Engineering topics

Grierson, B. A.

Publications and source records attributed to Grierson, B. A..

At least 19 records

DOE OSTI2023

Prompt core confinement improvement across the L–H transition in DIII-D: Profile stiffness, turbulence dynamics, and isotope effect

Here, we elaborate on the nature of the prompt core confinement improvement observed at the L–H transition in DIII-D, which is a long-standing issue unsolved for more than two decades and can impact future fusion reactor performance. Dynamic transport analysis suggests the essential role of the profile stiffness for understanding the mechanism of the prompt core confinement improvement. Beam emission spectroscopy shows that transport reduction at the core cannot be explained only by the ion scale turbulence density fluctuation suppression. Properties of nonlocal confinement improvement across the L–H transition are experimentally assessed in hydrogen (H) and deuterium (D) plasmas. Prompt core confinement improvement is found to be more rapid in the lighter hydrogen isotope.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

A New Integrated Analysis Suite for Fast-Ion Study in KSTAR

Here, an integrated workflow for fast-ion analysis was developed by adapting the One Modeling Framework for Integrated Task (OMFIT) workflow manager to support a standard and unified analysis platform for KSTAR users. The newly established analysis suite offers a graphical user interface–based workflow to enable users to readily access and handle experimental data archived in various data formats and servers. Further, users can analyze the data by importing modules designed for conducting certain tasks, such as profile fitting, equilibrium reconstruction, and postprocessing of tokamak data. The procedures for preparing the inputs for fast-ion simulations are streamlined by a common workflow manager, which enables the parallel processing of various tasks to efficiently analyze large fast-ion datasets. The OMFIT platform comprises a flexible Python-based application that enables users to freely manipulate the Python scripts for applications that are unavailable in the standard workflow. The framework also offers mapping tools to translate the output data into the Integrated Modeling and Analysis Suite format to maintain application compatibility for future ITER burning plasma experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Details of the neutral energy distribution and ionization source using spectrally resolved Balmer-alpha measurements on DIII-D

Spectrally resolved passive Balmer- α (D- α, H- α) measurements from the DIII-D 16 channel edge main-ion charge exchange recombination system confirm the presence of higher energy neutrals (“thermal” neutrals) in addition to the cold neutrals that recycle off the walls in the edge region of DIII-D plasmas. Charge exchange between thermal ions and edge neutrals transfers energy and momentum between the populations giving rise to thermal neutrals with energies approximating the ions in the pedestal region. Multiple charge exchange events in succession allow an electron to effectively take a random walk, transferring from ion to ion, providing a pathway of increasing energy and velocity, permitting a neutral to get deeper into the plasma before a final ionization event that contributes to the ion and electron particle fueling. Spectrally resolved measurements provide information about the density and velocity distribution of these neutrals, which has been historically valuable for validating Monte Carlo neutral models, which include the multi stage charge exchange dynamics. Here, in this study, a multi-channel set of such measurements is used to specifically isolate the details of the thermal neutrals that are responsible for fueling inside the pedestal top. Being able to separate the thermal from the cold emission overcomes several challenges associated with optical filter-based neutral density measurements. The neutral dynamics, deeper fueling by the thermal neutrals, and spectral measurement are modeled with the FIDASIM Monte Carlo collisional radiative code, which also produces synthetic spectra with a shape that is in close agreement with the measurements. By scaling the number of neutrals in the simulation to match the intensity of the thermal emission, we show it is possible to obtain local neutral densities and ionization source rates.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Mitigation of plasma–wall interactions with low-Z powders in DIII-D high confinement plasmas

Experiments with low-Z powder injection in DIII-D high confinement discharges demonstrated increased divertor dissipation and detachment while maintaining good core energy confinement. Lithium (Li), boron (B), and boron nitride (BN) powders were injected in H-mode plasmas (I p = 1 MA, B t = 2 T, P NB = 6 MW, < n e > = 3.6–5.0 · 10 19 m –3 ) into the upper small-angle slot divertor for 2 s intervals at constant rates of 3–204 mg s –1 . The multi-species BN powders at a rate of 54 mg s –1 showed the most substantial increase in divertor neutral compression by more than an order of magnitude and lasting detachment with minor degradation of the stored magnetic energy W mhd by 5%. Rates of 204 mg s –1 of boron nitride powder further reduce edge localized mode-fluxes on the divertor but also cause a drop in confinement performance by 24% due to the onset of an n = 2 tearing mode. The application of powders also showed a substantial improvement of wall conditions manifesting in reduced wall fueling source and intrinsic carbon and oxygen content in response to the cumulative injection of non-recycling materials. Furthermore, the results suggest that low-Z powder injection, including mixed element compounds, is a promising new core-edge compatible technique that simultaneously enables divertor detachment and improves wall conditions during high confinement operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Examination of stiff ion temperature gradient mode physics in simulations of DIII-D H-mode transport

