Table of contents

The 1997 DT experiment (DTE1) at the Joint European Torus included studies of the behaviour of alpha particles in high temperature plasmas. Alpha particle heating was clearly observed in a series of 10 MW hot ion H modes by scanning the DT mixture from 0% T to 93% T. Maxima in central temperature and energy content were obtained which corresponded with the maximum in fusion yield. Alfvén eigenmodes (AEs) have been detected in JET, driven by NBI or ICRH fast ions. However, in agreement with theory, no AE activity was observed in DT plasmas which could be attributed to alpha particle drive, except in the afterglow of some optimized shear pulses. Ion cyclotron emission (ICE) was detected at harmonics of the alpha particle cyclotron frequency at the outer edge of the plasma. The ICE is interpreted as being close to magnetoacoustic cyclotron instability, driven by inverted alpha distributions at the plasma edge. The high energy neutral particle spectra showed features which are ascribed to a mixture of alphas, neutralized by helium-like impurities, and deuterons born from elastic collisions with alpha particles and neutralized by hydrogen-like impurities. The results of all these studies are consistent with classical alpha particle trapping and slowing down. Future DT experiments will aim to increase alpha particle pressure so that interactions with plasma instabilities can be studied. The measurement of knock-on neutral triton spectra offers an unambiguous way to determine confined alpha densities in these future experiments.

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low q regime keeping a large radius of the internal transport barrier (ITB), and a record value of equivalent fusion multiplication factor in JT-60U, Q_DT^eq = 1.25, has been achieved at 2.6 MA. Operational schemes to reach the low q regime with good reproducibility have been developed. The reduction of Z_eff was obtained in the newly installed W shaped pumped divertor. The β limit in the low q_min regime, which limited the performance of L mode edge discharges, has been improved in H mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of the ITB and improved confinement for 5.5 s has been demonstrated in an ELMy H mode reversed shear discharge.

The radiative improved (RI) mode is a tokamak regime offering many attractive reactor features. In the article, the RI mode of TEXTOR-94 is shown to follow the same scaling as the linear ohmic confinement regime and is thus identified as one of the most fundamental tokamak operational regimes. The current understanding derived from experiments and modelling of the conditions necessary for sustaining the mode is reviewed, as are the mechanisms leading to L-RI mode transition. The article discusses the compatibility of high impurity seeding with the low central power density of a burning reactor, as well as RI mode properties at and beyond the Greenwald density.

A high ion temperature (T_i) mode is observed for neutral beam heated plasmas in the Compact Helical System (CHS) heliotron/torsatron. The high T_i mode plasma is characterized by a high central ion temperature, T_i(0), and is associated with a peaked electron density profile produced by neutral beam fuelling with low wall recycling. Transition from L mode to high T_i mode has been studied in CHS. T_i(0) in the high T_i mode discharges reaches 1 keV, which is 2.5 times higher than that in the L mode discharges. The ion thermal diffusivity is significantly reduced by a factor of more than 2-3 in the high T_i mode plasma. The ion loss cone is observed in neutral particle flux in the energy range 1-6 keV with a narrow range of pitch angle (90° ±20°) in the high T_i mode. However, the degradation of ion energy confinement due to this loss cone is negligible.

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 10¹⁸m^-3·keV·s at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 10¹⁹ m^-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to ⟨β⟩ = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (∼40 kW/m³) and electron density (3 × 10¹⁹ m^-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.

The effects of the plasma profile on the global energy confinement have been studied in Heliotron E with special regard to differences between heating methods (ECH, NBI and NBI + ECH). With high power NBI, peaked T_i and peaked n_e profiles (T_i(0)/⟨T_i ⟩ ≲ 2.7, n_e(0)/⟨n_e⟩ ≲ 4.5) were simultaneously achieved under low recycling conditions. A peaked n_e profile (n_e(0)/⟨n_e⟩ ≳ 2.5 ) could lead to the high T_i mode where the ion heat transport in the central region is substantially reduced. By changing the ECH launching condition (on-axis, off-axis and toroidally oblique injection), the peakedness of the T_e profile could be controlled in the range 1.3 ≲ T_e(0)/⟨T_e⟩ ≲ 4.5. A peaked T_e profile and a flat n_e profile (3.5 ≲ T_e(0)/⟨T_e⟩, n_e(0)/⟨n_e⟩ ≲ 1.8) were brought about by the well focused on-axis ECH. The ECH plasma with a peaked T_e profile has higher stored energy than that with a moderately peaked T_e profile for the same injected ECH power and the same density region. The global energy confinement time normalized by the LHD scaling, τ_E^G/τ_E^LHD, showed n_e(0)/⟨n_e⟩ dependence for the low T_i mode NBI plasmas. For the high T_i mode, the n_e(0)/⟨n_e⟩ dependence of τ_E^G/τ_E^LHD was weak. These findings suggest that the LHD scaling should be modified to scale the global energy confinement of the helical plasmas in a wide range of n_e(0)/⟨n_e⟩.

In order to improve the particle control capability of the divertor and to demonstrate the possibility of a dense and cold divertor with high confinement plasmas, a closed divertor study has been carried out on the JFT-2M tokamak. When a strong gas puff is applied into the divertor chamber, a baffling effect is revealed by enhanced radiation localization and high neutral pressure in the divertor chamber, low core fuelling and sustainment of core confinement quality. The baffling effect is further enhanced with E × B flow and/or current in the scrape-off layer by applying divertor biasing. A dense and cold divertor state (n_e^div ≈ 4 × 10¹⁹m^-3 and T_e^div ≈ 4 eV), together with the improved confinement modes, can be obtained by strong gas puffing. Furthermore, the improved core confinement is not affected significantly in the high density region (_e/n_e^G < 0.7). The UEDA code simulation reproduces the baffling effect and a dense and cold divertor plasma with a closed divertor structure.

