Improved Core Confinement Research Articles

Isotope physics of heat and particle transport with tritium in JET-ILW type-I ELMy H-mode plasmas

As part the DTE2 campaign in the JET tokamak, we conducted a parameter scan in T and D-T complementing existing pulses in H and D. For the different main ion masses, type-I ELMy H-modes at fixed plasma current and magnetic field can have the pedestal pressure varying by a factor of 4 and the total pressure changing from βN=1.0 to 3.0. We investigated the pedestal and core isotope mass dependencies using this extensive data set. The pedestal shows a strong mass dependence on the density, which influences the core due to the strong coupling between both plasma regions. To better understand the causes for the observed isotope mass dependence in the pedestal, we analysed the interplay between heat and particle transport and the edge localised mode (ELM) stability. For this purpose, we developed a dynamic ELM cycle model with basic transport assumptions and a realistic neutral penetration. The temporal evolution and resulting ELM frequency introduce an additional experimental constraint that conventional quasi-stationary transport analysis cannot provide. Our model shows that a mass dependence in the ELM stability or in the transport alone cannot explain the observations. One requires a mass dependence in the ELM stability as well as one in the particle sources. The core confinement time increases with pedestal pressure for all isotope masses due to profile stiffness and electromagnetic turbulence stabilisation. Interestingly, T and D-T plasmas show an improved core confinement time compared to H and D plasmas even for matched pedestal pressures. For T, this improvement is largely due to the unique pedestal composition of higher densities and lower temperatures than H and D. With a reduced gyroBohm factor at lower temperatures, more turbulent drive in the form of steeper gradients is required to transport the same amount of heat. This picture is supported by quasilinear flux-driven modelling using TGLF-SAT2 within Astra. With the experimental boundary condition TGLF-SAT2 predicts the core profiles well for gyroBohm heat fluxes >15 , however, overestimates the heat and particle transport closer to the turbulent threshold.

Revisiting the improved core confinement simulations for FT‐2 tokamak

AbstractIn the present paper, we revisit observations performed in FT‐2 tokamak from previous works. Improvements of core confinement are observed and believed to be caused by wide orbits going from collisionless to collisional regimes. Similar phenomena can occur whenever gradient lengths are comparable to the orbit widths at the top of the pedestal and the loss cone is continuously and increasingly filled by heated particles, collisions and turbulent effects. The lower hybrid heating operator is introduced into the ELMFIRE code to increase the ion temperature during the simulations while keeping the edge temperature low with logical boundary condition at the limiter. Particular focus is given on how the radial electric field deviates from the neoclassical value while introducing turbulent effects.

Overview of the isotope effects in the ASDEX Upgrade tokamak

In recent years, measurements on the ASDEX Upgrade tokamak and modelling performed for plasmas with hydrogen (H) and deuterium (D) as the main gas have improved our understanding of the ion mass dependencies in fusion plasmas. The observed isotope effects can be explained with established physics processes which highlight the importance of treating heat transport with coupled electron and ion heat channels. In the core of electron heated L-mode plasmas, the mass dependence of the electron–ion equipartition results in a reduction of with increasing ion mass. Combined with higher profile stiffness in the ions compared to the electrons, this results in improved core confinement for higher ion masses. At the edge of L-mode plasmas where a higher collisionality is observed, parallel electron dynamics is fundamental for turbulence. The parallel electron dynamics term in the gyrokinetic equations directly depends on , resulting in a different kinetic response with different ion mass. Higher turbulent fluxes are expected with lower ion mass. This is consistent with the difference in observed in the experiment. The mass dependence of turbulent transport in the L-mode edge has direct consequences for the L–H transition. More heating power is required to enter the H-mode at lower mass (). This is expected if the critical E × B shearing rate is important for the transition from L to H mode. In the H-mode pedestal, remains important to regulate the turbulent transport. The electrons do not contribute to and the enhanced equipartition for lower ion masses causes a shift from the ion channel to the electron channel in the absolute heat fluxes. Consequently, the inter edge localised mode (ELM) transport is found to be higher with lower isotope mass. This enhanced transport in H can prevent the pedestal from reaching the peeling–ballooning stability boundary with engineering parameters where D plasmas are peeling–ballooning unstable. Increasing the triangularity reduces the inter ELM transport in H stronger than in comparable D plasmas. For matched pedestal top and matched heat sources, the core heat transport is found to be similar for H and D when the fast-ion content is low. When ion temperature gradient turbulence stabilisation by fast ions becomes relevant, the mass dependent fast-ion slowing down results in higher fast-ion content in D and therefore in a reduction of ion heat transport in the core. Then, even for matched pedestals .