A systematic evaluation of gyrokinetic and gyrofluid model predictions of ion temperature gradient (ITG) stability and transport using parameters from DIII-D high confinement mode (H-mode) plasmas has been performed. The nonlinear CGYRO code is used to make the gyrokinetic predictions, and the quasilinear TGLF model for the corresponding gyrofluid predictions. The assessments are made at three radii (normalized toroidal flux ρtor = 0.4, 0.55, and 0.7) in three different plasma scenarios with varying levels of neutral beam heating and torque. For each of the nine cases (3 radii × 3 scenarios) considered, ITG turbulence is found to be the dominant long-wavelength instability and transport mechanism. The inclusions of both transverse magnetic fluctuations and dynamic fast beam ions are stabilizing for all cases considered, with strongest effects seen at ρor = 0.4 where the fast ion population and normalized plasma pressure β = 2μ0nT/B2 are highest. The further inclusion of parallel magnetic fluctuations does not have a meaningful impact on the ITG turbulence in these scenarios, but does destabilize (in combination with fast ions) new high-frequency instabilities at ρtor = 0.4 in the high power scenarios. In each case the linear and nonlinear ITG critical gradients are predicted to be lower than the measured ITG scale lengths and their associated uncertainties. Inclusion of equilibrium flow shear in the transport predictions generally leads to an upshift in effective critical gradient rather than a qualitative change in the predicted stiffness, with stronger responses typically seen in the gyrokinetic predictions than in the gyrofluid results. However, in most cases these upshifted gradients still remain below the measured values and their uncertainties. Although the predicted critical gradients are below the measured gradients, both models predicted flux-matching gradients consistent with measured values in six of the nine cases considered, with no clear systematic over- or underprediction. Thus, while the experimental ion temperature profiles do not appear to be closely pinned to the ITG critical gradient, both gyrokinetic and gyrofluid models are able to accurately match the measured gradients reasonably well in most cases.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

The role of toroidal rotation in the very high energy confinement quality observed in super H-mode experiments on DIII-D

In this paper, we report the key role that toroidal rotation and the related ExB shear physics played in the very high energy confinement quality (H98y2>1.5) of super H-mode experiments on DIII-D. Experiments show that the energy confinement quality decreases when toroidal rotation decreases due to the decreased externally controlled torque per particle. Meanwhile, the total pedestal pressure in the experiments remains very high during the rotation and confinement quality change. TGYRO transport modeling suggests the contribution from rotation in the ExB shear is responsible for the confinement quality in excess of standard H-mode (H98y2~1). CGYRO gyrokinetic simulations reveal the governing physics in the core plasma of super H-modes: significant up-shift of nonlinear the ITG critical gradient is observed when applying ExB shear physics in the modeling based on experimental data. The effects of other physical parameters and contribution from pedestal height, which may play minor roles in this study, are also discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Numerical modeling of pedestal stability and broadband turbulence of wide-pedestal QH-mode plasmas on DIII-D

The wide-pedestal quiescent high confinement mode discovered on DIII-D in recent years is a stationary and quiescent H-mode (QH-mode) with the pedestal width exceeding EPED prediction by at least 25%. Its characteristics, such as low rotation, high energy confinement and edge localized mode-free operation, make it an attractive operation mode for future reactors. Linear and nonlinear simulations using BOUT++ reduced two fluid MHD models and awere carried out to investigate the bursty broadband turbulence often observed in the edge of wide-pedestal QH-mode plasmas. Two kinds of MHD-scale instabilities in different spatial locations within the pedestal were found in the simulations: one mild peeling–ballooning (PB) mode γ PB < 0.04ω A ) located near the minimum in E r</:sub> well propagating in ion diamagnetic drift direction; and one drift-Alfvén wave locates at smaller radius compared to E r</:sub> well propagating in the electron diamagnetic drift direction and unstable only when the parallel electron dynamics is included in the simulation. The coupling between drift wave and shear Alfvén wave provides a possible cause of the experimentally observed local profile flattening in the upper-pedestal. The rotation direction, mode location, as well as the wavenumber of these two modes from BOUT++ simulations agree reasonably well with the experimental measurements, while the lack of quantitative agreement is likely due to the lack of trapped electron physics in current fluid model. This work presents improved physics understanding of the pedestal stability and turbulence dynamics for wide-pedestal QH-mode.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Explaining the lack of power degradation of energy confinement in wide pedestal quiescent H-modes via transport modeling