High density, high radiated power fraction and small ELMs are key elements of the current ITER design. In JET, these conditions are shown to be associated with high ELM frequency, low pedestal pressure and correspondingly reduced global energy confinement time. The article reviews the current understanding of the connections between these parameters and the mechanisms for the saturation of the density with gas fuelling.

The article presents an overview of recent experimental results obtained on the RFX device. The authors obtained and studied a reversed field pinch plasma with a plasma current of up to 1 MA, negligible radiation losses and low effective charge. The local power and particle balance shows that in standard operation the plasma core is dominated by magnetic turbulence and that the global confinement is mainly provided by the edge region, where a strongly sheared radial electric field is present. With poloidal current drive the amplitude of magnetic fluctuations and the thermal conductivity of the plasma core are reduced, leading to improved confinement. Reduced heat transport is also observed when the width of the n spectrum of magnetic fluctuations is reduced.

The central cell density and the diamagnetic signal were doubled as a result of plug potential formation by ECRH in hot ion mode experiments on the GAMMA 10 tandem mirror. In order to obtain these remarkable results, the axisymmetrized heating patterns of ECRH and ICRF heating were optimized. Furthermore, conducting plates were installed adjacent to the surface of the plasma along the flat shaped magnetic flux tube located in the anchor transition regions; the plates may contribute to the reduction of some irregular electric fields produced possibly with ECRH in these thin flux tube regions. The conducting plates contributed to reducing the radial loss rate to less than 3% of the total particle losses, along with improvements in the reproducibility of the experiments and the controllability of the potential confinement. The increases in central cell density and diamagnetism in association with the increase in plug potentials scaled well with increasing ECRH power. A plug potential of 0.6 kV and a density increase of 100% were achieved using an ECRH power of 140 kW injected into both plug regions. The plasma confinement was improved by an order of magnitude over a simple mirror confinement owing to the tandem mirror potential formation.

The global confinement and the local transport properties of improved core confinement plasmas in JT-60U were studied in connection with E_r shear formation. In the improved core confinement mode with internal transport barriers (ITBs), these are roughly classified into `parabolic type' ITBs and `box type' ITBs. The parabolic type ITB has a reduced thermal diffusivity χ in the core region; however, the E_r shear, dE_r/dr, is not as strong. The box type ITB has a very strong E_r shear at the thin ITB layer and χ decreases to the level of neoclassical transport there. The estimated E × B shearing rate, ω_{E × B}, becomes almost the same as the linear growth rate of the drift microinstability, γ_L, at the ITB layer in the box type ITB. Experiments with hot ion mode plasmas during the repetitive L-H-L transition showed that the thermal diffusivity clearly depends on the E_r shear and the strong E_r shear contributes to the reduced thermal diffusivity.

The article reports results of experiments to further determine the underlying physics behind the formation and development of internal transport barriers (ITBs) in the DIII-D tokamak. The initial ITB formation occurs when the neutral beam heating power exceeds a threshold value during the early stages of the current ramp in low density discharges. This region of reduced transport, made accessible by suppression of long wavelength turbulence by sheared flows, is most evident in the ion temperature and impurity rotation profiles. In some cases, reduced transport is also observed in the electron temperature and density profiles. If the power is near the threshold, the barrier remains stationary and encloses only a small fraction of the plasma volume. If, however, the power is increased, the transport barrier expands to encompass a larger fraction of the plasma volume. The dynamic behaviour of the transport barrier during the growth phase exhibits rapid transport events that are associated with both broadening of the profiles and reductions in turbulence and associated transport. In some but not all cases, these events are correlated with the safety factor q passing through integer values. The final state following this evolution is a plasma exhibiting ion thermal transport at or below neoclassical levels. Typically the electron thermal transport remains anomalously high. Recent experimental results are reported in which RF electron heating was applied to plasmas with an ion ITB, thereby increasing both the electron and the ion transport. Although the results are partially in agreement with the usual E × B shear suppression hypothesis, the results still leave questions that must be addressed in future experiments.

The plasma dynamics of enhanced confinement regimes in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Despite differences in transition timescale and location, as well as the sign of the radial electric field E_r, observations suggest that E × B shear effects on turbulence induced transport play a dominant role in governing barrier dynamics in all cases. Fast confinement bifurcations are observed in the TFTR core enhanced reverse shear (ERS) regime and in the edge DIII-D H mode. Both show spontaneous E_r shear layer formation prior to the confinement change and a negative E_r well that persists as steep gradients form. These dynamics differ from those of DIII-D negative central shear (NCS) plasmas. There, slow transitions are observed when the applied torque from unidirectional beam injection is small, while faster development and more dramatic confinement improvements occur at higher applied torques. Unlike the H mode and ERS cases, the NCS core generally has a positive E_r hill and no strong E_r shear precursor. However, similarity experiments performed on TFTR indicate that ERS, L mode and NCS-like regimes can all be accessed in a continuous fashion by varying the E × B shear through changes in the applied torque at constant power. As in the DIII-D NCS case, core confinement in TFTR reverse shear plasmas improves slowly as co-rotation begins to dominate the determination of E_r, no strong E_r shear layer develops prior to that improvement, and the plasma possesses a positive E_r hill. Reductions in transport with E_r gradients of either sign are consistent with the picture of E × B shear suppression and decorrelation of turbulence. At fixed input power, intermediate levels of confinement improvement are achieved by varying the E × B shear with changes in the applied neutral beam torque. The data suggest that control over the plasma pressure profile in a reactor may be possible if an external source of E × B shear, such as might be applied with RF techniques, is used to modify the shear which otherwise occurs.