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

Integration of divertor detachment with a high-performance (βN ∼ 3, βp > 2, H98 ∼ 1.5) core plasma has been demonstrated in DIII-D high-βp (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-βp plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te < 5 eV), low particle flux, and low heat flux across the entire divertor target plate were obtained by using nitrogen seeding. This detached high-βp plasma is compatible with a newly developed detachment control system which can help optimize the nitrogen gas flow rate. Several features, i.e., the high edge safety factor in the high-βp scenario, impurity injection, closed divertor and reduced heating power requirement due to the high confinement, facilitate the achievement of full divertor detachment at lower density. Instead of degrading global performance, the divertor detachment facilitates the access to an even stronger ITB at large radius with a relatively weak ETB through self-organized synergy between ITB and ETB, leading to sustained high confinement. The strengthening of the large-radius ITB compensates for the ETB degradation associated with divertor detachment. In addition, a weak ETB naturally has smaller edge localized modes (ELMs). In particular, with neon injection, a long-period no-ELM H-mode phase has been achieved simultaneously with high-performance core and partially detached divertor plasmas. These results demonstrate the possibility of integrating excellent core plasma performance with an effective divertor solution, an essential step toward steady-state operation of reactor-grade plasmas.

Integration of full divertor detachment with improved core confinement for tokamak fusion plasmas

Divertor detachment offers a promising solution to the challenge of plasma-wall interactions for steady-state operation of fusion reactors. Here, we demonstrate the excellent compatibility of actively controlled full divertor detachment with a high-performance (βN ~ 3, H98 ~ 1.5) core plasma, using high-βp (poloidal beta, βp > 2) scenario characterized by a sustained core internal transport barrier (ITB) and a modest edge transport barrier (ETB) in DIII-D tokamak. The high-βp high-confinement scenario facilitates divertor detachment which, in turn, promotes the development of an even stronger ITB at large radius with a weaker ETB. This self-organized synergy between ITB and ETB, leads to a net gain in energy confinement, in contrast to the net confinement loss caused by divertor detachment in standard H-modes. These results show the potential of integrating excellent core plasma performance with an efficient divertor solution, an essential step towards steady-state operation of reactor-grade plasmas.

Formation of the internal transport barrier in KSTAR

One of key objectives of tokamak experiments is the exploration of enhanced confinement regimes, and the access of the internal transport barrier (ITB) formation is dealt with an important physics issue in the most of major tokamaks. Also, the advanced tokamak scenario with ITB is expected to lead to a continuous reactor with high fusion power density. From that point of view, the formation of the ITB in KSTAR which is designed for long pulse operation capability is very important although its heating and current drive systems are not fully equipped yet. We have therefore assumed that an early injection of the full NBI power (∼5.5 MW) during the current ramp-up would give a chance to form an internal barrier if the plasma could stay in the L-mode. To avoid the H-mode transition, we have produced inboard limited plasmas with detaching from the both upper and lower divertors. Using this approach, an ITB formation during L-mode has been observed which shows improved core confinement. Ion and electron temperature profiles show the barrier clearly in the temperature, and it was sustained for about 7 s in the dedicated experiment. This is the first stationary ITB observed in a full superconducting tokamak. This operation scenario with the ITB could be an alternative way to achieve a high performance regime in KSTAR, and the length of the ITB discharge could be extended even longer. In this paper, we present the formation of the ITB using measured and simulated characteristic profiles.

Near-zero dispersion flattened, low-loss porous-core waveguide design for terahertz signal transmission

We demonstrate a photonic crystal fiber with near-zero flattened dispersion, ultralower effective material loss (EML), and negligible confinement loss for a broad spectrum range. The use of cyclic olefin copolymer Topas with improved core confinement significantly reduces the loss characteristics and the use of higher air filling fraction results in flat dispersion characteristics. The properties such as dispersion, EML, confinement loss, modal effective area, and single-mode operation of the fiber have been investigated using the full-vector finite element method with the perfectly matched layer absorbing boundary conditions. The practical implementation of the proposed fiber is achievable with existing fabrication techniques as only circular-shaped air holes have been used to design the waveguide. Thus, it is expected that the proposed terahertz waveguide can potentially be used for flexible and efficient transmission of terahertz waves.

Transition into the improved core confinement mode as a possible mechanism for additional electron heating observed in the lower hybrid current drive experiments at the FT-2 tokamak

In experiments on lower hybrid current drive (LHCD) carried out at the FT-2 tokamak, a substantial increase in the central electron temperature T e (r = 0 cm) from 550 to 700 eV was observed. A complex simulation procedure is used to explain a fairly high LHCD efficiency and the observed additional heating, which can be attributed to a transition into the improved core confinement (ICC) mode. For numerical simulations, data obtained in experiments with deuterium plasma at 〈n e 〉 = 1.6 × 1019 m–3 were used. Simulations by the GRILL3D, FRTC, and ASTRA codes have shown that the increase in the density and central temperature is apparently caused by a significant suppression of heat transport in the electron component. The mechanism for transition into the improved confinement mode at r < 3 cm can be associated with the broadening of the plasma current channel due to the lower hybrid drive of the current carried by superthermal and runaway electrons. In this case, the magnetic shear s = (r/q)(dq/dr) in the axial region of the plasma column almost vanishes during the RF pulse. In this study, the effect of lower hybrid waves on the plasma parameters, resulting in a transition into the ICC mode, is considered. New experimental and calculated data are presented that evidence in favor of such a transition. Special attention is paid to the existence of a threshold for the transition into the ICC mode in deuterium plasma.

