plasmas with...

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these condi...

H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the ? limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4?s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, Er, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge Er well. The Er at the bottom of the well in these cases is about ?100?kV?m?1 compared with ?20 to ?30?kV?m?1 in the standard H-mode. The well is about 1?cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal ? and ?* values matching the values desired for ITER have been achieved. In these higher density plasmas, the Er well is significantly shallower and broader.

Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The experimental ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co- and counter-injection; the ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (TORAY-GA) as well as a quasilinear Fokker–Planck model (CQL3D) and is found to be in better agreement with the more complete Fokker–Planck calculation, especially when the rf power density and/or loop voltage exceed criterion for substantial nonlinear modification of the electron distribution function. The width of the measured ECCD profile is consistent with the theoretically expected width in the absence of radial transport for the current carrying electrons.

Using off-axis electron cyclotron current drive (ECCD), self-consistent integrated advanced tokamak operation has been demonstrated on DIII-D [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159], combining high β (>3%) at high q (qmin>2.0) with good energy confinement (H89∼2.5) and high noninductive current fraction (fBS∼55%,fNI∼90%). Modification of the current profile by ECCD led to internal transport barrier formation even in the presence of type I edge localized modes. Improvements were observed in all transport channels, and increased peaking of profiles led to higher bootstrap current in the core. Separate experiments have shown the ability to maintain a nearly steady-state current profile for up to 1 s with qmin>1.5. Modeling indicates that this favorable current profile can be maintained indefinitely at a higher βN using tools available to the near-term DIII-D program. Modeling and simulation have become essential tools fo...

A magnetohydrodynamic (MHD) instability which has the characteristics that the displacement phase is in the same direction at all the affected flux surfaces (an interchange-like structure with no phase reversal across a flux surface), has been observed during the L mode in several negative central shear discharges of DIII-D tokamak [J. L. Luxon, P. Anderson, F. Baity et al., Plasma Physics and Controlled Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. The instability occurs when the minimum safety factor is around 2.0 and the profile of the safety factor q is deeply reversed in the center. Detailed stability analyses were carried out using standard numerical codes and the high quality magnetic probe, motional Stark effect and electron cyclotron emission (ECE) data. Analysis of the data after the onset shows that the instability has an n=1 mode number and a growth time of about 400 μs. The electron temperature fluctuations obtained from ECE measurements indicate a ...

Electron cyclotron current drive (ECCD) experiments on the DIII-D tokamak are solidifying the physics basis for localized, off-axis current drive, the goal being to validate a predictive model for ECCD. The ECCD profiles are determined from the magnetic field pitch angles measured by motional Stark effect polarimetry. The measured ECCD switches from the co- to the counter-direction as the toroidal injection angle is varied with a profile width that is in accordance with ray tracing calculations. Tests of electron trapping in low beta plasmas show that the ECCD efficiency decreases rapidly as the deposition is moved off-axis and towards the outboard side of the plasma, but the detrimental effects of electron trapping on the current drive are greatly reduced in high beta plasmas. Overall, the measured ECCD is in good agreement with theoretical calculations using a quasilinear Fokker–Planck code over a wide range of injection angles and plasma parameters.

A quick yet robust analysis technique has been developed to interpret measurements from the motional Stark effect (MSE) polarimetry diagnostic. The MSE diagnostic measures the vertical component of the magnetic field (Bz) as a function of the plasma major radius. Typically these measurements have been used as a constraint on a numerical calculation of the plasma equilibrium. The analysis technique presented here, which converts the MSE measurements of Bz to quantities of interest using Maxwell's equations and geometric constraints, has some advantages over equilibrium reconstruction analysis because it does not introduce assumptions about the shape of the current profile into the analysis. Further, the results are available immediately after the MSE data are collected. Examples of this MSE analysis technique are given using recent electron cyclotron current drive experiments on the DIII-D tokamak.

Sustained stabilization of the n=1 kink mode by plasma rotation at beta approaching twice the stability limit calculated without a wall has been achieved in DIII-D by a combination of error field reduction and sufficient rotation drive. Previous experiments have transiently exceeded the no-wall beta limit. However, demonstration of sustained rotational stabilization has remained elusive because the rotation has been found to decay whenever the plasma is wall stabilized. Recent theory [Boozer, Phys. Rev. Lett. 86, 5059 (2001)] predicts a resonant response to error fields in a plasma approaching marginal stability to a low-n kink mode. Enhancement of magnetic nonaxisymmetry in the plasma leads to strong damping of the toroidal rotation, precisely in the high-beta regime where it is needed for stabilization. This resonant response, or “error field amplification” is demonstrated in DIII-D experiments: applied n=1 radial fields cause enhanced plasma response and strong rotation damping at beta above the no wal...