Two types of high performance scenario have been produced in JET during the DTE1 campaign. One of them, the ELM free hot ion H mode scenario, which is well known and has been extensively used in the past, has two distinct regions: the plasma core and the edge transport barrier. The results obtained during the DTE1 campaign with D, DT and pure T plasmas confirm the JET Team's previous conclusion that the core transport scales as a gyro-Bohm in the inner half of the plasma volume, recovers its Bohm type nature closer to the separatrix and behaves as neoclassical ion transport in the transport barrier. Measurements on the top of the barrier suggest that the width of the barrier is dependent on isotope and that fast ions play a key role. The other high performance scenario is a relatively recently developed optimized magnetic shear scenario with small or slightly negative magnetic shear in the plasma core. Different mechanisms of internal transport barrier (ITB) formation were tested by predictive modelling and the results compared with experimentally observed phenomena. The observed non-penetration of heavy impurities through the strong ITB, which contradicts a prediction of the conventional neoclassical theory, is discussed.

Recent progress in the study of divertor and scrape-off layer plasma phenomena in JET is reviewed. Three pumped divertors (Mark I, Mark IIA/AP and Mark IIGB) have been installed and exploited under reactor relevant conditions. With increased divertor closure, the particle exhaust rate has increased and neutral compression factors of >100 have been obtained with the Mark IIGB divertor. Helium enrichment factors of >0.2 have been measured under a wide range of conditions and satisfy the minimum requirements for ITER. Excellent power handling has been demonstrated in all divertor configurations. Fast infrared camera measurements exhibited broad deposition profiles during type I ELMs and energy densities of ∼0.12 MJ/m², which extrapolate to unacceptable levels for ITER. During the recent DT experiments, the co-deposition of tritium on cold shadowed surfaces in the inner divertor was identified as an important form of long term tritium retention. This has serious implications for the divertor design and tritium inventory in a next step tokamak. Core plasma purity has not improved with enhanced divertor closure or decreased main chamber neutral pressure. Studies of the chemical sputtering yield showed a dependence on surface temperature and hydrogen isotope. This accounts for the observation of increased carbon impurity production in Mark II (at 500 K) relative to Mark I (at 300 K). The variation of the disruptive (L mode) density limit appears to be more strongly dependent on the local impurity production rate than on the divertor geometry. Significant progress has been made in the study of divertor detachment, and volume recombination has been spectroscopically identified. With increasing isotope mass, detachment and the disruptive density limit occur at lower main plasma density, as predicted by the EDGE2D/NIMBUS codes. Using differential gas fuelling in the Mark IIGB divertor, it was possible to modify the in-out asymmetry of the divertor plasma parameters and energy deposition for the first time.

The scaling of both the L-H threshold and confinement with the mass M of hydrogenic isotopes is discussed. The confinement was found to scale differently with M in the core and the edge, and a two region model has been developed to represent the physical behaviour of each region. Identity pulses with the same profiles of the dimensionless physics parameters ρ*, ν* and β were obtained with different isotopes, H and D; this result suggests that there is no explicit mass dependence of the transport in either the core or the edge region.

The feedback control system to control plasma current and position on the HT-7 superconducting tokamak was greatly improved in early 1998. Lower hybrid current drive (LHCD) experiments with the improved control system were performed to sustain long pulse discharges and to improve plasma confinement. Partial non-inductive current drive and full non-inductive current drive for several seconds by means of LHCD were demonstrated. It was observed that plasma confinement could be considerably improved by LHCD. Experimental evidence suggests that this improvement during the LHCD phase could be due to the modification of the current profile in the outer region of the plasma. MHD modes (especially m = 2) seem unstable with such a current profile. The EFIT code was modified for the reconstruction of the magnetic surfaces in HT-7 and a test computation was performed.

The physical mechanisms of small scale density fluctuations in the frequency range5-400 kHz were investigated with correlation reflectometry in different types of ohmically heated discharges in the T-10 tokamak. A temporal formation of velocity shear in the central region of the plasma column during the discharge transition from saturated to improved ohmic confinement resulted in suppression of long wavelength quasi-coherent turbulence, while the amplitude of fluctuations with a broader frequency spectrum was not affected. The potential of correlation reflectometry was extended by simultaneous plasma probing from the low field side and high field side. Enhancement of quasi-coherent turbulence at the low field side by a factor of 5 was measured, while the other turbulence type was poloidally symmetric. A high long distance toroidal correlation of up to 40% was observed experimentally for quasi-coherent density fluctuations at a distance of 10 m after one turn around the tokamak major axis. Fluctuations propagate at an angle of about 0.5° with respect to the perpendicular to the magnetic field line, demonstrating a drift mechanism of turbulence. Plasma rotation was estimated from a radial force balance equation for ions with the radial electric field measured with a heavy ion beam probe. A comparison of the calculated plasma rotation with the measured turbulence rotation showed that the turbulence rotates in the ion diamagnetic drift direction in the plasma frame. All experimental observations are consistent with `toroidal' and `slab' ion temperature gradient turbulence as the underlying physical mechanism.

The results of recent experimental and theoretical studies concerning the effects of plasma shape and current and pressure profiles on edge instabilities in DIII-D are presented. Magnetic oscillations with toroidal mode number n ≈ 2-9 and a fast growth time γ^-1 = 20-150 μs are often observed prior to the first giant type I ELM in discharges with moderate squareness. High n ideal second ballooning stability access encourages edge instabilities by facilitating the buildup of the edge pressure gradient and bootstrap current density, which destabilize the intermediate to low n modes. Analysis suggests that discharges with large edge pressure gradient and bootstrap current density are more unstable to n > 1 modes. Calculations and experimental results show that ELM amplitude and frequency can be varied by controlling access to the second ballooning stability regime at the edge through variation of the squareness of the discharge shape. A new method is proposed to control edge instabilities by reducing access to the second ballooning stability regime at the edge using high order local perturbation of the plasma shape in the outboard bad curvature region.