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ∼ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ∼ 1.0 only for βN = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to ne/nGW ∼ 0.5, long pulse operation (&amp;gt;1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (&amp;gt;50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (Tped ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.

Low-recycling conditions and improved core confinement in steady-state operation scenarios in JET (Joint European Torus)

In this paper, we discuss the phenomena that link particle recycling from the vessel walls in the L-mode during discharge start-up and the core confinement in the H-mode during the subsequent main heating phase. We consider available data of JET experiments that aimed at approaching fully non-inductive ITER-relevant steady-state conditions and show that the high electron temperature produced at the edge by a low recycling during start-up tends to favour the build-up of high normalized β(βN) regimes in the H-mode, the confinement being improved in a large plasma volume. To provide an insight into this complex phenomenon we have modelled the relation between particle recycling in the scrape-off layer and the evolution of plasma transport, plasma current density and shear as well as the stability properties for such experimental conditions. The results confirm the existence of a link between the confinement in the H-mode phase and the values at the edge of electron temperature, bootstrap current density and local magnetic shear during start-up. Such a link could favour these regimes to be self-sustained in time.

Chapter 2: Plasma confinement and transport

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137–2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

Intermittent β collapse after NBCD turn-off in JT-60U fully non-inductive reversed shear discharges

Non-disruptive β collapses with a regular intermittency have been observed after a forced turn-off of neutral beam current drive (NBCD) in JT-60U fully non-inductive, reversed shear (RS) discharges. Self-consistent transport simulations with improved core confinement have first clarified that the redistribution of ‘return current’ induced after the NBCD turn-off lowers the safety factor of magnetic shear reversal, qmin. Linear MHD stability analysis has also revealed that this lowering of safety factor leads to n = 1 kink-ballooning instability with the modes localized around the internal transport barrier (ITB) of higher qmin (∼6) discharges. It was also pointed out that an increase in the bootstrap current under continuous NB heating can lead to ITB reconstruction and thus causes subsequent β collapses.

Access to sustained high-beta with internal transport barrier and negative central magnetic shear in DIII-D

High values of normalized β (βN∼4) and safety factor (qmin∼2) have been sustained simultaneously for ∼2s in DIII-D [J.L. Luxon, Nucl. Fusion 42, 64 (2002)], suggesting a possible path to high fusion performance, steady-state tokamak scenarios with a large fraction of bootstrap current. The combination of internal transport barrier and negative central magnetic shear at high β results in high confinement (H89P&amp;gt;2.5) and large bootstrap current fraction (fBS&amp;gt;60%) with good alignment. Previously, stability limits in plasmas with core transport barriers have been observed at moderate values of βN (&amp;lt;3) because of the pressure peaking which normally develops from improved core confinement. In recent DIII-D experiments, the internal transport barrier is clearly observed in the electron density and in the ion temperature and rotation profiles at ρ∼0.5 but not in the electron temperature profile, which is very broad. The misalignment of Ti and Te gradients may help to avoid a large local pressure gradient. Furthermore, at low internal inductance ∼0.6, the current density gradients are close to the vessel and the ideal kink modes are strongly wall-coupled. Simultaneous feedback control of both external and internal sets of n=1 magnetic coils was used to maintain optimal error field correction and resistive wall mode stabilization, allowing operation above the free-boundary β limit. Large particle orbits at high safety factor in the core help to broaden both the pressure and the beam-driven current profiles, favorable for steady-state operation. At plasma current flat top and β∼5%, a noninductive current fraction of ∼100% has been observed. Stability modeling shows the possibility for operation up to the ideal-wall limit at β∼6%.

Transition phenomena and thermal transport properties in LHD plasmas with an electron internal transport barrier

Two types of improved core confinement were observed during centrally focused electron cyclotron heating (ECH) into plasmas sustained by counter (CNTR) and Co neutral beam injections (NBI) in the Large Helical Device. The CNTR NBI plasma displayed transition phenomena to the high-electron-temperature state and had a clear electron internal transport barrier, while the Co NBI plasma did not show a clear transition or an ECH power threshold but showed broad high temperature profiles with moderate temperature gradient. This indicated that the Co NBI plasma with additional ECH also had an improved core confinement. The electron heat transport characteristics of these plasmas were directly investigated using heat pulse propagation excited by modulated ECH. These effects appear to be related to the m/n = 2/1 rational surface or the island induced by NBI beam-driven current.