Wide pedestal quiescent H (WPQH)-mode is an attractive scenario for future burning plasmas as they operate without ELMs. WPQH is characterized by formation of a wider and higher pedestal (than quiescent H-mode), and broadband fluctuations in the pedestal. Unlike conventional H-modes, where the energy confinement time reduces with increasing heating power, the WPQH plasmas reported in this paper do not show power degradation of the energy confinement. As the injected neutral beam power was increased, reduced core (ρ ≤ 0.45) transport calculated by transp, as well as increased core temperatures, pressure gradient and diamagnetic E × B shear rate were observed. The reduction in the heat transport and rapid decrease in the ion temperature gradient scale length suggest the formation of an ion internal transport barrier (ITB) that was accompanied by increased stored energy in the core. Quasilinear turbulent transport modeling using the trapped gyro Landau fluid (tglf) code was used to predict the ITB and its turbulence stability properties. By using profiles and equilibria produced by matching the transp transport fluxes with the tglf transport model within the tgyro transport solver, the energy confinement time captures the experimentally observed insensitivity to the increased PNBI. Linear stability analysis reveals that drift-wave instabilities in the core are stabilized by E × B shear, T i /T e ratio and Shafranov shift; the latter was found to have the strongest effect on the turbulence suppression at the highest heating level.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Fusion pilot plant performance and the role of a sustained high power density tokamak

Here, recent U.S. fusion development strategy reports all recommend that the U.S. should pursue innovative science and technology to enable construction of a fusion pilot plant (FPP) that produces net electricity from fusion at low capital cost. Compact tokamaks have been proposed as a means of potentially reducing the capital cost of a FPP. However, compact steady-state tokamak FPPs face the challenge of integrating a high fraction of self-driven current with high core confinement, plasma pressure, and high divertor parallel heat flux. This integration is sufficiently challenging that a dedicated sustained-high-power-density (SHPD) tokamak facility is proposed by the U.S. community as the optimal way to close this integration gap. Performance projections for the steady-state tokamak FPP regime are presented and a preliminary SHPD device with substantial flexibility in lower aspect ratio (A = 2–2.5), shaping, and divertor configuration to narrow gaps to an FPP is described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Ion thermal transport in the H-mode edge transport barrier on DIII-D

The power balance ion heat flux in the pedestal region on DIII-D increases and becomes increasingly anomalous (above conventional neoclassical) in experiments with higher temperature and lower density pedestals where the ion collisionality ($v^*_i$) is lowered toward values expected on ITER. Direct measurements of the main-ion temperature are shown to be essential on DIII-D when calculating the ion heat flux due to differences between the temperature of $D^+$ and the more commonly measured $C^{6+}$ impurity ions approaching the separatrix. Neoclassical transport calculations from NEO and non-linear gyrokinetic calculations using CGYRO are consistent with these observations and show that while neoclassical transport plays an important role, the turbulent ion heat flux due to ion scale electrostatic turbulence is significant and can contribute similar or larger ion heat fluxes at lower collisionality. Beam emission spectroscopy and Doppler backscattering measurements in the steep gradient region of the H-mode pedestal reveal increased broadband, long-wavelength ion scale fluctuations for the low $v^*_i$ discharges at the radius where the non-linear CGYRO simulations were run. Taken together, increased fluctuations, power balance calculations, and gyrokinetic simulations show that the above neoclassical ion heat fluxes, including the increases at lower $v^*_i$, are likely due to weakly suppressed ion scale electrostatic turbulence. These new results are based on world first inferred ion and electron heat fluxes in the pedestal region of deuterium plasmas using direct measurements of the deuterium temperature for power balance across ion collisionalities covering an order of magnitude from high $v^*_i$ values of 1.3 down to ITER relevant $v^*_i$ ~0.1.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2021

Design and physics basis for the upcoming DIII-D SAS-VW campaign to quantify tungsten leakage and transport in a new slot divertor geometry

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10–15 μ m of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a ‘V’ shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned ‘V’ shape predict a substantial reduction in W sourcing and SOL accumulation in either B × ∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. Finally, this coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

The role of toroidal rotation in the very high energy confinement quality observed in super H-mode experiments on DIII-D

