Confinement Degradation Research Articles

Survey of H-mode transition and confinement from ASDEX Upgrade `H-mode standard shot'

The H-mode standard shot of ASDEX Upgrade provides an overviewof the long term evolution of machine conditions and scatter ofH-mode power threshold and confinement. This type-I ELMy H-mode isrun every day of plasma operation since 1999.After a long opening for in-vessel work,the power threshold drops rapidly by about 50% over100 discharges and then much slower by about 20% over the remaining 1000plasma dischargesmade in a campaign.At the natural density without gas puffing, confinement increases by about30%over about 500 discharges. Afterwards thepeak-to-peak scatter is about 20%, smaller athigh density.Conditioning causes changes of edge density and temperature,similar to the confinement degradation observed with gas puffing.

High density, high performance high-confinement-mode plasmas in the Joint European Torus (JET)

Recent experiments at the Joint European Torus [Rebut et al., Fusion Eng. Des. 22, 7 (1993)] aim to improve confinement quality in high-confinement-mode (H-mode) plasmas at high densities. Energy confinement time as predicted by the International Thermonuclear Experimental Reactor ITER-H98(y,2) scaling at densities near or in excess of 85% of the Greenwald density limit scaling has been obtained by (i) strong plasma shaping (triangularity 0.35<δ<0.5), or (ii) impurity seeding, or (iii) high-field side pellet injection. Slow peaking of central density without confinement degradation is observed. Loss of sawteeth and core impurity accumulation is prevented by central ion cyclotron resonance heating. In high triangularity and impurity seeded plasmas, reduction of average power loss associated with type I edge localized modes (ELMs) is found which is attributed to the occurrence of additional losses in between ELMs. Broad band magnetic fluctuations are seen which are reminiscent of regimes with small ELMs in other tokamaks. Plasma configurations have been varied to find best combinations of edge pedestal parameters and small ELM losses.

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.

Experimental Observation of the Density Limit in Hot Ion Mode Plasmas of the GAMMA 10 Tandem Mirror

We revealed experimental observation indicating density limit with gas-puff fueling in hot ion mode plasmas of the GAMMA 10 tandem mirror. The responses of the plasmas for the various gas fueling were studied. Despite the increase in the quantity of the particle source, the line average density was saturated with the degradation of the radial particle confinement at the periphery of the central plasma. It was clarified that the line density obtained with the pellet fueling is about 20–50% higher than that with gas-puffing in the same average ion temperature. The clear difference between the pellet fueling and the gas fueling was observed in the radial profile of particle deposition at the central midplane. From the consideration based on the experimental observation, a radial-loss mechanism of ions by charge exchange reaction was discussed as an explanation for the degradation of particle confinement at the plasma periphery.

Local physics basis of confinement degradation in JET ELMy H mode plasmas and implications for tokamak reactors

First results of gyrokinetic analysis of JET ELMyH mode plasmas are presented.ELMy H mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER.Relatively high performance for long durations has been achieved and thescaling appears to be favourable.It will be necessary to sustain low Zeff and highdensity for high fusion yield.The article studies the degradation in confinement and the increase inthe anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition from type I to III ELMs at the heating power threshold.Linear gyrokinetic analysis gives the growth rate γlinof the fastest growing modes.The flow shearing rate γE×B and γlin are large near the top of the pedestal. Their ratio decreases approximately at the time whenthe confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently highγE×B/γlin near the top of the pedestal for high confinement.

Dynamics of Plasma during the Formation of a Weak Negative Central Magnetic Shear Configuration Using Counter Neutral Beam Injection in the JFT-2M Tokamak.

In the JFT-2M tokamak, discharges exhibiting good performance have been produced by counterNBI after boronization. An improved core confinement mode under an H-mode edge condition is sustained for -5~~E With H89p - 1.5 and pN - 2.0 at a line-averaged electron density of around 70%-80% of the Greenwald density limit. It is found that the q-profile changes from a monotonic one having qo ~ 1 to a zero or weak negative central magnetic shear configuration having I < qmin ~ qo < 3 during improved core confinement mode using counter-NBI under an H-mode edge condition. The central electron temperature very clearly increases when the sawtooth activity disappears, while the previous result without boronization exhibited a significant degradation of the energy confinement due to the accumulation of impurities. The mechanism governing the improved core confinement mode including the density peaking and/or ITB formation due to negative shear is presently unclear.

