Degradation Of Energy Confinement Research Articles

First-principles based plasma profile predictions for optimized stellarators

In the present Letter, first-of-its-kind computer simulations predicting plasma profiles for modern optimized stellarators—while self-consistently retaining neoclassical transport, turbulent transport with 3D effects, and external physical sources—are presented. These simulations exploit a newly developed coupling framework involving the global gyrokinetic turbulence code GENE-3D, the neoclassical transport code KNOSOS, and the 1D transport solver TANGO. This framework is used to analyze the recently observed degradation of energy confinement in electron-heated plasmas in the Wendelstein 7-X stellarator, where the central ion temperature was ‘clamped’ to keV regardless of the external heating power. By performing first-principles based simulations, we provide key evidence to understand this effect, namely the inefficient thermal coupling between electrons and ions in a turbulence-dominated regime, which is exacerbated by the large ratios, and show that a more efficient ion heat source, such as direct ion heating, will increase the on-axis ion temperature. This work paves the way towards the use of high-fidelity models for the development of the next generation of stellarators, in which neoclassical and turbulent transport are optimized simultaneously.

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 P NBI. 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.

Characterization of density scanning experiments with NBI and LHW on EAST

This paper presents the results of the density scanning experiment on the 2018 EAST campaign to study the effect of gas fueling on energy confinement of the ELMy H-mode. The experiment is carried out in a USN configuration, with neutral beam and lower hybrid wave heating and gas fueling, with the upper triangularity δup ∼ 0.47. The total stored energy, H98, and βN decrease with normalized density. Compared to the variations in temperature at the pedestal, the core temperature decreases more significantly for both Te and Ti, leading to a large reduction in core pressure and an increase in the pedestal electron collisionality ν*e,ped. The increase in ν*e,ped could reduce the pedestal current and result in a decrease in the value of q in the core region. It was observed that the frequency of type I ELMs increases with density and the edge localized mode size becomes smaller at high density plasma. An m/n = 2/2 tearing mode was observed at the core of the plasma and can coexist with a sawtooth at low density plasma while this tearing mode disappeared at high gas fueling plasma. The reversal radius of the sawtooth (where q = 1) moves toward the magnetic axis as density increases. The degradation in performance with density may be due to two reasons: the more monotonic shear q profile and the weakening of the stabilizing effect of fast ions on ion temperature gradient modes at high density by D2 gas fueling. It seems that there is a strong link between core transport and pedestal parameters which are influenced by gas fueling, resulting in a significant degradation of energy confinement.

A new mechanism for increasing density peaking in tokamaks: improvement of the inward particle pinch with edge E × B shearing

Developing successful tokamak operation scenarios, as well as confident extrapolation of present-day knowledge requires a rigorous understanding of plasma turbulence, which largely determines the quality of the confinement. In particular, accurate particle transport predictions are essential due to the strong dependence of fusion power or bootstrap current on the particle density details. Here, gyrokinetic turbulence simulations are performed with physics inputs taken from a JET power scan, for which a relatively weak degradation of energy confinement and a significant density peaking is obtained with increasing input power. This way physics parameters that lead to such increase in the density peaking shall be elucidated. While well-known candidates, such as the collisionality, previously found in other studies are also recovered in this study, it is furthermore found that edge E × B shearing may adopt a crucial role by enhancing the inward pinch. These results may indicate that a plasma with rotational shear could develop a stronger density peaking as compared to a non-rotating one, because its inward convection is increased compared to the outward diffusive particle flux as long as this rotation has a significant on E × B flow shear stabilization. The possibly significant implications for future devices, which will exhibit much less torque compared to present day experiments, are discussed.