In this paper, we report the key role that toroidal rotation and the related E × B shear physics played in the very high energy confinement quality (H 98y2 > 1.5) of super H-mode experiments on DIII-D. Experiments show that the energy confinement quality decreases when toroidal rotation decreases due to the decreased externally controlled torque per particle. Meanwhile, the total pedestal pressure in the experiments remains very high during the rotation and confinement quality change. TGYRO transport modeling suggests the contribution from rotation in the E × B shear is responsible for the confinement quality in excess of standard H-mode (H 98y2 ~ 1). CGYRO gyrokinetic simulations reveal the governing physics in the core plasma of super H-modes: significant up-shift of nonlinear the ITG critical gradient is observed when applying E × B shear physics in the modeling based on experimental data. Furthermore, the effects of other physical parameters and contribution from pedestal height, which may play minor roles in this study, are also discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Testing the DIII-D co/counter off-axis neutral beam injected power and ability to balance injected torque

DIII-D has undergone a major upgrade and successfully injected high power off-axis neutral beams (~4 MW) in both co-current and counter-current directions. This capability of high power co/counter steerable off-axis neutral beams on a major tokamak opens a unique parameter space of broad pressure and current profiles for high beta steady-state advanced tokamak (AT) scenarios, while retaining the ability to balance the injected torque for low rotation studies. This co/counter off-axis neutral beam capability is being used to validate physics-based energetic particle and thermal transport models for designing next-step facilities based on the steady-state AT approach. This paper reports on the critical evaluation of the transmitted power and energetic ion population produced by this heating and current drive system, which is assessed through visible imaging, neutron measurements and rotation profile measurements at balanced torque. Minimal losses of neutral beam power have been achieved by optimizing the strongly focused ion sources required to pass through the aperture. Tilting of the ion source has been guided by fast visible imaging and resulted in neutral beam injection along the design centerline with empirical characterization of each beam's divergence derived from the imaging data and used in the NUBEAM description of the beam injection. Through exclusive power injection of each neutral beam into MHD quiescent plasmas across a range of neutral beam voltage, perveance and plasma current we conclude that a modest reduction (~10%–15%) of transmitted power compared to on-axis, standard focus has been incurred. We report corrections that more accurately represent the injected power. Good ability to balance the neutral beam torque has been demonstrated by injecting the new off-axis counter injecting beam against the existing off-axis co-injected beam in 2.0 T, 1.0 MA, MHD quiescent L-mode plasmas. Furthermore, the torque balance studies verify the ability to operate with balanced injection, which is critical for achieving low torque and low rotation operation for physics studies and in ITER demonstration discharges.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

High fusion performance in Super H-mode experiments on Alcator C-Mod and DIII-D

The ‘Super H-Mode’ regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This regime exists due to a bifurcation of the pedestal pressure, as a function of density, that is predicted by the EPED model to occur in strongly shaped plasmas above a critical pedestal density. Experiments on Alcator C-Mod and DIII-D have achieved access to the Super H-Mode (and Near Super H) regime, and obtained very high pedestal pressure, including the highest achieved on a tokamak (pped ~ 80 kPa) in C-Mod experiments operating near the ITER magnetic field. DIII-D Super H experiments have demonstrated strong performance, including the highest stored energy in the present configuration of DIII-D (W ~ 2.2–3.2 MJ), while utilizing only about half of the available heating power (Pheat ~ 7–12 MW). These DIII-D experiments have obtained the highest value of peak fusion gain, QDT,equiv ~ 0.5, achieved on a medium scale (R < 2 m) tokamak. Sustained high performance operation (βN ~ 2.9, H98 ~ 1.6) has been achieved utilizing n = 3 magnetic perturbations for density and impurity control. Pedestal and global confinement has been maintained in the presence of deuterium and nitrogen gas puffing, which enables a more radiative divertor condition. A pair of simple performance metrics is developed to assess and compare regimes. Super H-Mode access is predicted for ITER and expected, based on both theoretical prediction and observed normalized performance, to allow ITER to achieve its goals (Q = 10) at Ip < 15 MA, and to potentially enable more compact, cost effective pilot plant and reactor designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Observation of fully detached divertor integrated with improved core confinement for tokamak fusion plasmas