MHD phenomena in ASDEX Upgrade in standard shear discharge regimes as well as in weak and reversed shear discharges are investigated. The onset of neoclassical tearing modes leads to the most serious β limit at ASDEX Upgrade. The β_p value for the onset of neoclassical tearing modes is found to be proportional to the poloidal ion gyroradius for collisionless plasmas, as proposed by the ion polarization current model. Larger collisionalities have a stabilizing effect. Sawtooth crashes or fishbones can trigger the mode, and in a few cases it appears spontaneously. The fractional energy loss due to a (3,2) mode saturates for large pressures at around 25%, in agreement with the result of theoretical calculations. Owing to its large and flexible heating power, ASDEX Upgrade is well equipped for investigations of fast particle driven or influenced MHD modes. The observed TAEs have been shown to agree well with theoretical predictions regarding the frequency as well as the drive by fast particles. Fishbones have been shown to not only affect the fast particle distribution but also directly to affect the background plasma. For low and reversed shear discharges, theoretical predictions for the growth of neoclassical tearing modes in regimes with a large fraction of bootstrap current at the rational surface, as well as for neoclassical double tearing modes, are given. In discharges with large impurity accumulation in the plasma centre, unusual MHD phenomena such as cascades of high n tearing modes and modes driven by positive pressure gradients have been found at ASDEX Upgrade. For similar discharges, double tearing modes have been investigated at TEXTOR.

Confinement in TCV electron cyclotron heated discharges was studied as a function of plasma shape, i.e. as a function of elongation, 1.1 < κ< 2.15, and triangularity, -0.65 ⩽ δ ⩽ 0.55. The electron energy confinement time was found to increase with elongation, owing in part to the increase of plasma current with elongation. The beneficial effect of negative triangularities was most effective at low power and tended to decrease at the higher powers used. The large variety of sawtooth types observed in TCV for different power deposition locations, from on-axis to the q = 1 region, was simulated with a model that included local power deposition, a growing m/n = 1 island (convection and reconnection), plasma rotation and finite heat diffusivity across flux surfaces. Furthermore, a model with local magnetic shear reproduced the experimental observation that the sawtooth period is at a maximum when the heating is close to the q = 1 surface.

Localized MHD activity observed in JT-60U and TFTR near transport barriers with their associated large pressure gradients is investigated. Stability analysis of equilibria modelling the experiments supports an identification of the MHD activity as being due to an ideal MHD n = 1 instability. Theoretical models are used to identify the dependence of the instability on the salient features of the plasma profiles: the local pressure gradient, local shear in the q profile and the proximity of rational surfaces where q ≈ m/n, with m and n the poloidal and toroidal mode numbers, respectively. The mode width is shown to depend on the value of q where the pressure gradient is largest, and increases as q decreases. The edge current density is shown to play an important role in coupling the internal mode to the plasma edge. Energetic particles in these neutral beam heated discharges are shown to drive fishbone-like modes when they exist in sufficient numbers in the vicinity of the maximum perturbation owing to the ideal instability.

New experiments on JET, COMPASS-D and DIII-D have identified the critical scalings of error field sensitivity and harmonic content effects, enabling predictions to be made of the requirements for larger devices such as ITER. Thresholds are lowest at low density, a regime proposed for H mode access on ITER. Results suggest a moderate error field sensitivity (δB/B ≈ 10^-4) for ITER, comparable with the size of its intrinsic error, although there are uncertainties in scaling behaviour. Other studies on COMPASS-D and DIII-D show that sideband harmonics to the (2, 1) component play an important role. Thus, a correction system for ITER will be important, with flexibility to correct sidebands desirable, possibly assisted by beam rotation. Such a system has been designed and is capable of reducing multiple harmonic error levels to ∼ 2 × 10^-5.

The excitation and stabilization of Alfvén eigenmodes and their impact on energetic ion confinement were investigated with negative ion based neutral beam injection at 330-360 keV into weak or reversed magnetic shear plasmas on JT-60U. Toroidicity induced Alfvén eigenmodes (TAEs) were observed in weak shear plasmas with ⟨β_h⟩ ⩾ 0.1% and 0.4 ⩽ v_b||/v_A ⩽ 1. The stability of TAEs is consistent with predictions by the NOVA-K code. New burst modes and chirping modes were observed in the higher β regime of ⟨β_h⟩ ⩾ 0.2%. The effect of TAEs, burst modes and chirping modes on fast ion confinement has been found to be small so far. It was found that a strongly reversed shear plasma with internal transport barrier suppresses AEs.

The article treats the recent development of quasi-steady ELMy high β_p H mode discharges with enhanced confinement and high β stability, where long sustainment time, an increase in absolute fusion performance and extension of the discharge regime towards low q₉₅ (∼3) are emphasized. After modification to the new W shaped pumped divertor, a long heating time (9 s) with a high total heating energy input of 203 MJ became possible without a harmful increase in impurity and particle recycling. In addition, optimization of the pressure profile characterized by the double transport barriers, optimum electron density and/or high triangularity δ made it possible to extend the performance in long pulses. The DT equivalent fusion gain Q^eq_DT ≈ 0.1 (δ = 0.16) was sustained for ∼9 s (∼50τ_E, ∼10τ*_p) and Q_DT^eq ≈ 0.16 (δ = 0.3) for 4.5 s at I_p = 1.5 MA. In the latter case with higher δ, an H factor ( = τ_E/τ^ITER89PL_E) of ∼2.2, β_N ≈ 1.9 and β_p ≈ 1.6 were sustained with 60-70% of the non-inductively driven current. In the low q₉₅ (∼3) region, the β limit was improved by the high δ (∼0.46) shape, where β_N ≈ 2.5-2.7 was sustained for ∼3.5 s with the collisionality close to that of ITER-FDR plasmas. The limit of the edge α parameter in the ELMy phase increases with δ, which is the main reason behind the improved β limit in a long pulse at high δ. The sustainable value of β_NH also increases with δ. Sustainable β_N is limited by the onset of low n resistive modes. Direct measurement of island width shows agreement with the neoclassical tearing mode theory.

