Strong Internal Transport Barriers Research Articles

Impact of T i/T e ratio on ion transport based on EAST H-mode plasmas

At the EAST tokamak, the ion temperature (T i) is observed to be clamped around 1.25 keV in electron cyclotron resonance (ECR)-heated plasmas, even at core electron temperatures up to 10 keV (depending on the ECR heating power and the plasma density). This clamping results from the lack of direct ion heating and high levels of turbulence-driven transport. Turbulent transport analysis shows that trapped electron mode and electron temperature gradient-driven modes are the most unstable modes in the core of ECR-heated H-mode plasmas. Nevertheless, recently it was found that the T i/T e ratio can increase further with the fraction of the neutral beam injection (NBI) power, which leads to a higher core ion temperature (T i0). In NBI heating-dominant H-mode plasmas, the ion temperature gradient-driven modes become the most unstable modes. Furthermore, a strong and broad internal transport barrier (ITB) can form at the plasma core in high-power NBI-heated H-mode plasmas when the T i/T e ratio approaches ~1, which results in steep core T e and T i profiles, as well as a peaked n e profile. Power balance analysis shows a weaker T e profile stiffness after the formation of ITBs in the core plasma region, where T i clamping is broken, and the core T i can increase further above 2 keV, which is 80% higher than the value of T i clamping in ECR-heated plasmas. This finding proposes a possible solution to the problem of T i clamping on EAST and demonstrates an advanced operational regime with the formation of a strong and broad ITB for future fusion plasmas dominated by electron heating.

Recent results of fusion triple product on EAST tokamak

High fusion triple product has been obtained in the advanced scenarios with high normalized beta (β N) on the Experimental Advanced Superconducting Tokamak (EAST). A record value of n i0 T i0 τ E ∼ 1.0 × 1019 m−3 keV s for EAST deuterium plasma has been achieved, which is due to the formation of strong and broad internal transport barriers (ITBs) in n e, T e and T i profiles. Analysis shows that the strong ITB formation could be attributed to the reduction of transport from ITG modes. Based on the analysis, the physical mechanisms and methods to further improve the plasma performance are discussed.

Prediction of high-performance scenario with localized magnetic shear reversal on EAST tokamak

An advanced scenario with high poloidal beta, large-radius internal transport barriers (ITBs) and localized reversed-shear q-profile, has been demonstrated in DIII-D/EAST joint experiments recently on the DIII-D tokamak. This scenario is also one of the future explorations on EAST. In this article, integrated modeling has been utilized to explore the regime with localized reversed-shear q-profile and high performance based on the developed long-pulse scenario on EAST. A scenario with a strong reversed-shear q-profile has been achieved with the combination of 2 MW off-axis ECH, 2.6 MW LHW and 2 MW NBI (H98y2 ∼ 1.35, ∼ 2.56, ∼ 2.23, ). In addition, a small amount (∼0.1 MW) of centrally deposited ECH facilitates the generation of the bootstrap current on or near the axis, and hence reduces the level of the q-profile in this region; thus, a scenario with localized magnetic shear reversal and strong large-radius ITBs is obtained (H98y2 ∼ 1.4, ∼ 2.66, ∼ 2.31, ). Modeling predictions indicate that off-axis ECH/LHW and high plasma density promote the formation of large-radius ITBs. A further increase in the injected power of centrally deposited ECH (0.15 MW) or NBI (4 MW) leads to an increase in the temperature gradient, especially the electron temperature near the axis, and large-radius ITBs disappear in all channels (T e, T i, n e); also, the confinement quality decreases. The bootstrap current density profiles and LHW-driven current profiles tend to peak near the axis due to the increase in electron temperature profiles, which eventually leads to peaked total current profiles, similar to the EAST long-pulse scenario. Changes in magnetic shear result in the disappearance of large-radius ITBs. Modeling validation will be carried out in the near future on the EAST tokamak.

