levels which might have a significant role in the light shift of the 22p level due to the 1.06-/im laser field are 6s, 7s, Ad, and 5d. These are far from being resonantly coupled to the 22p level, at least 1700 cm" 1 away. Their relative positions are such that their combined effects are partially cancelled* A rough evaluation showed that under these conditions the 5d level, which is expected to be responsible for the largest effect, contributes to the shift of the 22p level an amount of approximately 3xl0" 3 MHz/ MW-cm' 2 . This is at least 4 orders of magnitude less than the measured shift, and is thus completely negligible, With respect to the shift Lv g of the ground state, since it cannot be measured alone the best procedure is to calculate it as carefully and precisely possible. A calculation based on Fig. 1 has been carried out. 6 The result is &v g = -26.3 MHz/MW-cm" 2 . The dashed line in Fig. 3 corresponds to the sum of the two calculated shifts Ai/ e + Ay g , whereas the straight line corresponds to a least-squares fit on the measured shifts. Agreement between experimental and theoretical results is satisfactory.To conclude, this experiment provides clear evidence for the shift of a Rydberg level, due to an intense and strongly nonresonant em field. It is of interest to note that in a pure quantum treat-PACS numbers: 52.55.Gb, 52.35.Py On the PDX tokamak, large-amplitude magnetohydrodynamic (MHD) fluctuations have been observed during plasma heating by injection of high-ment, radiative corrections can be interpreted as the sum of spontaneous and stimulated radiative corrections. The net effect of spontaneous radiative corrections due to vacuum fluctuations is well known to be responsible for the Lamb shift. In the same spirit, the light shifts which have been studied in our experiment can perhaps be viewed as resulting from the stimulated radiative corrections induced by an intense and nonresonant em field.We thank Professor CI. Cohen-Tannoudji for many helpful discussions concerning both the experiment and its interpretation. We are indebted to Dr. M. Aymar and Dr. M. Crance for their calculation of the shift of the ground state.Strong magnetohydrodynamic activity has been observed in PDX neutral-be am-heated discharges. It occurs for fi T q^ 0.045 and is associated with a significant loss of fast ions and a drop in neutron emission. As much as 20%~-40% of the beam heating power may be lost. The instability occurs in repetitive bursts of oscillations of ^ 1 msec duration at 1-6-msec intervals. The magnetohydrodynamic activity has been dubbed the "fishbone instability" from its characteristic signature on the Mirnov coils.

Research in the National Spherical Torus Experiment, NSTX, has been conducted to establish spherical torus plasmas to be used for high-, auxiliary heated experiments. The device has a major radius R 0 = 0.86 m, a midplane half-width of 0.7 m, and has been operated with toroidal magnetic field B 0 ≤ 0.3 T and I p ≤ 1.0 MA. The evolution of the plasma equilibrium is analyzed between shots with an automated version of the EFIT code. Limiter, double-null, and lower single-null diverted configurations have been sustained for several energy confinement times. Plasma stored energy has reached 92 kJ (t = 17.8 %) with neutral beam heating. Plasma elongation of 1.6 ≤ ≤ 2.0 and triangularity in the range 0.25 ≤ ≤ 0.45 have been sustained, with values of = 2.5 and = 0.6 being reached transiently. The reconstructed magnetic signals are fit to the corresponding measured values with low error. Aspects of the plasma boundary, pressure, and safety factor profiles are supported by measurements from non-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limit in deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilities including sawteeth and reconnection events (REs), characterized by Mirnov oscillations, and perturbation of the I p , , and i evolution, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

A transport code (TRANSP) is used to simulate future deuterium-tritium (DT) experiments in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modelling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modelling. Fusion energy yields and 01 particle parameters are calculated, including profiles of the 01 slowing down time, the 01 average energy, and the AlfvBn speed and frequency. Two types of simulation are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the Do in the neutral beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that 01 heating will enhance the plasma parameters or that confinement will increase with the addition of tritium. The maximum relative fusion yield calculated for these simulations is QDT-0.3, and the maximum a contribution to the central toroidal 0 is PJO)-0.5%. The stability of toroidicity induced Alfvkn eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the simulations, particularly after the termination of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed.