Improving confinement and β limits simultaneously in long pulse ELMy H mode discharges is investigated. The product β_N H_98y serves as a useful figure of merit for performance, where β_N ≡ β/(I/aB) and H_98y is the ratio of the thermal confinement time to the most recent ELMy H mode confinement scaling established by the ITER confinement database working group. In discharges with q₀ ≈ 1 (no sawteeth) and discharges with q_min > 1.5 and negative central magnetic shear, β_N ≈ 2.9 and H_98y ≈ 1.4 are sustained for up to 2 s. Although peaked profiles are observed, steep internal transport barriers are not present. Further increases in β_N in these discharges are limited by neoclassical tearing modes (NTMs) in the positive shear region. In another recently developed regime, β_N ≈ 3.8 and H_98y ≈ 2 have been sustained during large infrequent ELMs in non-sawtoothing discharges with q₀ ≈ 1. This level of performance is similar to that obtained in ELM free regimes such as VH mode. The limitation on β_N and pulse length in these discharges is also the onset of NTMs.

The H_α line shape in front of the limiter in the HT-6M tokamak is analysed by multi-Gaussian fitting. The energy distribution of neutral hydrogen atoms reveals that H_α radiation is contributed by Franck-Condon atoms, atoms reflected at the limiter surface and charge exchange. Multi-Gaussian fitting of the H_α spectral profile indicates contributions of 60% from reflection particles and 40% from molecule dissociation to recycling. Ion temperatures in central regions are obtained from the spectral width of charge exchange components. Dissociation of hydrogen molecules and reflection of particles at the limiter surface are dominant in edge recycling. Reduction of particle reflection at the limiter surface is important for controlling edge recycling. The measured profiles of neutral hydrogen atom density are reproduced by a particle continuity equation and a simplified one dimensional Monte Carlo simulation code.

A multishot in situ pellet injection system has been constructed in the Institute of Plasma Physics. Single- and multi-pellet injection experiments were performed on the HT-6M and superconducting HT-7 tokamaks. The system proved to be convenient and reliable to operate. Pellets were fired into ohmically and LHCD and ICRF heated plasmas. Single pellet injection in ohmic discharge was found to increase the central density of HT-7 by about one half, while two pellet injection increased the central density in a step-like fashion by one half with each shot. Peaking of the electron density profile and a hollow electron temperature profile were obtained.

Internal transport barriers (ITBs) in which both the ion thermal diffusivity and electron thermal diffusivity are substantially reduced have been observed in JET. Such discharges have been obtained with DD and DT plasmas. Central ion temperatures of 40 keV and plasma pressure gradients of 10⁶Pa/m were observed in DT plasmas leading to a fusion triple product n_i0T_i0τ_E = 1.1 × 10²¹ m^-3·keV·s and producing 8.2 MW of fusion power. ITBs have been produced in both the MkII and the new Gas Box divertor configuration with similar behaviour. With the Gas Box divertor an L mode edge has so far only been produced using edge radiation cooling. For the first time, ITBs have been triggered by radiating about 40% of the power with a krypton puff. A possible scaling of the power needed to trigger an ITB with magnetic field is suggested.

Core confinement and the overall performance of the tokamak have been considerably enhanced with the combination of two improved confinement regimes. Internal transport barriers characteristic of the optimized shear regime and an edge transport barrier of the high confinement H mode regime have been superposed in the double barrier (DB) mode. In DT discharges the DB mode has resulted in a fusion gain Q a factor of 2 higher than in conventional sawtoothing steady state ELMy H mode plasmas. The DB mode has been routinely established in the new Gas Box divertor configuration on JET. Off-axis lower hybrid current drive and edge radiation cooling through impurity seeding have been used for current and pressure profile control. This has made it possible to increase the performance up to β_N = 2.4 and H⁸⁹ = 2.7 and to extend flat-top phases up to a duration of three energy confinement times, approaching steady state conditions. Modelling studies of DB mode DT operation on JET predict a fusion power in the range 20-30 MW. In ITER transport code modelling, steady state operation with a fusion power output of 1.3 GW is obtained for advanced tokamak scenarios in the DB mode.

Trace amounts of tritium (1-5% of the deuterium content) were introduced in short puffs (∼40 ms) into the low field midplane edge of steady state deuterium plasmas in JET. The transport properties of the edge and core plasma, and the time evolution of the fraction of tritium in the recycling flux, determine the evolution in space and time of the 14 MeV neutron emission following these puffs. The derived tritium transport coefficients were used to model the deuterium density profile. Good agreement of the profile shape was found near the edge. In the core, the predicted deuterium density is more peaked than the measured profile. An attempt was made to study the scaling of tritium transport with dimensionless plasma parameters. For H mode discharges the transport is best described by gyro-Bohm scaling in the core and Bohm scaling near the edge. These conclusions are based on the assumption that the dependence of particle transport on β and ν* is weak within the range of variation in these parameters in the data set.

Experiments in the ELM free hot ion H mode regime have been carried out in DT plasmas in JET. Initial experiments undertaken at constant neutral beam (NB) power (∼11 MW) demonstrated that core fuelling was dominated by wall/target recycling rather than NB fuelling and made it possible to arrange an optimum core DT mix by adjusting the DT mix in wall/target, gas and NB. High power experiments at 4.2 MA/3.6 T have successfully and reliably delivered fusion power (P_fus) up to 16.1 MW and plasma stored energy (W_dia) up to 17 MJ. The results are in good agreement with extrapolations, carried out with the TRANSP and JETTO codes, from similar deuterium discharges. Transiently, values of Q_tot up to 0.95 ±0.17 were achieved, consistent with values of n_DT(0)τ_{E, dia}T_i (0) ≈ 8.7 × 10²⁰ m^-3·s·keV±20%. The ratio of fusion power to input power, Q_in, is in excess of 0.6. There are indications of an isotope effect on the edge pressure pedestal, but no net dependence of global confinement on isotopic plasma composition has been found.