Overview of LHD experiments

During the first two years of the LHD experiment the following resultshave been achieved: (i) higher Te (Te(0) = 4.4 keV at ⟨ne⟩ = 5.3 × 1018 m-3and Pabs = 1.8 MW); (ii) higher confinement (τE = 0.3 s, Te(0) = 1.1 keV at ⟨ne⟩ = 6.5 × 1019 m-3and Pabs = 2.0 MW); (iii) higher stored energy, Wpdia = 880 kJ at B = 2.75 T. High performance plasmas have been realized in the inward shifted magneticaxis configuration (R = 3.6 m) where helical symmetry is recovered and the particle orbitproperties are improved by a trade-off of MHD stability properties due to the appearance of amagnetic hill. Energy confinement was systematically higher than that predicted by theInternational Stellarator Scaling 95 by up to a factor of 1.6 and was comparable with theELMy H mode confinement capability in tokamaks. This confinement improvement is attributed to configuration control (inward shift of the magnetic axis) and to the formation of a high edgetemperature. The average beta value achieved reached 2.4% at B = 1.3 T, the highest betavalue ever obtained in a helical device, and so far no degradation of confinement by MHDphenomena has been observed. The inward shifted configuration has also led to successful ICRFminority ion heating. ICRF powers up to 1.3 MW were reliably injected into the plasma withoutsignificant impurity contamination, and a plasma with a stored energy of 200 kJ was sustainedfor 5 s by ICRF alone. As another important result, long pulse discharges of more than 1 min were successfully achieved separately with an NBI heating of 0.5 MW and with an ICRFheating of 0.85 MW.

R&D of the heat-resistant LH antenna

Two different types of Lower Hybrid Current Drive (LHCD) antenna having high-heat resistant property are developed according to a pulse length and/or working frequency. In the first type of LH antenna, heat resistive material is installed at the antenna mouth only and should be attachable using bolts. This heat-resistant mouth is exchanged for a new one when it is damaged. The heat-resistant mouth is made of Carbon Fiber Composite (CFC)-block bonded with stainless steel-base machined to fit the waveguides. This mouth has no Cu-plating for reducing rf-losses, because the antenna is designed to operate during 10–100 s injection of 2 GHz in JT-60SC. The second antenna consists of a compact multi-junction (MJ) part and a power divider designed for the next generation tokamak machine like ITER. This MJ must be made of low-Z material as a plasma facing component allowing no degradation in confinement. The test module for the MJ is made of rods and plates of CFC with Cu-plating ∼30 cm in length. The test module shows transmitted power availability up to 45 MW/m 2 over 1000 s with cooling at 3.7 GHz. For improvement of transmitted power availability, R&D of strong Cu-plating on CFC surface has been progressed using an electrochemical method.

MHD characteristics in the high beta regime of the Large Helical Device

The highest volume averaged beta values βt of 2.2% at Bt = 0.75 T (gas puff) and 2.4% at Bt = 1.3 T (pellet) in helical devices have been achieved in LHD. The βt dependence of MHD activities has been investigated in NBI plasmas. The n/m = 1/2 mode, excited in the core region, and the ι - = 1 resonant modes, in the peripheral region, have been observed. Both of the fluctuation amplitudes increase with βt and pressure gradient. The strong n/m = 1/2 mode, which affects the plasma profile, has been observed in high βt discharges, and the abrupt disappearance of the mode leads to restoration of the Te profile. Violent instabilities which terminate the plasma and degradation of global energy confinement have not been observed so far.

A calculation model for density limits in auxiliary heated, gas fueled tokamaks and application to DIII-D model problems

A model for the calculation of density limits in high confinement (H-mode) tokamaks is described. The model consists of coupled calculations of (1) the power, particle and momentum balances for the core plasma and for the divertor/SOL (scrape-off layer) plasma; (2) the transport of fueling and recycling neutrals; (3) pedestal gradient scale lengths and MHD (magnetohydrodynamic) limits; (4) confinement degradation due to thermal instabilities in the edge transport barrier; (5) divertor and core MARFE (multifaceted asymmetric radiation from edge) onset; (6) radiative collapse leading to disruption, and (7) power threshold for high-to-low mode transition. The model is applied to study the effects of different operational parameters (e.g., P, I, B) on the density limit for auxiliary heated, gas fueled DIII-D [J. L. Luxon, F. Batty, C. B. Baxi et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] model problems. Predicted values of densities, temperatures, and edge gradient scale length are generally consistent with measured values, and many of the phenomena observed experimentally in association with density limits are predicted. It is found that the MARFE density limit increases strongly with increasing auxiliary heating power per unit plasma surface area (i.e., power flux exiting the core into the SOL) and with increasing plasma current, and decreases weakly with increasing magnetic field and with increasing carbon impurity concentration.