Dynamics of weak-magnetic-shear-sustained internal transport barrier during supersonic molecular-beam injection in JT-60U

The characteristics of ion heat transport inside the internal transport barrier (ITB), which is sustained by weak magnetic shear, are investigated using a cold pulse induced by supersonic molecular-beam injection (SMBI) in JT-60U. It is known that cold-pulse propagation in ITBs usually significantly reduces the temperature and degrades the temperature gradient. When subsequent SMBIs are launched before the temperature has recovered, it is observed that the temperature gradient of the ITB is not monotonically decreased but alternately decreased and increased. Alternating decreasing and increasing phases of the ITB temperature gradient continue for about 1 s (∼6τ E), and the properties of the cold-pulse propagation and the flux-gradient relations differ according to the phase. The usual transient transport analysis is also provided by inducing a cold pulse in stationary ITBs. A rapid reduction and recovery of the temperature is observed inside the ITBs, and it is found that the ion-heat flux changes without a variation in the local ion-temperature gradient or a change to the other observable local parameters. The flux-gradient relations exhibit significant hysteresis in two ITB cases: (i) ITBs with strong electron-density gradients and (ii) ITBs with weak ion-temperature gradients. In addition, the range in which hysteresis appears is most likely to depend on the width of the ITB.

Effects of triangularity on the fusion performance of CFETR

An important task of the China Fusion Engineering Test Reactor physics design is to develop operation scenarios with high fusion power (1 GW), high bootstrap current fraction for steady-state and a plasma edge compatible with heat and particle exhaust. To achieve these goals, triangularity (δ) effects on the fusion performance of two candidate scenarios, with or without reversed magnetic shear (RS), namely conventional H-mode and RS H-mode, are evaluated using core-edge coupled integrated modeling in this paper. For fixed pedestal density, it is shown that higher δ is favorable for higher fusion performance in the conventional H-mode scenario while the fusion performance decreases with increasing δ in the RS H-mode scenario. In conventional H-mode, the higher fusion performance at high δ mainly comes from a higher pedestal temperature as predicted by EPED in combination with stiff core kinetic profiles. In the RS H-mode scenario with a local reversed shear region, the profiles are non-stiff and a strong internal transport barrier (ITB) exists at low δ. This results in higher density and temperature inside the ITB for low δ, leading to higher fusion power. If the pedestal temperature is kept fixed, in both scenarios the significant increase in pedestal density, which extends into the core, dominates at high δ and leads to much higher fusion power. For conventional H-mode, destabilization from increasing δ is partially balanced by stabilization due to increasing ν*. Since the normalized heat sources are quite similar, it results in minimal changes in the temperature profiles except for the lowest density case. For RS H-mode, destabilization from increasing δ is approximately balanced by stabilization due to increasing ν* in foot region, but a strong temperature ITB is still evident for low δ. The ability to take advantage of the high pedestal density in conventional H-mode and reversed shear scenario depends on its compatibility with edge density requirements from efficient heat and particle exhaust. Transport analysis is presented to elucidate the roles of δ, collisionality and magnetic shear in altering the profiles and the ITB, which contribute to the different behavior in the two scenarios.

Scenario development for high βp low torque plasma with qmin above 2 and large-radius internal transport barrier in DIII-D

A recent experiment on DIII-D, which was conducted by the joint research team from DIII-D and EAST, has extended the previous high , high qmin regime, which has been tested in the 2013 DIII-D/EAST joint experiment, to inductive operation at higher plasma current ( MA) and significantly higher normalized fusion performance (). The experiment aims at exploring high performance scenario with and reduced torque for long pulse operation, which can be potentially extrapolated to EAST. The effort was largely motivated by the interest in developing a feasible scenario for long-pulse high performance operation with low torque on EAST. Very high confinement, H89 = 3.5 or with , has been achieved transiently in this experiment together with and reduced NBI torque ( N m). The excellent confinement is associated with the spontaneous formation of an internal transport barrier (ITB) in plasmas with MA at large minor radius (normalized ) in all channels (ne, Te, Ti, , especially strong in the Te channel). Fluctuation measurements show a significant reduction in the fluctuation levels, including AE modes and broadband turbulence, at the location where an ITB forms. Linear gyrokinetic simulations also support the decrease of the growth rate of the most unstable mode during strong ITB formation. The simulation implies that strong suppression of turbulence and a positive feedback loop may be active in this process and is responsible for the spontaneous formation of large-radius ITB. In an unstable ITB phase, an ELM crash is observed to have a positive effect on transient formation of large-radius ITB. The formation of this kind of ITB is found to have a shielding (protecting) effect on the core plasma while isolating the perturbation due to ELM crash.

Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN, despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

Internal Transport Barrier Broadening through Subdominant Mode Stabilization in Reversed Field Pinch Plasmas.