Transport modelling and gyrokinetic analysis of advanced high performance discharges

Predictive transport modelling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and advanced tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. E × B shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET and AUG tokamaks. GLF23 transport modelling and gyrokinetic stability analysis indicate that E × B shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of E × B shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveal some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and E × B shear stabilization can dominate parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent E × B shear quenching of the turbulent transport, and only E × B suppression is achieved.

Ion Cyclotron Resonance Heating on TEXTOR

The original work done on TEXTOR on ion cyclotron resonance heating of the plasma is reviewed. After a brief introduction outlining the principles, the radio-frequency (rf) system is described, with its substantial evolution during time. Then, the different physics results are reviewed. Ion cyclotron heating has been performed in a large number of scenarios and under a wide range of conditions. Aside from the various minority or mode-conversion scenarios, the interaction with beam ions and the possibility of controlling fast-ion diffusion with rf have been deeply investigated. Both the interaction with the wall or edge plasma and the impact on improved core confinement were studied in detail. Pioneering work was done to demonstrate efficient heating with unshielded antennas and plasma production in a tokamak by rf alone for plasma startup assistance or wall conditioning.

High recycling steady H-mode regime in the JFT-2M tokamak

A new operating regime, ‘high recycling steady’ (HRS) H-mode regime, has been discovered in the JFT-2M tokamak. It is easily reproduced with neutral beam heating of any co-, balanced-, or counter-injection under wall fuelling from the boronized first wall. The HRS is characterized by good energy confinement (H89P ∼ 1.6) at a high density (ne/nGW ∼ 0.7; nGW is the Greenwald density), low radiation loss power fraction (, typically), and the complete disappearance of large ELMs. Accompanying the HRS H-mode transition, the coherent magnetic fluctuations in the frequency range of the order of 10–100 kHz with significant variation are seen on magnetic probes at the vessel wall. In addition, the HRS H-mode edge condition is compatible with an improved core confinement. We have demonstrated that an internal transport barrier can be produced under a HRS H-mode edge condition, achieving βNH89P ∼ 6 at ne/nGW ∼ 0.7–1.0, transiently.

Status of and prospects for advanced tokamak regimes from multi-machine comparisons using the ‘International Tokamak Physics Activity’ database

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses (‘hybrid’ scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, , is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

Observation of High Recycling Steady H-mode edge and compatibility with improved core confinement mode on JFT-2M

A new operational regime has been discovered on JFT-2M under the boronized first wall condition to produce the High Recycling Steady (HRS) H-mode, which is characterized by good energy confinement (H89P ∼ 1.6) at a high density of around 70% of the Greenwald density (ne/nGW), low radiated power fraction, and complete disappearance of large ELMs. Accompanying the HRS H-mode transition, coherent magnetic fluctuation is seen on magnetic probes at the vessel wall. These coherent fluctuations have a frequency of the order of 10–100 kHz with significant variation. They are recognized as important in enhancing particle transport, whose characteristics are similar to the Enhanced Dα (EDA) H-mode regime reported from Alcator C-Mod. The H′-mode, previously observed on JFT-2M, had features in common with the EDA mode in terms of the coherent density fluctuation, but it appeared transiently. The HRS operating regime is also similar to the EDA mode, except that HRS is seen even at low q95 (2 < q95 ⩽ 3). The most important feature of the HRS H-mode edge condition is compatibility with an improved core confinement mode at a high density without large ELMs. We have demonstrated that an internal transport barrier can be produced under a HRS H-mode edge condition, achieving βNH89P ∼ 6.2 at ne/nGW ∼ 70%, transiently.

Dynamics of the transport barrier formation on the FT-2 tokamak caused by lower hybrid heating

Experiments at the FT-2 tokamak had demonstrated effective plasma LH heating, which was accounted for by both direct absorption of RF power and plasma transport suppression. The improved core confinement accompanied by Internal Transport Barrier (ITB) formation was observed. The RF pulse switch off is followed by triggering of LH transition and the External Transport Barrier (ETB) formation near the last closed flux surface. The present paper is devoted to a much more detailed study of the radial electric fieldEr behaviour in the region of ITB and ETB and its influence on the tokamak microturbulence in these regions. The new experimental data were obtained by spatial spectroscopic technique using additional pulse helium puffing in hydrogen plasma. Simultaneously microscale plasma oscillations in the frequency band (0.01–2) MHz are observed with local enhanced microwave scattering diagnostics and by x-mode fluctuation reflectometry. Experiments demonstrate that the improved confinement is associated with the modification of microturbulence by the shear of theE×B poloidal velocity. This conclusion is also confirmed by the data obtained by Langmuir probes in the edge plasma.