Modelling of confinement degradation in the radiative improved mode causedby a strong gas puff

Confinement deterioration and rollover from the radiative improved (RI)mode to the low (L) confinement mode in plasmas with a strong gas puff inthe tokamak TEXTOR-94 are modelled by the one-dimensional transport codeRITM (radiation from impurities in the transport model). The anomaloustransport coefficients in the plasma core include contributions from theion temperature gradient and dissipative trapped electron instabilities.This model allows us to reproduce the L-RI bifurcation with impurityseeding in good agreement with experimental observations. Whereas thetransport at the plasma edge under the RI-mode conditions might bedescribed by the electrostatic turbulence caused by electric currents inthe scrape-off layer of the limiter, the present computations show thatthe level of the edge transport must be increased by at least a factor ofthree in order to reproduce the evolution of the plasma density profileand the effective ion charge during RI-to-L back transitions. An increaseof this order is in agreement with reflectometer measurements and couldtentatively be explained by the effect of neutrals on resistive driftmodes.

Improved plasma performance on Large Helical Device

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position Rax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with Rax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta 〈β〉 of 2.4%. An improved radial profile of electron temperature was also achieved in the configuration with magnetic islands, minimized by an external perturbation coil system for the Local Island Divertor (LID). The LID has been proposed for remarkable improvement of plasma confinement like the high (H) mode in tokamaks, and the LID function was suggested in limiter experiments.

High-density H-mode operation achieved using efficient plasma refueling by inboard pellet launch

A systematic study was performed on high-density H-mode operation in the tokamaks ASDEX Upgrade and JET using inboard pellet launch refueling. The pellet particle flux was found to correlate with the achieved density enhancement. After injection of each pellet the decay of the density enhancement starts on a fast time scale until about half of the pellet inventory is expelled, slowing down significantly as the base density is approached. Whereas the overall slow decay happens in the particle confinement time, its first phase results from a sequence of ELMs following each injection. Loss of particles in an ELM sequence is accompanied by a loss of energy, causing a reduction of the plasma energy. Full plasma energy recovery after an ELM sequence occurs faster than the slow density decay, allowing transient operation at high densities maintaining full confinement. However, confinement degradation by inappropriate discharge scenarios must be avoided. Pellet-induced ELM bursts result in a particle flux from the plasma recycling at the wall, adding up with gas fluxes from other sources. Insufficient pumping can then lead to a neutral gas pressure increase causing confinement degradation. Also, excessive temperature reduction by pellets close to rational surfaces can create conditions likely to catalyze the growth of neoclassical tearing modes (NTMs) at high β N, which may then be triggered by a succeeding pellet.

Calculation of the effects of pumping, divertor configuration and fueling on density limit in a tokamak model problem

Several series of model problem calculations have been performed to investigate the predicted effect of pumping, divertor configuration and fueling on the maximum achievable density in diverted tokamaks. Density limitations due to thermal instabilities (confinement degradation and multifaceted axisymmetric radiation from the edge) and to divertor choking are considered. For gas fueling the maximum achievable density is relatively insensitive to pumping (on or off), to the divertor configuration (open or closed), or to the location of the gas injection, although the gas fueling rate required to achieve this maximum achievable density is quite sensitive to these choices. Thermal instabilities are predicted to limit the density at lower values than divertor choking. Higher-density limits are predicted for pellet injection than for gas fueling.

Understanding of H mode pedestal characteristics using the multimachine pedestal database

With use of a multimachine pedestal database, essential issues for each regime of ELM types areinvestigated. They include (i) understanding and prediction of pedestal pressure during type I ELMs,a reference operation mode of a future tokamak reactor; (ii) identification of the operation regime oftype II ELMs, which have small ELM amplitude with good confinement characteristics; (iii) identificationof the upper stability boundary of type III ELMs for access to the higher confinement regimes withtype I or II ELMs; (iv) understanding the relation between core confinement and pedestal temperaturein conjunction with the confinement degradation in high density discharges. Both scaling and modelbased approaches for expressing pedestal pressure are shown to roughly scale the experimental datawell and could be used to make initial predictions for a future reactor. q and δ are identifiedas important parameters for obtaining the type II ELM regime. A theoretical model of type III ELMs isshown to reproduce the upper stability boundary reasonably well. It is shown that there exists somecritical pedestal temperature below which the core confinement starts to degrade. It is also shown thatit is possible to obtain improved pedestal conditions for good confinement in high density dischargesby increasing the plasma triangularity.