The energy confinement response of DIII-D plasmas to resonant magnetic perturbations

Resonant magnetic perturbations (RMPs) are a leading method for edge localized modes (ELMs) Control in fusion plasmas. However they can also cause a rapid degradation in energy confinement. In this paper we show that the energy confinement in low collisionality ( < 0.3) DIII-D ITER similar shape (ISS) plasmas often recovers after several energy confinement times for RMP amplitudes up to the threshold for ELM suppression. Immediately following the application of the RMP, the plasma stored energy decreases in proportion to the decrease in the line-averaged density during density ‘pump-out’. Later in the discharge confinement recovery is observed in the thermal ion channel and is correlated with the increase in the ion temperature at the top of the H-mode pedestal. A correlation between the inverse scale length of the ion temperature () and the shearing rate at the top of the pedestal is seen during the confinement recovery phase. Transport analysis reveals that the confinement improvement in the ion channel results from the self-similarity in the ion temperature profiles in the plasma core combined with the observed increase in in the plasma edge following density pump-out. In contrast the electron temperature scale length () remains essentially unchanged in response to the application of the RMP. At significantly higher RMP levels the edge shearing rate and does not increase and the confinement does not recover following density pump-out.

Introduction to Neoclassical Tearing Modes and the Role of Rotation

The creation of large-sized magnetic islands through excitation of neoclassical tearing modes (NTMs) and the concommitant degradation of energy confinement is a major concern for ITER. The basic physical mechanisms governing the dynamics of this instability are introduced, and important experimental observations and techniques of controlling or suppressing these modes are briefly reviewed. The effects of plasma rotation on the excitation threshold and saturated island sizes of NTMs, as observed in many recent experiments, and their present understanding as obtained from model analytical and numerical simulations are discussed.

Modification of H-mode pedestal structure with lower hybrid waves on Alcator C-Mod

The application of lower hybrid range of frequencies (LHRF) waves in H-mode plasmas on Alcator C-Mod can result in a significant reduction in core particle inventory, with no significant degradation of energy confinement. This phenomenon has been observed in steady enhanced Dα (EDA) H-mode targets, which are sustained by ion cyclotron RF auxiliary heating, in which pedestal density nped is usually tied firmly to plasma current IP and shows a strong resilience to changes in the edge neutral source. Upon application of up to 1 MW LHRF power, nped is reduced by up to 30%, while the temperature profile increases simultaneously such that the pressure pedestal remains constant or is slightly increased. Steady EDA H-mode operation with no edge-localized modes can be maintained while edge collisionality is reduced by factors of reduction of 2–4. Elevation of scrape-off layer (SOL) density and electric currents accompany the application of LHRF (at levels as low as 400 kW) with a fast time response (∼10−2 s), while full density pedestal relaxation and core density reduction occur on longer time scales (∼10−1 s). A similarly prompt counter-IP change in the edge toroidal velocity is also observed in response to LHRF, followed on longer time scales by a counter-IP change in the central rotation. The range of time scales of the plasma response may indicate that the radial locations of LHRF interactions (i.e. SOL versus core), and power deposition mechanisms, are evolving in time. Understanding the responsible physical mechanisms and applying them to a broad range of discharges could provide a tool for improved H-mode density control.

Confinement degradation with beta for ELMy H-mode plasmas in JT-60U tokamak

The degradation of energy confinement with increased toroidal beta βT was shown by non-dimensional analysis in JT-60U. The dependence of the energy confinement on βT was examined by both the JT-60U ELMy H-mode confinement database and the dedicated experiment on a single βT scan while and ν* were kept fixed as well as the other magnetic geometrical parameters. In both cases, the degradation of energy confinement with increasing βT was observed, satisfying the relation of . This dependence is a little weaker than that predicted by the IPB98(y, 2) scaling. The fusion power production rate was estimated to increase in proportion to .

Influence of Neutral Gas Fueling on Edge Instabilities in the Scrape-Off Layer of TEXTOR-94

Strong neutral gas fueling in discharges with radiative improved confinement in TEXTOR-94 is always accompanied by a considerable degradation of energy confinement, whereas moderate fueling permits to maintain the desirable features of this operational regime. Model calculations indicate that an additional edge transport effect has to be taken into account for scenarios with strong gas puffing to match the experimental observations. The suggestion that the impact of localized neutral gas distributions on high toroidal mode number ballooning-type edge instabilities might be responsible for the enhanced edge transport is investigated and it is shown that the linear perturbation analysis indeed exhibits a pronounced effect depending on the strength of the puff rate and on the position of the neutral puff as well. Mixing length estimates provide transport coefficients of the expected order of magnitude serving a possible explanation for the experimentally found effects.