Perturbative experiments on TFTR have investigated the transport of multiple ion species in reverse shear (RS) plasmas. The profile evolutions of trace tritium and helium and intrinsic carbon indicate the formation of core particle transport barriers in enhanced reverse shear (ERS) plasmas. There is an order of magnitude reduction in the particle diffusivity inside the RS region. The diffusivities for these species in ERS plasmas agree with neoclassical theory.

Compact toroid (CT) injection experiments with H mode plasmas were carried out for the first time in the JFT-2M tokamak. The soft X ray emission profile shows deep CT penetration into the core region of H mode plasmas heated by 1.2 MW NBI as well as in OH plasmas, with a toroidal magnetic field of 0.8 T. The line averaged electron density rapidly increased by Δ_e ≈ 0.2 × 10¹⁹ m^-3 at a rate of 4 × 10²¹ m^-3·s^-1 in H mode and the fuelling efficiency was roughly 20%. An asymmetric radial profile in the soft X ray emission was produced for ∼50 μs by deep penetration of the CT.

Particle confinement and transport have been systematically analysed for improved confinement modes in JT-60U. A scaling law for the total number of ions in the main plasma of ELMy H mode plasmas was proposed for the first time, with confinement times of the particles fuelled by NBI (centre fuelling) and by recycling and gas puffing (edge fuelling) separately defined. The confinement time increases and decreases with density for centre and edge fuelling, respectively. A comparison between the experimental data and the scaling was made for high β_p ELMy H mode and reversed shear plasmas. In the reversed shear plasma, particle confinement was enhanced by a factor of about 2 compared with the scaling, when the enhancement factor of the energy confinement time over the ELMy H mode scaling was larger than 1.2. Density controllability is discussed on the basis of the scaling in ELMy H mode, and it was found that additional fuelling with a confinement time of 0.1-0.4 s is necessary for density control in JT-60U. The particle diffusivity and convection velocity were evaluated by means of a perturbation technique using modulated helium gas puffing in ELMy H mode and reversed shear plasmas. The particle diffusivity and convection velocity were estimated to be 0.2-2 m²/s and from -4 (inward) to 0 m/s for the reversed shear plasma, and 0.4-2 m²/s and from -5.5 (inward) to +2.5 (outward) m/s for the ELMy H mode plasma, respectively. The particle diffusivity around the internal transport barrier was reduced by a factor of 5-6 compared with that in the inside and outside regions in the reversed shear plasma.

Fishbone-like burst modes (FBs) and toroidal Alfvén eigenmodes (TAEs) excited by energetic ions were observed for the first time in CHS heliotron/torsatron plasmas heated by co-injected neutral beams, where the rotational transform is increased by the induced beam driven current. FBs of m = 3/n = 2 (where m, n are the poloidal and toroidal mode numbers) induce a pulsed increase in the energetic ion loss flux. FBs of m = 2/n = 1 induce sawtooth oscillations in the latter half of a discharge where the plasma β becomes high. Only when the beam velocity exceeds about half of the central Alfvén velocity and the net plasma current is induced to the required level are TAEs with n = 1 and n = 2 excited; these are localized in the plasma core region, where the magnetic shear is appreciably reduced by the net plasma current. So far, TAE induced energetic ion loss has not been observed.

Both the dynamic and equilibrium thermal responses of an L mode plasma to repetitive ECH heat pulses were measured and compared with predictions from several thermal transport models. While no model consistently agreed with all observations, the GLF23 model was most consistent with the perturbed electron and ion temperature responses for one of the cases studied, which may indicate a key role played by electron modes in the core of these discharges. Generally, the IIF and MM models performed well for the perturbed electron response while the GLF23 and IFS/PPPL models agreed with the perturbed ion response for all three cases studied. No single model agreed well with the equilibrium temperature profiles measured.

A technique of fitting a modified hyperbolic tangent to the edge profiles has improved the localization of plasma edge parameters. Non-dimensional edge parameters are broadly consistent with several theories of the L-H transition that use edge gradients in their formulation of a critical threshold parameter. The ion ∇ B drift direction has only a small effect on the edge plasma conditions measured near the plasma midplane but a large effect on the divertor plasma. The dramatic change of power threshold with the ion ∇ B drift direction implies that phenomena in the divertor region may be critical for the L-H transition.

To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR supershots and high internal inductance plasmas. At NBI powers P_B ⩾ 30 MW, carbon influxes (`blooms') were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. A change in the radial electric field E_r is associated with the improved confinement observed.

A high ion temperature discharge in the lower hybrid current drive scheme has been obtained for the first time in the world in the TRIAM-1M superconducting tokamak. The high ion temperature mode is triggered by a transition which occurs within an operation window for the density, horizontal plasma position and antenna phasing. A steep ion temperature gradient (>80 keV/m) is formed near the core region at the transition. Long duration operation of this discharge has been successfully demonstrated in both the limiter and single null configurations by multiple control techniques.

Neoclassical tearing modes have been identified in JET ELMy H mode discharges at medium to high β. The n = 2 modes, as observed in both ECE and SXR data, have a clear island structure and a global character. The critical normalized β at the onset of the modes is found to scale with the normalized Larmor radius. The scaling with collisionality is much weaker and has a negative power. At high β, large amplitude n = 1 modes limit the maximum normalized β.

The behaviour of MHD perturbations before and during disruptions in TFTR reversed shear plasmas with q_min ≈ 2 was analysed. In the q_min region, tearing modes, wavelike modes, and mixed tearing plus wavelike modes are followed by disruption. Sometimes a helical snake (helix) appears at the X point of the q_min island. The local outward electron energy transport near the X point can be explained by the development of `positive' magnetic islands (islands with positive current perturbations). It is proposed that the disruption is initiated when the X point of the magnetic islands coincides in one toroidal position near the torus equator.