Divertor target heat load reduction by electrical biasing, and application to COMPASS-D

A toroidally asymmetric potential structure in the scrape-off layer (SOL) plasma may be formed by toroidally distributed electrical biasing of the divertor target tiles. The resulting E× B convective motions should increase the plasma radial transport in the SOL and thereby reduce the heat load at the divertor [R.H. Cohen, D.D. Ryutov, Nucl. Fus. 37 (1997) 621]. In this paper, we develop theoretical modelling and describe the implementation of this concept to the COMPASS-D divertor. We show that a strong magnetic shear near the X-point should cause significant squeezing of the convective cells preventing convection from penetrating above the X-point. This should result in reduced heat load at the divertor target without increasing the radial transport in the portion of the SOL in direct contact with the core plasma, potentially avoiding any confinement degradation. Implementation of divertor biasing is in hand on COMPASS-D involving insulation of, and modifications to, the present divertor tiles. Calculations based on measured edge parameters suggest that modest currents ∼8 A/tile are required, at up to 150 V, to drive the convection. A technical test is preceding full bias experiments.

Pellet fuelling and ELMy H mode physics in JET

As the reference operating regime for ITER, ELMy H mode investigations have received high priority in the JET experimental programme. Recent experiments have concentrated in particular on operation simultaneously at high density and high confinement using high field side (HFS) pellet launch. The enhanced fuelling efficiency of HFS pellet fuelling is found to scale favourably to a large machine such as JET. The achievable density of ELMy H mode plasmas in JET has been significantly increased using HFS fuelling, although at the expense of confinement degradation down to L mode levels. Initial experiments using control of the pellet injection frequency have shown that density and confinement can simultaneously be increased close to the values necessary for ITER. The boundaries of the available ELMy H mode operational space have also been extensively explored. The power necessary to maintain the high confinement normally associated with ELMy H mode operation is found to be substantially higher than the H mode threshold power. The compatibility of ELMy H modes with divertor operation acceptable for a fusion device has been studied. Narrow energy scrape-off widths are measured which place stringent limits on divertor power handling. Deuterium and tritium co-deposition profiles are measured to be strongly in-out asymmetric. Successful modelling of these profiles requires inclusion of the (measured) scrape-off layer flows and of the production in the divertor of hydrocarbon molecules with sticking coefficients below unity. Helium exhaust and compression are found to be within the limits sufficient for a reactor.

ECRH at high heating power density in FTU tokamak

Plasma absorption at the fundamental EC resonance of e.m. waves at 140 GHz, 0.8 MW, Ordinary mode, provides in the FTU tokamak a localized (minimum δ r ≃2.5 cm) and intense (up to ≃100 MW/m 3) heat source for the electrons. It is shown that at high density (≃10 20 mγ− -3) ions are heated (Δ T i / T i =0.25) by collisional e−i energy transfer and assuming ion thermal diffusivity about three times larger (at most) than predictions in order to account for the observed T i . Off-center localization ( r abs/ a ≃0.2) of the ECRH heat source changes the electron temperature (and current density) profile in a way to fully stabilize sawteeth. On-axis absorption accelerates the sawtooth crash repetition rate if P ecrh⩽ P oh, but with P ecrh> P oh the sawtooth period is increased. A critical shear model for the sawtooth crash repetition rate is shown to be in agreement with its experimental dependence on resonance position and heating power. In discharges characterized by a flat temperature profile in the ohmic phase, ECRH may de-stabilize tearing modes of different order, with a strong degradation of core energy confinement.

The role of thermal instabilities in limiting the density in DIII-D

The mechanisms which apparently govern the maximum achievable density in several DIII-D [Luxon, Anderson, Batty et al., Plasma Physics on Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 159] shots in which different operating procedures were used in building up the density were investigated and compared with the predictions of thermal instability theory. Core MARFEs (multifaceted asymmetric radiation from the edge) followed by a H–L (high-to-low confinement mode) transition limit the density well below the Greenwald limit in continuous gas puffed lower single-null discharges with low triangularity. Similar continuous gas puff fueled discharges with higher triangularity or with pumping in the lower divertor achieve densities at or above the Greenwald value, apparently limited by confinement degradation, without the formation of core MARFEs. Pellet fueled discharges achieve densities up to twice the Greenwald value, limited by global radiative collapse. Thermal instability theory predictions of the limiting core MARFEs, confinement degradation or global radiative collapse are in good agreement with the experimental observations for the shots examined. Evidence for an important role of neutral particles in the plasma edge in core MARFE onset and in confinement degradation was identified.