Thermal instability explanation of similar density limits in gas fueled, DIII-D H-mode shots with different operating conditions

Recent experiments on DIII-D [J. L. Luxon, F. Batty, C. Baxi et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] examined the effect of different operating conditions (“open” and “closed” divertor geometry, active pumping, fueling location) on the maximum achievable density in gas fueled H-mode (high confinement mode) discharges. Several phenomena observed at these higher densities (≈0.8 the Greenwald density)—degradation in energy confinement, detachment of the core plasma from the divertor plate, multifaceted asymmetric radiation from edge formation—are found to be correlated with the predicted onset of various thermal instabilities in the plasma edge or divertor regions. The similarity of the maximum achievable densities under the different operating conditions can be related to a similarity of edge thermal instability characteristics.

Studies in JET divertors of varied geometry. II: Impurity seeded plasmas

In current large tokamaks, non-intrinsic seeded impurities have been used to produce divertor power loads which would be considered acceptable when extrapolated to ITER. Many devices have achieved the goals of high fractional radiated powers, small frequent ELMs and detachment which are characteristic of radiative H mode regimes. The influence of divertor geometry on these characteristics is described. It has been a matter of concern that the Zeff associated with the seeded impurities may exceed that allowable in ITER and also that the degradation in energy confinement may be unacceptable. Confidence can only be built in the prediction of these parameters in ITER if reliable scalings are available for impurity content and energy confinement which have a sound physics basis. Work is described at JET in this area whilst using multimachine data to characterize the size scaling and provide a context for the JET data. Predicted levels for the impurity content of seeded ITER plasmas appear to be of marginal acceptability. Discharges run in the JET Mark I, Mark IIA and Mark IIAP divertors are compared and indicate that increased divertor closure has brought relatively minor benefits in highly radiative discharges. The acceptability of the energy confinement of radiation for ITER remains unclear. Dimensionless parameter scaling experiments have been conducted in which β, q25, fractional radiated power and Zeff are held constant for a range of ρ*. The price paid for high edge radiation and small ELMs appears to be a 25% loss in total stored energy as a result of edge pedestal degradation. However, the underlying energy confinement scaling may still be consistent with gyro-Bohm scaling, which would give an adequate margin for ITER. This conclusion is, however, sensitive to the scaling of confinement with collisionality, which is difficult to determine due to the coupling between ρ* and ν* which is a consequence of radiation dominated regimes.

Energy confinement properties of H-mode discharges in the DIII-D tokamak

Neutral beam heated DIII-D expanded boundary divertor discharges have exhibited ASDEX-like H-mode behaviour over a wide parameter range. The deuterium H-mode energy confinement of 120 ms remained near the Ohmic value for up to 6 MW of neutral beam heating, where it was 2-2.5 times higher than the L-mode value at a plasma current of 1 MA. The hydrogen and helium H-mode energy confinement times were similar and substantially below the deuterium H-mode confinement time. The H-mode confinement times decreased with increasing neutral beam power and were only 30% better than the L-mode confinement times at 5 MW. In an H-mode with a mixture of hydrogen and deuterium ([H]/[H+D] ≅ 40%), the confinement time was in between the values obtained in the pure hydrogen and deuterium cases, increased linearly with plasma current for q95 > 3.2, and decreased with increasing neutral beam power. The confinement quality in these plasmas was 85 ms per MA at a heating power of 5.6 MW. The lower energy confinement in the non-deuterium H-modes and the degradation of energy confinement with neutral beam power were both accompanied by an increase in the edge localized mode (ELM) amplitude and frequency. The changing ELM characteristics make a determination of the intrinsic isotopic and neutral beam effect on confinement difficult. For values of BT < 0.9 T and q95 < 3, the confinement quality in the deuterium and hydrogen/deuterium H-modes deteriorated to values near the L-mode level. This deterioration in energy confinement was not related to operation at high beta but instead appears to be due to a combined action between sawteeth and ELMs that becomes more pronounced at low q and low BT. L-mode energy confinement was independent of ion species and in good agreement with Kaye-Goldston scaling. Odajima-Shimomura scaling disagreed with the present L-mode τE data in terms of isotopic mass dependence; their prediction for the hydrogen L-mode exceeds the present measurements by a factor of two.

