Reversed Shear Discharges Research Articles

Evolution of ηi(r,t) in JT-60U reverse shear discharges

Evolution of the ion temperature gradient parameter (ηi) has been studied in JT-60U reverse shear (RS) discharges experimentally. The ηi(r,t) behaviours for RS with and without internal transport barrier (ITB) discharges were observed, where the ITB appeared as steep gradients in the ion temperature profile, the electron density and temperature profiles. The relationship between the electron density scale length (Lne) and the ion temperature scale length (LTi) for different heating methods (NBI only and NB{} + {}ICRF) in RS discharges was compared with each other. It was observed that ηi is a key parameter to characterize ITB formation for the two cases. Kinetic stability for the toroidal drift-type modes (s<0 and ηi<ηicrit) was identified experimentally in the shear reversal plasma with ITB discharges.

Ion heat pinch in reversed magnetic shear tokamak plasmas

In the well-known reversed shear discharges, it is observed that the ion thermal diffusivity (χi) falls below the standard neoclassical value (χineo), i.e., χi&amp;lt;χineo. In this paper, local turbulent ion thermal pinch (χit&amp;lt;0) is proposed as a candidate for interpreting the experimental results from χi=χineo+χit&amp;lt;χineo. To test the idea, the two-fluid theory, developed by Weiland and the Chalmers group [J. Weiland et al., Nucl. Fusion 29, 1810 (1989); H. Nordman et al., ibid 30, 983 (1990)], is used in the reversed magnetic shear tokamak plasma to study the drift mode and associated ion heat transport. The theory is extended here to include both the radial electrical field shear (dEr/dr) and electron fluid velocity (Ve) in the sheared coordinate system. Remarkably different from B−1dEr/dr, k⋅Ve directly includes the safety factor q as well as the E×B velocity VE itself, where k is the magnetic configuration-dependent wave vector. As a result, the synergetic effects of B−1dEr/dr and k⋅Ve, especially those of k⋅Ve, lead to the local turbulent ion heat pinch in the negative and weak magnetic shear region because of the wave-particle resonance. The impact of B−1dEr/dr and k⋅Ve on the growth rate and ion heat pinch is numerically investigated. Qualitatively, the present results are in good agreement with the experimental trends.

JT-60U experimental results aimed at steady-state operation of a tokamak reactor

Recent experiments on JT-60U have been focused on the establishment of the scientific basis of advanced steady-state operation, utilizing the bootstrap current, and accordingly strove for the achievement and sustainment of improved confinement discharges. As a result, a equivalent fusion multiplication factor of Q DT eq=1.25 was achieved in a transient state and Q DT eq=0.5 was sustained for ∼0.8 s in the reversed shear discharges. It was then shown that active control schemes of stored energy and neutron emission rate were effective for the production of these high performances. The internal transport barrier was sustained for ∼5 s in reversed shear plasma with off-axis LHCD. In the high- β p mode of operation, simultaneous achievement of high confinement, high beta, and full non-inductive current drive with a large fraction of bootstrap current was demonstrated. For the stabilization of the neoclassical tearing mode, the ECRF system was introduced and the injection power into the plasma of 680 kW for 2 s and 250 kW for 5 s was achieved. With the negative-ion based neutral beam injection, the current drive efficiency increased to ∼1.3×10 19A/W per m 2. The W-shaped divertor with pumping was effective for reducing deuterium recycling, carbon impurity influx, heat load on the target tiles and the H-mode threshold power as well. Effectiveness of impurity gas puffing for the reduction of halo current was also demonstrated, which can mitigate the mechanical fatigue of the tokamak at disruptions.

MHD phenomena in reversed shear discharges on ASDEX Upgrade

Ideal and resistive MHD instabilities degrade the confinement or even lead to disruptions in ASDEX Upgrade reversed shear discharges. Double tearing modes temporarily arise when the minimum q value passes the q = 2 surface. They usually become stabilized with increasing distance between the rational surfaces or by the large pressure gradient at the inner rational surface provided by central electron heating. Large pressure gradients in the region of weak shear between the two q = 2 surfaces drive infernal modes unstable. These modes can couple to external kink modes if the q value at the plasma edge is close to low order rational values. The resulting global mode usually causes disruptions. An optimized current profile has to avoid, therefore, close double rational q surfaces as well as large pressure gradients in the weak shear region.