Two approaches to achieving long timescale stabilization of the ideal kink mode with a real, finite conductivity wall are considered: plasma rotation and active feedback control. DIII-D experiments have demonstrated stabilization of the resistive wall mode (RWM) by sustaining β greater than the no-wall limit for up to 200 ms, much longer than the wall penetration time of a few milliseconds. These plasmas are typically terminated by an m = 3, n = 1 mode as the plasma rotation slows below a few kilohertz. Recent temperature profile data show an ideal MHD mode structure, as expected for the RWM at β above the no-wall limit. The critical rotation rate for stabilization is in qualitative agreement with recent theories for dissipative stabilization in the absence of magnetic islands. However, drag by small amplitude RWMs or damping of stable RWMs may contribute to an observed slowing of rotation at high β, rendering rotational stabilization more difficult. An initial open loop active control experiment, using non-axisymmetric external coils and a new array of saddle loop detectors, has yielded encouraging results indicating a delayed onset of the RWM.

The geometry effects of the W shaped divertor on the divertor plasma were investigated quantitatively. The ion flux was increased near the divertor strike point, which is effective for reducing the local electron temperature and decreasing the onset _e of divertor detachment. The plasma profile and parallel plasma flow in the scrape-off layer were systematically measured using reciprocating Mach probes installed at the midplane and the divertor X point. For the ion ∇B drift direction towards the divertor, `flow reversal' was observed at the midplane. A quantitative evaluation of the parallel plasma flow suggesting that the flow is produced in a torus to keep the pressure constant along the field lines was consistent with the measurements.

Radiofrequency waves in the ion cyclotron range of frequency (ICRF) are mainly used for plasma production and heating in the central cell of the GAMMA 10 tandem mirror. GAMMA 10 has minimum-B anchor cells with a non-axisymmetric magnetic field configuration. The ICRF heating system in the central cell has been improved to create a more axisymmetric plasma. A high ion temperature is attained with the system, and high energy ions with energies of more than 50 keV are detected both parallel and perpendicular to the magnetic field lines. Strong temperature anisotropy is observed and strong Alfvén ion cyclotron (AIC) modes are excited due to the anisotropy. With the AIC modes, the number of high energy ions detected at the end of the mirror increase and the number of high energy ions with a pitch angle in the central cell of nearly 90° decrease.

A novel high β relaxation to a field reversed configuration (FRC) has been investigated by axially colliding two spheromaks with opposing toroidal magnetic fields. The β value of the merging toroids increases from 0.1 to 0.7-1.0 within 15 μs, indicating an equilibrium transition from the low β spheromak to the high β FRC. An important finding is that the merging spheromaks relax either to a high β FRC or to another low β spheromak, depending on whether the initial normalized magnetic helicity given to these spheromaks is smaller or larger than a threshold value. This fact suggests that the FRCs are equipped with some global stability as robust as the Taylor magnetic energy minimum state.

A new concept for plasma heating using axial magnetic compression of a field reversed configuration (FRC) plasma is proposed. In this concept, the FRC plasma is compressed only axially, keeping the magnetic flux between the separatrix and the confining chamber (flux conserver) wall unchanged, while allowing the plasma to expand radially. A simple model based on an empirical scaling law of FRC confinement and on the assumption that the compression is done adiabatically predicts that, in addition to heating the plasma, improved confinement will also be accomplished with this concept. This compression is done by energizing segmented mirror coils successively in such a way as to decrease the length of the confinement region between the coils. The apparatus for this axial compression was developed and an experiment was carried out. In this experiment the plasma was compressed by about 30% and the plasma lifetime of about 500 μs was increased by about 50 μs.

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D₂ puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D₂ puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to n_e/n_Gr ≈ 0.25 (with n_Gr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.

Ion cyclotron resonance heating experiments have been carried out in reactor relevant scenarios during the deuterium-tritium (DTE1) campaign in JET. For the first time it was possible to assess the scheme with a deuterium minority in tritium plasmas: this produced a steady state DT fusion power of up to 1.66 MW for input ICRF power of 6 MW. Strong ion heating, with a core ion temperature of up to 13 keV, was observed when using ³He minority heating in both 50:50 DT and tritium dominated H mode plasmas. Second harmonic tritium heating was also studied and, as expected for JET plasma conditions, produced mainly electron heating. Finally, the `inverted' scenario of tritium minority in deuterium was successfully demonstrated.

The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. The article identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems.

The remote maintenance strategy and the associated component design of the International Thermonuclear Experimental Reactor (ITER) have reached a high degree of completeness, especially with respect to those components that are expected to require frequent or occasional remote maintenance. Large scale test stands, to demonstrate the feasibility in principle of the remote maintenance procedures and to develop the required equipment and tools, were operational at the end of the Engineering Design Activities phase. The initial results are highly encouraging: major remote equipment deployment and component replacement operations have been successfully demonstrated.

An assessment is presented of the impact of recent magnetohydrodynamic research results on performance projections for reactor scale tokamaks as exemplified by the ITER Final Design Report (ITER/FDR) facility. For nominal ELMy H mode operation, the presence and amplitude of neoclassical tearing modes governs the achievable β value. Recent work finds that the scaling of β at which such modes onset agrees well with a polarization drift model, with the consequence that, with reasonable assumptions regarding seed island width, the mode onset β will be lower in reactor scale tokamaks than in contemporary devices. Confinement degradation by such modes, on the other hand, depends on relative saturated island size which is governed principally by β and secondarily by ν* effects on bootstrap current density. Relative saturated island size should be comparable in present and reactor devices. DT ITER demonstration discharges in JET exhibited no confinement degradation at the planned ITER operating value of β_N = 2.2. Theory indicates that electron cyclotron current drive can either stabilize these modes or appreciably reduce saturated island size. Turning to operation in candidate steady state, reverse shear, high bootstrap fraction configurations, wall stabilization of external kink modes is effective while the plasma is rotating but (so far) rotation has not been maintained. Recent error field observations in JET imply an error field size scaling that leads to a projection that the ITER/FDR facility will be somewhat more tolerant to error fields than thought previously. ICRF experiments on JET and Alcator C-Mod indicate that plasmas heated by central energetic particles have benign ELMs compared with the usual type 1 ELM of NBI heated discharges.