The reversed field pinch (RFP) device RFX-mod features strong internal transport barriers when the plasma accesses states with a single dominant helicity. Such transport barriers enclose a hot helical region with high confinement whose amplitude may vary from a tiny one to an amplitude encompassing an appreciable fraction of the available volume. The transition from narrow to wide thermal structures has been ascribed so far to the transport reduction that occurs when the dominant mode separatrix, which is a preferred location for the onset of stochastic field lines, disappears. In this Letter we show instead that the contribution from the separatrix disappearance, by itself, is marginal and the main role is instead played by the progressive stabilization of secondary modes. The position and the width of the stochastic boundary encompassing the thermal structures have been estimated by applying the concept of a 3D quasiseparatrix layer, developed in solar physics to treat reconnection phenomena without true separatrices and novel to toroidal laboratory plasmas. Considering the favorable scaling of secondary modes with the Lundquist number, these results open promising scenarios for RFP plasmas at temperatures higher than the presently achieved ones, where lower secondary modes and, consequently, larger thermal structures are expected. Furthermore, this first application of the quasiseparatrix layer to a toroidal plasma indicates that such a concept is ubiquitous in magnetic reconnection, independent of the system geometry under investigation.

Ion Internal Transport Barrier in the Large Helical Device

AbstractThe internal transport barrier (ITB) appears in ion root plasmas with NBI heating, when the Te/Ti ratio is held below unity in LHD. The improvement of transport is observed only in the ion heat transport, not in the electron heat transport nor in the particle transport. During the formation of ion ITB, the region of low ion thermal diffusivity expands toward the plasma center and results in a wide weak ITB, which is in contrast to the narrow strong ITB typically observed in tokamaks (© 2010 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Cyclic scenarios for steady-state operation of tokamak reactors

A new concept of steady-state scenario for tokamak reactors is proposed. It is based on cyclic operations, alternating phases of positive and negative loop voltage with no magnetic flux consumption on average. Localized non-inductive current drive by electron cyclotron waves is used to trigger and sustain an internal transport barrier (ITB), whereas neutral beam current drive is used to periodically recharge the tokamak transformer. The fact of operating in cycles relaxes the hard constraint of simultaneous fusion performance maximization and full non-inductive operation, within the MHD stability limits. Integrated modelling simulations are performed to apply this concept for the ITER steady-state regime. A linear MHD analysis of the instabilities that could appear in this type of scenario is performed, showing that MHD stability would be strongly improved with respect to a steady regime with a strong ITB.

Identity physics experiment on internal transport barriers in JT-60U and JET

A series of experiments have been carried out in 2008 at JT-60U and JET to find common characteristics and explain differences between internal transport barriers (ITBs). The identity experiments succeeded in matching the profiles of most dimensionless parameters at the time ITBs were triggered. Thereafter the q-profile development deviated due to differences in non-inductive current density profile, affecting the ITB. Furthermore, the ITBs in JET were more strongly influenced by the H-mode pedestal or edge localized modes. It was found to be difficult to match the plasma rotation characteristics in both devices. However, the wide range of Mach numbers obtained in these experiments shows that the rotation has little effect on the triggering of ITBs in plasmas with reversed magnetic shear. On the other hand the toroidal rotation and more specifically the rotational shear had an impact on the subsequent growth and allowed the formation of strong ITBs.

Effects of low central fuelling on density and ion temperature profiles in reversed shear plasmas on JT-60U

Effects of low central fuelling on density and ion temperature profiles have been investigated using negative ion based neutral beam injection and electron cyclotron heating (ECH) in reversed shear plasmas on JT-60U. Strong internal transport barrier (ITB) was maintained in density and ion temperature profiles, when central fuelling was decreased by switching positive ion based neutral beam injection to ECH after the strong ITB formation. Similar density and ion temperature ITBs were formed for the low and high central fuelling cases during the plasma current ramp-up phase. Strong correlation between the density gradient and the ion temperature gradient was observed, indicating that particle transport and ion thermal transport are strongly coupled or the density gradient assists the ion temperature ITB formation through suppression of drift wave instabilities such as ion temperature gradient mode. These results support that the density and ion temperature ITBs can be formed under reactor relevant conditions.