Reversed-shear experiments in the T-10 tokamak

Results from T-10 experiments in regimes with nonmonotonic plasma current profiles are presented. The possibility of controlling the current profile j(r) by electron-cyclotron current drive is demonstrated experimentally. Nonmonotonic q profiles with the reversed shear are obtained in which the qmin value varies in a wide range, qmin=1–2.3. It is shown that the current profiles with qmin∼2 (in this case, there are two resonant magnetic surfaces q=2 in the plasma) can cause the onset of MHD instabilities. The possibility of the formation of an internal transport barrier in reversed-shear discharges in the T-10 tokamak is analyzed. In T-10, electron transport is governed by short-wavelength electron turbulence. As a result, there is no clear evidence of the formation of an inner transport barrier in these experiments.

Chapter 8: Plasma operation and control

Wall conditioning of fusion devices involves removal of desorbablehydrogen isotopes and impurities from interior device surfaces to permitreliable plasma operation. Techniques used in present devices include baking, metal film gettering, deposition of thin films of low-Z material, pulsedischarge cleaning, glow discharge cleaning, radio frequency discharge cleaning, andin situ limiter and divertor pumping. Although wall conditioning techniqueshave become increasingly sophisticated, a reactor scale facility willinvolve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficientremoval of tritium incorporated into co-deposited layers on plasma facing componentsand their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized.Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactorscale plasmas will rely on the same principles, but will face additionalchallenges, exemplified by the ITER/FDR design. The absolute positioning of outermost flux surface and divertor strike points will have to be precise andreliable in view of the high heat fluxes at the separatrix. Long pulses will requireminimal control actions, to reduce accumulation of AC losses in superconducting PF and TF coils. To this end, more complex feedback controllers areenvisaged, and the experimental validation of the plasma equilibrium response models onwhich such controllers are designed is encouraging. Present simulation codes providean adequate platform on which equilibrium response techniques can be validated.Burning plasmas require kinetic control in addition to traditional magnetic shapeand position control. Kinetic control refers to measures controlling density,rotation and temperature in the plasma core as well as in plasma periphery anddivertor. The planned diagnostics (Chapter 7) serve as sensors forkinetic control, while gas and pellet fuelling, auxiliary power and angular momentuminput, impurity injection, and non-inductive current drive constitute the controlactuators. For example, in an ignited plasma, core density controls fusion poweroutput. Kinetic control algorithms vary according to the plasma state, e.g.H- or L-mode. Generally, present facilities have demonstrated the kinetic controlmethods required for a reactor scale device. Plasma initiation - breakdown,burnthrough and initial current ramp - in reactor scale tokamaks will not involvephysics differing from that found in present day devices. For ITER, the inducedelectric field in the chamber will be ∼0.3V· m-1 - comparable to that requiredby breakdown theory but somewhat smaller than in present devices.Thus, a start-up 3MW electron cyclotron heating system will be employed to assure burnthrough. Simulations show that plasmacurrent ramp up and termination in a reactor scale device can follow proceduresdeveloped to avoid disruption in present devices. In particular, simulations remain inthe stable area of the li-q plane. For design purposes, theresistive V·s consumed during initiation is found, by experiments, to follow theEjima expression, 0.45μ0 RIp. Advanced tokamak control has two distinct goals. First, control of density, auxiliary power, and inductive current rampingto attain reverse shear q profiles and internal transport barriers, which persistuntil dissipated by magnetic flux diffusion. Such internal transport barriers can leadto transient ignition. Second, combined use poloidal field shape control withnon-inductive current drive and NBI angular momentum injection to create and controlsteady state, high bootstrap fraction, reverse shear discharges. Active n = 1 magneticfeedback and/or driven rotation will be required to suppress resistive wall modes forsteady state plasmas that must operate in the wall stabilized regime for reactorlevels of β ⩾ 0.03.

MHD phenomena in ASDEX Upgrade and TEXTOR

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 βp 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.

High performance experiments in JT-60U reversed shear discharges

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, QDTeq = 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 Zeff was obtained in the newly installed W shaped pumped divertor. The β limit in the low qmin 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.

