J.A. Goetz

Central impurity toroidal rotation in ICRF heated Alcator C-Mod plasmas

Central impurity toroidal rotation has been observed in Alcator C-Mod ICRF heated plasmas, from the Doppler shifts of argon X ray lines. Rotation velocities of up to 1.3 × 105 m/s in the co-current direction have been observed in H mode discharges with no direct momentum input. There is a strong correlation between the increase in the central impurity rotation velocity and the increase in the plasma stored energy, induced by ICRF heating, although other factors may be involved. This implies a close association between energy and momentum confinement. Co-current rotation is also observed during purely ohmic H modes. In otherwise similar discharges with the same stored energy increase, plasmas with lower current rotate faster. For hydrogen minority (D(H)) heating, plasmas with the highest rotation have an H/D ratio between 5 and 10% and have the resonance location in the inner half of the plasma, i.e. in the same conditions that are conducive to the best ICRF absorption and heating. Comparisons with neoclassical theory indicate that the ion pressure gradient is an unimportant contributor to the central impurity rotation and the presence of a substantial core radial electric field is inferred during the ICRF pulse. An inward shift of ions induced by ICRF waves could give rise to a non-ambipolar electric field in the plasma core.

Observations of impurity toroidal rotation suppression with ITB formation in ICRF and ohmic H mode Alcator C-Mod plasmas

Co-current central impurity toroidal rotation has been observed in Alcator C-Mod plasmas with on-axis ICRF heating. The rotation velocity increases with plasma stored energy and decreases with plasma current. Very similar behaviour has been seen during ohmic H modes, which suggests that the rotation, generated in the absence of an external momentum source, is not mainly an ICRF effect. A scan of ICRF resonance location across the plasma has been performed in order to investigate possible influences on the toroidal rotation. With a slight reduction of toroidal magnetic field from 4.7 to 4.5 T and a corresponding shift of the ICRF resonance from r/a = -0.36 to -0.48, the central toroidal rotation significantly decreased together with the formation of an internal transport barrier (ITB). During the ITB phase, electrons and impurities peaked continuously for |r/a| ≤ 0.5. Comparison of the observed rotation and neoclassical predictions indicates that the core radial electric field changes from positive to negative during the ITB phase. Similar rotation suppression and ITB formation have been observed during some ohmic H mode discharges.

Tokamak-like confinement at a high beta and low toroidal field in the MST reversed field pinch

Energy confinement comparable with tokamak quality is achieved in the Madison Symmetric Torus (MST) reversed field pinch (RFP) at a high beta and low toroidal magnetic field. Magnetic fluctuations normally present in the RFP are reduced via parallel current drive in the outer region of the plasma. In response, the electron temperature nearly triples and beta doubles. The confinement time increases ten-fold (to ~10 ms), which is comparable with L- and H-mode scaling values for a tokamak with the same plasma current, density, heating power, size and shape. Runaway electron confinement is evidenced by a 100-fold increase in hard x-ray bremsstrahlung. Fokker–Planck modelling of the x-ray energy spectrum reveals that the high energy electron diffusion is independent of the parallel velocity, uncharacteristic of magnetic transport and more like that for electrostatic turbulence. The high core electron temperature correlates strongly with a broadband reduction of resonant modes at mid-radius where the stochasticity is normally most intense. To extend profile control and add auxiliary heating, rf current drive and neutral beam heating are in development. Low power lower-hybrid and electron Bernstein wave injection experiments are underway. Dc current sustainment via ac helicity injection (sinusoidal inductive loop voltages) is also being tested. Low power neutral beam injection shows that fast ions are well-confined, even in the presence of relatively large magnetic fluctuations.

Comparison of detached and radiative divertor operation in Alcator C-Mod

The divertor of the Alcator C‐Mod tokamak [Phys. Plasmas 1, 1511 (1994)] routinely radiates a large fraction of the power entering the scrape‐off layer. This dissipative divertor operation occurs whether the divertor is detached or not, and large volumetric radiative emissivities, up to 60 MW m−3 in ion cyclotron range of frequency (ICRF) heated discharges, have been measured using bolometer arrays. An analysis of both Ohmic and ICRF‐heated discharges has demonstrated some of the relative merits of detached divertor operation versus high‐recycling divertor operation. An advantage of detached divertor operation is that the power flux to the divertor plates is decreased even further than its already low value. Some disadvantages are that volumetric losses outside the separatrix in the divertor region are decreased, the neutral compression ratio is decreased, and the penetration efficiency of impurities increases.