Characteristics of the H-mode pedestal in improved confinement scenarios in ASDEX Upgrade, DIII-D, JET and JT-60U

Pedestal and global plasma parameters are compared in conventional ELMy H-modes and improved confinement discharges from ASDEX Upgrade (AUG), DIII-D, JET and JT-60U with varying net input power. Both electron and ion pedestal pressures are studied. The pedestal top pressure pPED increases moderately with power in all tokamaks, in broad agreement with the power dependence of the IPB98(y, 2) scaling. Higher pedestal pressures are observed in AUG improved H-modes and in JT-60U high βpol discharges at q95 ∼ 6.5 and high triangularity. For all machines and all scenarios a robust correlation between the total and the pedestal thermal stored energy is observed, with the ratio of the two varying between ∼0.3 and 0.5. However the relative importance of pedestal and core confinement varies from regime to regime. In AUG the confinement improvement with respect to the IPB98(y, 2) scaling is due to improved pedestal confinement in improved H-modes with early heating and to both improved pedestal and core confinement in improved H-modes with late heating. In DIII-D hybrid discharges the increase in confinement factor compared with conventional H-modes is due to improved confinement in the plasma core. JT-60U reversed shear H-modes have strong internal transport barriers and thus improved core performance. In all four tokamaks improved edge stability is correlated with increasing total βpol and H98(y,2) increases with pedestal βpol. The analysed multimachine data set supports a scaling expression for the pedestal stored energy derived under the assumption that the dominant loss term for the pedestal is by thermal conduction in the edge transport barrier region.

Physics of strong internal transport barriers in JT-60U reversed-magnetic-shear plasmas

The physics of strong internal transport barriers (ITBs) in JT-60U reversed-magnetic-shear (RS) plasmas has been studied through the modelling on the 1.5 dimensional transport simulation. The key physics to produce two scalings on the basis of the JT-60U box-type ITB database are identified. As for the scaling for the narrow ITB width proportional to the ion poloidal gyroradius, the following three physics are important: (1) the sharp reduction of the anomalous transport below the neoclassical level in the RS region, (2) the autonomous formation of pressure and current profiles through the neoclassical transport and the bootstrap current and (3) the large difference between the neoclassical transport and the anomalous transport in the normal-shear region. As for the scaling for the energy confinement inside ITB (ϵfβp,core ≈ 0.25, where ϵf is the inverse aspect ratio at the ITB foot and βp,core is the core poloidal beta value), the value of 0.25 is found to be a saturation value due to the MHD equilibrium. The value of ϵfβp,core reaches the saturation value, when the box-type ITB is formed in the strong RS plasma with a large asymmetry of the poloidal magnetic field, regardless of the details of the transport and the non-inductively driven current.

Density profile peaking in the presence of ECRH heating in TCV

Experimental observations in stationary ECRH discharges in TCV show that additional heating has a strong effect on the electron density profile. In the absence of MHD activity or strong internal transport barriers, additional electron heating generally leads to a broadening of the density profiles with respect to the Ohmic target plasma profiles. In the ECRH L-mode and in the presence of weak electron internal transport barriers, the density peaking factor depends on the edge safety factor, the power deposition profile and ECRH power. Beyond a critical power of some 0.5 MW, the power dependence saturates. The edge safety factor dependence is supportive of turbulent equipartition (TEP) theory, which predicts inward convection in the presence of turbulence. The observation of a reduction in the peaking with central electron heating supports drift wave turbulence theory, which predicts the decrease of inward particle convection in addition to the inward convection by TEP, when trapped electron modes are destabilized, thereby reducing the net inward convection.

MHD instabilities leading to disruptions in low beta JT-60U reversed shear plasmas

High performance reversed shear (RS) discharges with strong internal transport barrier (ITB) and flat pressure profile in the plasma core region disrupt frequently even with low beta. We analysed MHD instabilities leading to low beta disruption with measuring fluctuations and current profile with MSE measurement. We mainly observed two types of disruptions. One is the disruption without precursor at surface q ∼ integer. The other is the disruption with n = 1 precursor of γ > 100 ms. The poloidal mode number of the n = 1 mode is equal to the outermost integer of q. The n = 1 mode exists continuously from the peripheral region to the ITB layer or the n = 1 modes exist separately at the peripheral region and at the ITB layer. The phase has a difference of 180° between at the peripheral region and at the ITB layer. To explain these characteristics of disruption, we introduce a simple model where disruption occurs when both MHD instabilities at the plasma surface and at the q = m/n surface in the RS region, where m and n are the poloidal and the toroidal mode numbers of the surface MHD instability, are simultaneously unstable. This simple model can explain almost all observed disruption by two processes. One is the surface mode triggered disruption, which occurs when surface q changes; the corresponding q = m/n surface at ITB layer changes discretely. The other is the internal mode triggered disruption, which occurs when the corresponding q = m/n surface becomes unstable gradually.