Integration of divertor detachment with a high-performance (β N ~ 3, β p > 2, H 98 ~ 1.5) core plasma has been demonstrated in DIII-D high-β p (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-β p plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te < 5 eV), low particle flux, and low heat flux across the entire divertor target plate were obtained by using nitrogen seeding. This detached high-β p plasma is compatible with a newly developed detachment control system which can help optimize the nitrogen gas flow rate. Several features, i.e., the high edge safety factor in the high-β p scenario, impurity injection, closed divertor and reduced heating power requirement due to the high confinement, facilitate the achievement of full divertor detachment at lower density. Instead of degrading global performance, the divertor detachment facilitates the access to an even stronger ITB at large radius with a relatively weak ETB through self-organized synergy between ITB and ETB, leading to sustained high confinement. The strengthening of the large-radius ITB compensates for the ETB degradation associated with divertor detachment. In addition, a weak ETB naturally has smaller edge localized modes (ELMs). In particular, with neon injection, a long-period no-ELM H-mode phase has been achieved simultaneously with high-performance core and partially detached divertor plasmas. Furthermore, these results demonstrate the possibility of integrating excellent core plasma performance with an effective divertor solution, an essential step toward steady-state operation of reactor-grade plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Predicting operational windows of ELMs suppression by resonant magnetic perturbations in the DIII-D and KSTAR tokamaks

A newly developed plasma response model, combining the nonlinear two-fluid MHD code TM1 and toroidal MHD code GPEC run in ideal mode, quantitatively predicts the narrow isolated q 95 windows (Δq 95 ~ 0.1) of ELM suppression by n = 1, 2 and 3 resonant magnetic perturbations (RMPs) in both DIII-D and KSTAR tokamaks across a wide range of plasma parameters. The key physics that unites both experimental observations and our simulations is the close alignment of essential resonant q-surfaces and the location of the top of the pedestal prior to an ELM. This alignment permits an applied RMP to produce field penetration due to the lower E × B rotation at the pedestal top rather than being screened. The model successfully predicts that narrow magnetic islands form when resonant field penetration occurs at the top of pedestal, and these islands are easily screened when q 95 moves off resonance, leading to very narrow windows of ELM suppression (typically Δq 95 ~ 0.1). Furthermore, the observed reduction in the pedestal height is also well captured by the calculated classical collisional transport across the island. We recover observed q 95 , β N and plasma shape dependence of ELM suppression due to the effect of magnetic islands on pedestal transport and Peeling-Ballooning- Mode (PBM) stability. Importantly, experiments do occasionally observe wide windows of ELM suppression (Δq 95 > 0.5). Our model reveals that at low pedestal-top density multiple islands open, leading to wide operational windows of ELM suppression consistent with experiment. The model indicates that wide q95 windows of ELM suppression can be achieved at substantially higher pedestal pressure with less confinement degradation in DIII-D by operating at higher toroidal mode number (n = 4) RMPs. As a result, this can have significant implications for the operation of the ITER ELM control coils for maintaining high confinement together with ELM suppression.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

On the very high energy confinement observed in super H-mode DIII-D experiments

Analysis of recent super H-mode experiments on DIII-D shows that high rotation, not high pedestal, plays the essential role in achieving very high conﬁnement H98y2 > 1.5. Very high conﬁnement is reached early on in the H-mode phase of these discharges, when the pedestal is still very low, but after the toroidal rotation has already built-up to very high levels in the core. As the discharge evolves, the rotation drops, and so does the energy conﬁnement, despite a sustained very high pressure pedestal. During this evolution, the conﬁnement quality is linearly correlated with the core toroidal rotation, which varies according to diﬀerent levels of injected neutral beam torque per particle. Core transport modeling shows that the contribution from rotation in the E×B shear is responsible for conﬁnement quality signiﬁcantly in excess of standard H-mode (H98y2 ∼ 1).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Multi-scale transport in the DIII-D ITER baseline scenario with direct electron heating and projection to ITER

Multi-scale fluctuations measured by turbulence diagnostics spanning long and short wavelength spatial scales impact energy confinement and the scale-lengths of plasma kinetic profiles in the DIII-D ITER baseline scenario with direct electron heating. Contrasting discharge phases with ECH + neutral beam injection (NBI) and NBI only at similar rotation reveal higher energy confinement and lower fluctuations when only NBI heating is used. Modeling of the core transport with TGYRO using the TGLF turbulent transport model and NEO neoclassical transport reproduces the experimental profile changes upon application of direct electron heating and indicates that multi-scale transport mechanisms are responsible for changes in the temperature and density profiles. Intermediate and high-k fluctuations appear responsible for the enhanced electron thermal flux, and intermediate-k electron modes produce an inward particle pinch that increases the inverse density scale length. Projection to ITER is performed with TGLF and indicates a density profile that has a finite scale length due to intermediate-k electron modes at low collisionality and increases the fusion gain. For a range of E×B shear, the dominant mechanism that increases fusion performance is suppression of outward low-k particle flux and increased density peaking.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