It has long been recognized that attainment of the safety and environmental potential of fusion energy requires the successful development of low activation materials for the first wall, blanket and other high heat flux structural components. Only a limited number of materials potentially possess the physical, mechanical and low activation characteristics required for this application. The current US structural materials research effort is focused on three candidate materials: advanced ferritic steels, vanadium alloys and silicon carbide composites. Recent progress has been made in understanding the response of these materials to neutron irradiation.

The work presented deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts are usually observed over a broad range of time and spatial scales. The existence of these fronts provides one possible way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyro-Bohm type in spite of these large scale transport events. Some departure from the gyro-Bohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux as for a turbulence calculated at fixed temperature gradient, with the same time averaged profile.

Non-linear MHD simulation results of pellet injection show that MHD forces can accelerate large pellets injected on the high field side of a tokamak to the plasma centre. Magnetic reconnection can produce a reverse shear q profile. Ballooning instability caused by pellets is also reduced by high field side injection. Studies are also reported of the current quench phase of disruptions, which can cause 3-D halo currents and runaway electrons.

For tokamaks the ideal localized interchange mode does not play a role for cases of q (safety factor) >1 over the whole plasma region, whereas in heliotron plasmas the mode becomes unstable in the edge region owing to the magnetic hill. However, for negative shear tokamaks, the resistive localized interchange mode becomes unstable where q₀ is much larger than q_min, where q₀ (q_min) is the central (minimum) q value. With the appropriate ellipticity and triangularity the stability against the resistive interchange mode is significantly improved. It is also shown that ideal and resistive non-resonant global interchange modes appear easily in the central region of heliotrons and in negative shear tokamaks; the modes are similar to the infernal mode in low shear tokamaks.

The process of kinetic stabilization of the tilt disruption in a field reversed configuration is investigated by means of a three dimensional particle simulation. For the case of no ion beam the growth rate of the tilt instability decreases as the plasma β value at the magnetic separatrix, β_sp, increases. This stabilization effect originates from the `anchoring ions' which exist in the vicinity of the magnetic separatrix and act as an `anchor' to hold the internal plasma to the external plasma. The tilt mode is also found to be stabilized by injecting an ion beam with about 20% of the ion thermal energy in the vicinity of the null point even for small β_sp plasmas.

A canonical profiles transport model and a semi-empirical transport model are tested against JT-60U plasmas. The simulations were carried out with the 1.5-D transport code ASTRA, comprising the particle transport, and the electron and ion heat transport.

Relying on the good agreement observed between the gyrokinetic PENN model and the low toroidal mode number n damping measurements from JET, the stability of Alfvén eigenmodes (AEs) is predicted for reactor relevant conditions. Full non-local wave-particle power transfers are computed between global wavefields and alpha particles in an ITER reference equilibrium, showing that low-n eigenmodes (n ≅ 2) are strongly damped and intermediate-n eigenmodes (n ≅ 12) with a global radial extension are stable with a damping rate γ/ω ≅ 0.02. Even though an excitation of alpha particle driven instabilities remains in principle possible in reactors, this study suggests that realistic operation scenarios exist where all the AEs of global character are stable.

It is shown that tokamaks can be intrinsically steady state without seed currents by coupling potato bootstrap current in the region close to the magnetic axis to the banana bootstrap current away from the axis. The equilibria found are highly elongated (e.g. κ = 3) and have naturally reversed shear profiles. The vertical instability of elongated tokamaks and the theory of the enhanced reversed shear mode for this class of equilibria are also developed.

Experimental observations suggest that toroidal plasma rotation can play a crucial role in determining the stability of a tokamak. Conventional MHD stability analysis ignores rotation. Results from analytic and numerical analysis of a number of tokamak stability issues including toroidal plasma rotation are reported. Depending on the plasma and rotation profiles, the net effect of rotation can be either stabilizing or destabilizing; it also has a substantial effect on the observed mode structure. Localized interchange, the double kink mode, low-n resistive MHD modes and the effect of sheared rotation on magnetic island structure are discussed.

The neoclassical behaviour of plasma rotation under a driven radial electrical current is studied in a tokamak geometry. An ambipolar radial electric field develops instantly in such a way that the driven current is balanced by a return current j^p in the plasma. The j^p × B torque pushes the plasma into a new rotation state both toroidally and poloidally. An anomalous toroidal viscosity is needed to avoid an extreme toroidal rotation speed. It is shown that the poloidal rotation relaxes to a new equilibrium speed, which is in general smaller than the E × B poloidal speed, and that the timescale for the relaxation of poloidal rotation is the same as that of toroidal rotation generation, which is usually much longer than the ion-ion collision time.

Kinetic or slow Alfvén waves, which appear in tokamak plasmas as a result of RF wave conversion, have small radial localization. For this reason, these waves are used here to create strongly sheared plasma flows at any point of the plasma minor radius and to obtain the locally reversed magnetic shear configuration. It is also demonstrated that Alfvén waves may be a convenient trigger for formation of transport barriers in tokamaks. They can be used for suppressing plasma turbulence and squeezing ion banana and potato orbits. Estimates of the absorbed Alfvén wave power, which is necessary to achieve these effects in some tokamaks, are about the level of the absorbed power in early completed experiments. The use of Alfvén waves can also be helpful for self-organization processes in tokamaks.

An equation which includes a non-local effect in the heat flux is introduced to study transient transport phenomena. A non-local heat flux, which is expressed in terms of an integral equation, is superimposed on the conventional form of the heat flux. This model is applied to describe the fast responses in the transition from L mode to H mode. A small fraction of non-local component in the heat flux is found to be very effective in modifying the response against an L-H transition. The transient features of the transport property, which are observed in the response of heat pulse propagation, are qualitatively reproduced by the transport simulations based on this model.