Neutral-beam-heated plasmas in TFTR show evidence of substantial non-Ohmically driven toroidal current, even for balanced beam momentum input. The observations are inconsistent with calculations including only Ohmic and beam-driven currents, and presently can only be matched by models including the neoclassical bootstrap current.

The use of high resolution x-ray crystal spectrometers to diagnose fusion plasmas has been limited by the poor spatial localization associated with chord integrated measurements. Taking advantage of a new x-ray imaging spectrometer concept [M. Bitter et al., Rev. Sci. Instrum. 75, 3660 (2004)], and improvements in x-ray detector technology [Ch. Broennimann et al., J. Synchrotron Radiat. 13, 120 (2006)], a spatially resolving high resolution x-ray spectrometer has been built and installed on the Alcator C-Mod tokamak. This instrument utilizes a spherically bent quartz crystal and a set of two dimensional x-ray detectors arranged in the Johann configuration [H. H. Johann, Z. Phys. 69, 185 (1931)] to image the entire plasma cross section with a spatial resolution of about 1 cm. The spectrometer was designed to measure line emission from H-like and He-like argon in the wavelength range 3.7 and 4.0 A with a resolving power of approximately 10,000 at frame rates up to 200 Hz. Using spectral tomographic techniques [I. Condrea, Phys. Plasmas 11, 2427 (2004)] the line integrated spectra can be inverted to infer profiles of impurity emissivity, velocity, and temperature. From these quantities it is then possible to calculate impurity density and electron temperature profiles. An overview of the instrument, analysis techniques, and example profiles are presented.

In Alcator C-Mod discharges lower hybrid waves have been shown to induce a countercurrent change in toroidal rotation of up to 60 km=s in the central region of the plasma (r=a $ <0:4). This modification of the toroidal rotation profile develops on a time scale comparable to the current redistribution time ($100 ms) but longer than the energy and momentum confinement times ($20 ms). A comparison of the co-and countercurrent injected waves indicates that current drive (as opposed to heating) is responsible for the rotation profile modifications. Furthermore, the changes in central rotation velocity induced by lower hybrid current drive (LHCD) are well correlated with changes in normalized internal inductance. The application of LHCD has been shown to generate sheared rotation profiles and a negative increment in the radial electric field profile consistent with a fast electron pinch. The beneficial effects of rotation on toroidal plasmas have been well documented. Strong rotation can help stabilize destructive magneto-hydrodynamic instabilities (i.e., resistive wall modes [1,2]) while gradients in rotation can improve confinement by suppressing turbulence [3,4]. In many experiments the rotation profiles associated with improved performance are generated through the use of neutral beam injection. This approach may prove impractical in the large, high density plasmas envisioned for next generation devices such as ITER [5,6]. As a result, there is a need to develop alternative methods for driving plasma rotation. Significant self-generated flows have been observed on a number of tokamaks [7] suggesting that it may be possible to reap the benefits of rotation without the use of neutral beams.Self-generated flows associated with lower hybrid current drive (LHCD) have been observed in both L-mode and H-mode discharges on Alcator C-Mod when lower hybrid waves are launched such as to drive positive current. These changes to the toroidal rotation profile are core localized (r=a $ <0:4) and always in the countercurrent direction. When the waves are launched against the inductive toroidal electric field, very little current is driven and no effect on the rotation profile is observed. This result indicates that it is the LHCD (as opposed to heating) that is responsible for the countercurrent change in toroidal rotation. In discharges with sufficient LHCD, a region of enhanced velocity shear forms concurrently with a negative increment in the radial electric field profile.The results presented in this Letter were obtained on Alcator C-Mod [8], a compact tokamak (major radius R ¼ 0:67 m, typical minor radius ¼ 0:21 m) that operates at the high magnetic fields (B t > 5 T) and high densities (n e $ 10 20 m À3 ) envisaged for burning plasma reactors such as ITER and DEMO [9]. In the experiments described here, the lower hybrid waves were injected by an 88 wave guide launcher capable of delivering up to 1.2 MW of power at 4.6 GHz with an n k range of 1.5-4 in either direction [10]. (Here n k is the refractive index of the injected ...

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