JT-60U high performance regimes

High performance regimes of JT-60U plasmas are presented with an emphasis upon the results from the use of a semiclosed pumped divertor with a W shaped geometry. Plasma performance in transient and quasi-steady states has been significantly improved in reversed shear and high βp regimes. The reversed shear regime elevated an equivalent QeqDT transiently up to 1.25 (nD(0) τETi(0) = 8.6 × 1020m-3·s·keV) in a reactor relevant thermonuclear dominant regime. Long sustainment of enhanced confinement with internal transport barriers (ITBs) with fully non-inductive current drive in a reversed shear discharge was successfully demonstrated with LH wave injection. Performance sustainment has been extended in the high βp regime with high triangularity, achieving a long sustainment of plasma conditions equivalent to QeqDT ≈ 0.16 (nD(0) τETi(0) ≈ 1.4 × 1020 m-3·s·keV) for ∼4.5 s with a large non-inductive current drive fraction of 60-70% of the plasma current. Thermal and particle transport analyses show significant reduction of thermal and particle diffusivities around the ITB, resulting in a strong Er shear in the ITB region. The W shaped divertor is effective for helium ash exhaust, demonstrating a steady exhaust capability of τ*He/τE ≈ 3-10, which is favourable for ITER. Suppression of neutral backflow and chemical sputtering effects has been observed, while MARFE onset density is rather decreased. Negative ion based neutral beam injection (N-NBI) experiments have created a clear H mode transition. An enhanced ionization cross-section due to multistep ionization processes was confirmed as theoretically predicted. A current density profile driven by N-NBI is measured in good agreement with the theoretical prediction. N-NBI induced TAEs characterized as persistent and bursting oscillations have been observed from a low hot β of ⟨βh⟩ ≈ 0.1-0.2% without a significant loss of fast ions.

The radial electric field in a tokamak with reversed magnetic shear

Neoclassical theory with the impurity rotational velocity is used to evaluate the radial electric field, Er, in a tokamak. The result of using the complete matrix method for the deuterium–carbon plasma is compared with a reduced analytic formula for determining Er [Ernst et al., Phys. Plasmas 5, 665 (1998)]. The analytic formula is shown to overestimate the Er magnitude and its gradient. Two transport measures of the effect of the Er shear are compared for the reversed shear and enhanced reversed shear discharges in Tokamak Fusion Test Reactor [Mazzucato et al., Phys. Rev. Lett. 77, 3145 (1996)]. It is shown that the combined Er and magnetic shear measure, Υs, from linear stability theory gives a higher correlation with the observed transition between the two discharges than the vorticity measure ωs from Er shear alone.

Numerical Analysis of Ripple Loss in Reversed Shear Tokamak Operation.

Ripple loss of energetic alpha particles and neutral beam ions was calculated for reversed shear discharges in ITER-FDR(Final Design Report). The result indicates that, compared with normal operation with positive shear, reversed shear operation dramatically enhances the ripple loss. Ripple loss of alpha particles can reach 25 % with the maximum heat load of 3.7 MW/m2 on the outboard wall, while that of a 1 MeV negative ion source beam can be as high as 20 % as well. The heat load due to alpha particle loss is marginal in the light of a wall tolerance. The calculation suggests that the Toroidal Field (TF) ripple in a fusion reactor should be designed to be less than 0.6 % at the plasma surface so that ripple loss of alpha particles or beam ions can be acceptably low. Ferritic steel insert to the vacuum vessel, is a probable solution to reduce the ripple loss to an allowable level in the ITER-FDR design with the TF ripple reduction of by a factor of 0.4. With TF ripple reduction, ripple loss for alpha particles and beam ions is expected to decrease to 10 % and 4 %, respectively.

The ARIES-RS power core—recent development in Li/V designs

The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.

High bootstrap current in RF-heated discharges

Purely non-inductive steady-state discharges in tokamak reactors will require a large bootstrap current, Iboot, in order to minimize the amounts of power in external current drive systems. In present day tokamaks, high bootstrap current discharges are produced under various conditions. It has been demonstrated that the on-axis RF heatings are good candidates for producing a large bootstrap current, especially the fast magnetosonic wave direct electron heating (FWEH), for which power deposition is centrally peaked. Moreover, the heating and current drive methods using RF waves could be used as external sources to optimize Iboot through the current profile control. In this paper, the dependence of Iboot on the profiles of the plasma density, temperature, pressure and current density is analysed using a theoretical approach with a matrix formulation. Examples of quasi-steady-state discharges with large bootstrap current fraction using FWEH on Tore Supra are given. Examples of high bootstrap current fractions on reverse shear discharges (TFTR, DIII-D, ...) are also mentioned. Purely non-inductive discharges with improved confinement by combination of FWEH and lower hybrid current drive are analysed and the extrapolation to the so-called `Tore Supra CIEL phase' is reported.