Compatibility of advanced tokamak plasma with high density and high radiation loss operation in JT-60U

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high βp H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operating regime is extended to high density above the Greenwald density (nGW) with high confinement (HHy2 > 1) and high radiation loss fraction (frad > 0.9) by tailoring the internal transport barriers (ITBs). With a small plasma-wall gap, the radiation loss in the main plasma (inside the magnetic separatrix) reaches 80% of the heating power due to metal impurity accumulation. However, high confinement of HHy2 = 1.2 is sustained even with such a large radiation loss in the main plasma. By neon seeding, the divertor radiation loss is enhanced from 20% to 40% of the total radiation loss. In the high βp H-mode plasmas, high confinement (HHy2 = 0.96) is maintained at high density () with high radiation loss fraction (frad ∼ 1) by utilizing high-field-side pellets and argon (Ar) injection. The high is attributed to the formation of strong density ITB. Strong core-edge parameter linkage for confinement improvement is observed, where the pedestal pressure and the core plasma confinement increase together. The measured radiation profile including contributions from all impurities in the main plasma is peaked, and the central radiation is ascribed to the contribution from Ar accumulated inside the ITB. Impurity transport analyses indicate that the Ar density profile, twice as peaked as the electron density profile, which is the same level as that observed in the high βp H-mode plasma, can yield an acceptable radiation profile even with a peaked density profile in a fusion reactor.

Timescale and magnitude of plasma thermal energy loss before and during disruptions in JET

In this paper we analyse and discuss the thermal energy loss dynamics before and during JET disruptions that occurred between 2002 and 2004 in discharges which reached >4.5 MJ of thermal energy. We observe the slow thermal energy transients with diamagnetic loops and the fast ones with electron cyclotron emission and soft x-ray diagnostics. For most disruption types in JET, the plasma thermal energy at the time of the thermal quench is substantially less than that of the full performance plasma, typically in the range of 10–50% depending on plasma conditions and disruption type. The exceptions to this observation are disruptions in plasmas with a strong internal transport barrier (ITB) and in discharges terminating in a pure vertical displacement event, in which the plasma conserves a very high energy content up to the thermal quench. These disruption types are very sudden, leaving little scope for the combined action of soft plasma landing strategies and intrinsic performance degradation, both requiring >500 ms to be effective, to decrease the available thermal energy. The characteristic time for the loss of energy from the main plasma towards the PFCs in the thermal quench of JET disruptions is in the range 0.05–3.0 ms. The shortest timescales are typical of disruptions caused by excessive pressure peaking in ITB discharges. The available thermal energy fraction and thermal quench duration observed in JET can be processed (with due caution) into estimates for the projected PFC lifetime of the ITER target.

Properties of internal transport barrier formation in JT-60U

The dependence of the thermal diffusivity on the heat flux has been investigated in JT-60U plasmas by varying the neutral beam power. In positive magnetic shear (PS) plasmas, the ion thermal diffusivity (χi) in the core region generally increases with the heating power, as with the L mode at low heating powers. However, as a result of the intensive central heating, a weak internal transport barrier (ITB) is formed, and the χi value in the core region starts to decrease. Corresponding to a further increase in the heating power, a strong ITB is formed and the χi value is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of the χi value is observed in any of the heating regimes. The electron thermal diffusivity (χe) is correlated with the χi value in PS and RS plasmas. Furthermore, the dependence of the ion thermal diffusivity on the radial electric field (Er) shear has also been investigated. By considering the non-locality in the relation between the Er shear and the χi value, a threshold in the effective Er shear to change the state from a weak to a strong ITB is identified.

Formation conditions for electron internal transport barriers in JT-60U plasmas

The formation of electron internal transport barriers (ITBs) was studied using electron cyclotron (EC) heating in JT-60U positive shear (PS) and reversed shear (RS) plasmas with scan of neutral beam (NB) power. With no or low values of NB power and with a small radial electric field (Er) gradient, a strong, box-type electron ITB was formed in RS plasmas while a peaked profile with no strong electron ITBs was observed in PS plasmas within the available EC power. When the NB power and the Er gradient were increased, the electron transport in strong electron ITBs with EC heating in RS plasmas was not affected, while electron thermal diffusivity was reduced in conjunction with the reduction of ion thermal diffusivity, and strong electron and ion ITBs were formed in PS plasmas.