Confinement and ballooning stability at the beta limit

Cross-field transport and ballooning stability in neutral beam heated ASDEX divertor tokamak plasmas below and at the β limit are analysed by computer simulations. It is found that the discharges below the limit are ballooning stable and exhibit H transport. No gradual reduction of confinement happens when beta approaches the limit (‘hard’ β limit). The degradation of energy confinement at the β limit is shown to be due to enhanced electron heat conduction. Resistive ballooning modes with high wavenumbers are found to be marginally stable or unstable in a radial zone where the electron heat diffusivity is enhanced by a factor of four. Broader zones correspond to stronger degradation of global energy confinement. The diffusion coefficient is raised much less than the electron heat diffusivity. Fast flattening of the current profile which keeps resistive ballooning modes close to marginal stability seems to occur. Such modes with high wavenumbers generate small scale, fluctuating Br fields which can cause magnetic braiding. It is shown that the confinement properties at the β limit are consistent with transport in stochastic magnetic fields.

First neutral-beam heating experiments in JET

Hydrogen beams at particle energies of up to 65 keV, total beam powers of up to 5.5 MW , and beam-pulse durations of up to 7 s have been injected into deuterium plasmas. Experiments were performed over a wide range of plasma parameters with limiter plasmas and inner-wall plasmas. The operational regime was extended by 70% over the current ohmic density limit. In medium density experiments, ion temperatures of ca . 6.5 keV were reached with electron temperatures of 4.8 keV. The expected degradation of energy confinement with additional heating was observed. At 4 MA plasma current and 8 MW total input power, the global energy confinement time is ca. 0.4 s. The metallic impurity concentration and Zeff drop with the rise of plasm a density during beam pulses. The rise of radiated power closely follows that of the density. In most cases, the highest value of the radiated power stays below 50 % of the power input, with very low radiation from the centre of the plasma.

ICRF studies on JET

Two antennae have been installed in JET and operated to the maximum design capability of the generators. 4.5 MW, 10 MJ have been coupled to the plasma which heated up to a maximum stored energy of 3 MJ with central temperatures of Te0=5 keV and Ti0=4 keV without increase of the relative impurity concentration. Degradation of energy confinement is observed according to an L mode scaling. The effect of k/sub //// shaping is discussed using a quadrupole antenna. Hydrogen and helium 3 minority heating regimes give similar results.

Power flow in neutral beam-heated limiter discharges in the D-III tokamak

Bolometric, thermocouple, Langmuir probe and infrared camera data are used to deduce the power flow in neutral beam-heated ( P input ⩽ 5 MW), dee-shaped, limiter discharges. Power loss measurements typically account for ⩽ 60% of the total input power, with the majority either deposited on the primary limiter (20–40%) or radiated away (10–25%). Infrared camera measurements indicate that the heat flux to the ion drift side of the primary limiter is significantly greater than that to the electron drift side. Charge exchange neutral analysis, diamagnetic loop data and photodiode observations are consistent with a non-thermal flux of fast, beam produced particles to the limiter. If such an effect is indeed present, it could be partially responsible for the degradation in energy confinement observed during beam heating. Analysis of data comparing neutral beam injection from the two beamlines indicates that there is a variance in the energy cofinement time which is beamline specific; this variance may be correlated to the limiter location relative to the beamlines.