Mode structure of disruption precursors in TFTR enhanced reversed shear discharges

The mode structure of the disruption precursors in the TFTR enhanced reversed shear (ERS) parameters has been studied by using Te fluctuation profiles and q profiles obtained from electron cyclotron emission (ECE) and motional Stark effect (MSE) measurements. The observed profiles of the radial displacement associated with the MHD modes were consistent with the displacement profiles expected from the ideal MHD external kink mode. The observed mode frequencies differ from the plasma toroidal rotation frequency measured with C VI charge exchange recombination light. The independence of the mode frequency from the plasma rotation frequency supports the ideal MHD hypothesis. Possible causes of the frequency difference are discussed.

Study of internal transport barriers by comparison of reversed shear and high- discharges in JT-60U

Reduction of electron thermal diffusivity was investigated by comparing the internal transport barrier (ITB) with that of high- discharges. In reversed shear discharges heated by neutral beams, reduction was significant when negative magnetic shear was strong. Radial location of ITB propagated outwards and then localized near the region of q-minimum for both the ion and the electron channels. Reduction in was also observed in the electron heating regime by ion cyclotron heating

Dynamics and control of internal transport barriers in reversed shear discharges

Transitions to an enhanced confinement regime in tokamak plasmas with negative central magnetic shear have been observed in a number of devices. A simple model incorporating the nonlinear coupling between the turbulent fluctuations and the sheared radial electric field is added to a transport model in order to investigate the dynamics of the transition to this enhanced confinement mode. In this model, by incorporating both the instability growth rate profiles and particle and/or power deposition profiles, a rich variety of transition dynamics is uncovered. Transition dynamics and their concomitant thresholds are examined within the context of these models. In the course of investigating these transitions, potential methods for triggering and controlling these enhanced confinement regimes have been discovered and are discussed.

Achievement of High Fusion Performance in JT-60U Reversed Shear Discharges

Fusion performance of reversed shear discharges with an {ital L}-mode edge has been significantly improved in a thermonuclear dominant regime with up to 2.8 MA of plasma current in the JT-60U tokamak. The core plasma energy is efficiently confined due to the existence of persistent internal transport barriers formed for both ions and electrons at a large minor radius of r/a{approximately}0.7 near the boundary of the reversed shear region. In an assumed deuterium-tritium fuel, the peak fusion amplification factor defined for transient conditions involving the dW/dt term would be in excess of unity. {copyright} {ital 1997} {ital The American Physical Society}

Internal Transport Barrier for Electrons in JT-60U Reversed Shear Discharges

A new type of internal transport barrier (ITB) has been observed in JT-60U reversed shear discharges. It accompanies clear electron temperature and density pedestals and significant reduction of effective thermal diffusivities of electrons and ions ( ${\ensuremath{\chi}}_{e}^{\mathrm{eff}}$ and ${\ensuremath{\chi}}_{i}^{\mathrm{eff}}$); ${\ensuremath{\chi}}_{e}^{\mathrm{eff}}$ sharply drops by a factor of 20 within 5 cm while ${\ensuremath{\chi}}_{i}^{\mathrm{eff}}$ is smaller than the conventional neoclassical value by a factor of 4 or more. The ratio of ion temperature to electron temperature was less than 1.5 inside the ITB. The ITB lies in the negative shear region and extended beyond 60% of the plasma minor radius. High density plasmas with high confinement have been obtained with the formation of ITB.

Dynamics of Transition to Enhanced Confinement in Reversed Magnetic Shear Discharges

A simple model of the transition to enhanced confinement in reversed shear discharges is presented. The proposed transition mechanism relies on a synergism between electric field shear suppression of turbulence aided by reduced curvature drive due to magnetic shear reversal or reduction. Profile structure and transport barrier propagation dynamics are predicted. A novel analytical theory for the time evolution of the barrier foot-point location is presented. The model predicts that the transition threshold has favorable dependence on pretransition temperature ratio $({T}_{i}{/T}_{e})$, in-out asymmetry in the $E\ifmmode\times\else\texttimes\fi{}B$ shearing rate (i.e., lower for larger Shafranov shift), density profile peakedness, and unfavorable scaling with density. Optimal confinement occurs in discharges where deposition is peaked within the magnetic shear reversal radius.

Equilibrium reconstructions of reversed-shear discharges based on motional Stark effect measurements

Internal magnetic field measurements are now routinely made in tokamak discharges with a motional Stark effect polarimeter. Using these measurements, free-boundary equilibrium reconstructions of the temporal evolution of tokamak discharges can be made. The variational moments equilibrium code (VMEC) has been upgraded with an improved plasma boundary positioning algorithm and an improved spline representation of the pressure and q profiles. The uncertainty of the reconstruction is assessed for discharges with reversed magnetic shear where q(0)≫qmin&amp;gt;1.