Investigation of performance limiting phenomena in a variable phase ICRF antenna in Alcator C-Mod

High power density, phased antenna operation can often be limited by antenna voltage handling and/or impurity and density production. Using a pair of two-strap antennas for comparison, the performance of a four-strap, fast wave antenna is assessed for a variety of configurations and antenna phases in Alcator C-Mod. To obtain robust voltage handling, the antenna was reconfigured to eliminate regions where the RF E-field is parallel to B or to reduce the RF E-field to <1.0 MV m−1. To limit impurity generation, BN tiles were used to replace the original Mo tiles, a BN clad septum was inserted to limit field line connection length, and BN–metal interfaces were shielded from the plasma. With these modifications, the antenna heating efficiency and impurity generation are nearly identical to those of the two-strap antennas and independent of antenna phase in L-mode discharges. This antenna has achieved 11 MW m−2 in both heating and current drive phases in both L-mode and H-mode discharges.

20 years of research on the Alcator C-Mod tokamak

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-modes performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included.

Electron heating via mode converted ion Bernstein waves in the Alcator C-Mod tokamak

Highly localized direct electron heating [full width at half-maximum (FWHM)≅0.2a] via mode converted ion Bernstein waves has been observed in the Alcator C-Mod Tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Electron heating at or near the plasma center (r/a⩾0.3) has been observed in H(3He) discharges at B0=(6.0–6.5) T and ne(0)≅1.8×1020 m−3. [Here, the minority ion species is indicated parenthetically.] Off-axis heating (r/a⩾0.5) has also been observed in D(3He) plasmas at B0=7.9 T. The concentration of 3He in these experiments was in the range of n3He/ne≅(0.2–0.3) and the locations of the mode conversion layer and electron heating peak could be controlled by changing the 3He concentration or toroidal magnetic field (B0). The electron heating profiles were deduced using a rf modulation technique. Detailed comparisons with one-dimensional and toroidal full-wave models in the ion cyclotron range of frequencies have been carried out. One-dimensional full-wave code predictions were found to ...

High confinement dissipative divertor operation on Alcator C-Mod

Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] has operated a High-confinement-mode (H-mode) plasma together with a dissipative divertor and low core Zeff. The initially attached plasma is characterized by steady-state enhancement factor, HITER89P [P. N. Yushmanov et al., Nucl. Fusion 30, 1999 (1990)], of 1.9, central Zeff of 1.1, and a radiative fraction of ∼50%. Feedback control of a nitrogen gas puff is used to increase radiative losses in both the core/edge and divertor plasmas in almost equal amounts. Simultaneously, the core plasma maintains HITER89P of 1.6 and Zeff of 1.4 in this nearly 100% radiative state. The power and particle flux to the divertor plates have been reduced to very low levels while the core plasma is relatively unchanged by the dissipative nature of the divertor.

Improved-confinement plasmas at high temperature and high beta in the MST RFP

We have increased substantially the electron and ion temperatures, the electron density, and the total beta in plasmas with improved energy confinement in the Madison Symmetric Torus (MST). The improved confinement is achieved with a well-established current profile control technique for reduction of magnetic tearing and reconnection. A sustained ion temperature >1?keV is achieved with intensified reconnection-based ion heating followed immediately by current profile control. In the same plasmas, the electron temperature reaches 2?keV, and the electron thermal diffusivity drops to about 2?m2?s?1. The global energy confinement time is 12?ms. This and the reported temperatures are the largest values yet achieved in the reversed-field pinch (RFP). These results were attained at a density ~1019?m?3. By combining pellet injection with current profile control, the density has been quadrupled, and total beta has nearly doubled to a record value of about 26%. The Mercier criterion is exceeded in the plasma core, and both pressure-driven interchange and pressure-driven tearing modes are calculated to be linearly unstable, yet energy confinement is still improved. Transient momentum injection with biased probes reveals that global momentum transport is reduced with current profile control. Magnetic reconnection events drive rapid momentum transport related to large Maxwell and Reynolds stresses. Ion heating during reconnection events occurs globally, locally, or not at all, depending on which tearing modes are involved in the reconnection. To potentially augment inductive current profile control, we are conducting initial tests of current drive with lower-hybrid and electron-Bernstein waves.

Dynamo-free plasma in the reversed-field pinch : Advances in understanding the reversed-field pinch improved confinement mode

Generation and sustainment of the reversed field pinch (RFP) magnetic configuration normally relies on dynamo activity. The externally applied electric field tends to drive the equilibrium away from the relaxed, minimum energy state which is roughly described by a flat normalized parallel current density profile and is at marginal stability to tearing modes. Correlated fluctuations of magnetic field and velocity create a dynamo electric field which broadens the parallel current density profile, supplying the necessary edge current drive. These pervasive magnetic fluctuations are also responsible for destruction of flux surfaces, relegating the standard RFP to a stochastic-magnetic transport-limited device. Application of a tailored electric field profile (which matches the relaxed current density profile) allows sustainment of the RFP configuration without dynamo-driven edge current. The method used to ascertain that a dynamo-free RFP plasma has been created is reported here in detail. Several confinement...