R. E. Bell

The effect of lithium surface coatings on plasma performance in the National Spherical Torus Experiment

National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosi...

Nondimensional transport scaling in the Tokamak Fusion Test Reactor: Is tokamak transport Bohm or gyro-Bohm?

General plasma physics principles state that power flow Q(r) through a magnetic surface in a tokamak should scale as Q(r)= {32π2Rr3Te2c nea/[eB (a2−r2)2]} F(ρ*,β,ν*,r/a,q,s,r/R,...) where the arguments of F are local, nondimensional plasma parameters and nondimensional gradients. This paper reports an experimental determination of how F varies with normalized gyroradius ρ*≡(2TeMi)1/2c/eBa and collisionality ν*≡(R/r)3/2qRνe(me/ 2Te)1/2 for discharges prepared so that other nondimensional parameters remain close to constant. Tokamak Fusion Test Reactor (TFTR) [D. M. Meade et al., in Plasma Physics and Controlled Nuclear Fusion Research, 1990, Proceedings of the 13th International Conference, Washington (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] L‐mode data show F to be independent of ρ* and numerically small, corresponding to Bohm scaling with a small multiplicative constant. By contrast, most theories predict gyro‐Bohm scaling: F∝ρ*. Bohm scaling implies that the largest scale size f...

High poloidal beta equilibria in the Tokamak Fusion Test Reactor limited by a natural inboard poloidal field null

Recent operation of the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Research 1, 51 (1986)] has produced plasma equilibria with values of Λ≡βp eq+li/2 as large as 7, eβp dia≡2μ0e〈p⊥〉/〈〈Bp〉〉2 as large as 1.6, and Troyon normalized diamagnetic beta [Plasma Phys. Controlled Fusion 26, 209 (1984); Phys. Lett. 110A, 29 (1985)], βNdia≡108〈βt⊥〉aB0/Ip as large as 4.7. When eβp dia≳1.25, a separatrix entered the vacuum chamber, producing a naturally diverted discharge that was sustained for many energy confinement times, τE. The largest values of eβp and plasma stored energy were obtained when the plasma current was ramped down prior to neutral beam injection. The measured peak ion and electron temperatures were as large as 24 and 8.5 keV, respectively. Plasma stored energy in excess of 2.5 MJ and τE greater than 130 msec were obtained. Confinement times of greater than 3 times that expected from L‐mode predictions have been achieved. The fusion power gain QDD reached a value of 1.3×10−...

Active core profile and transport modification by application of ion Bernstein wave power in the Princeton Beta Experiment-Modification

Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment‐Modification (PBX‐M) [Phys. Fluids B 2, 1271 (1990)] tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH‐assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x‐ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large ∇ne, ∇Te, ∇νφ, and ∇Ti, delimiting the confinement zone. This regime is reminiscent of the H(high) mode, but with a confinement zone moved inward. The core region has better than H‐mode confinement while the peripheral region is L(low)‐mode‐like. The peaked profile enhances NBI core deposition and increases nuclear reactivity. An increase in central Ti results from χi reduction (compared to the H mode) and better beam penetration. Boot...

Simulation of microtearing turbulence in national spherical torus experimenta)

Thermal energy confinement times in National Spherical Torus Experiment (NSTX) dimensionless parameter scans increase with decreasing collisionality. While ion thermal transport is neoclassical, the source of anomalous electron thermal transport in these discharges remains unclear, leading to considerable uncertainty when extrapolating to future spherical tokamak (ST) devices at much lower collisionality. Linear gyrokinetic simulations find microtearing modes to be unstable in high collisionality discharges. First non-linear gyrokinetic simulations of microtearing turbulence in NSTX show they can yield experimental levels of transport. Magnetic flutter is responsible for almost all the transport (∼98%), perturbed field line trajectories are globally stochastic, and a test particle stochastic transport model agrees to within 25% of the simulated transport. Most significantly, microtearing transport is predicted to increase with electron collisionality, consistent with the observed NSTX confinement scaling....

Snowflake divertor configuration studies in National Spherical Torus Experimenta)

Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width λq was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flatt...

Role of the stabilizing shell in high- beta , low-q disruptions in PBX-M

The characteristics of high- beta , low-q disruptions have been studied in PBX-M, a device with a nearby conducting shell. The coupling between the wall and the plasma was varied by choosing different plasma shapes, including nearly circular plasmas, D-shaped plasmas and bean-shaped plasmas (indented on the midplane), and by increasing the effective coverage of the plasma by the shell. Disruption precursors were observed to have a strong dependence on the coupling between the plasma and the shell. Measured mode growth times vary from between several times the Alfven time-scale (~100 mu s) to the L/R time-scale of the wall (~20 ms). The behaviour of observed disruption precursors is interpreted in terms of the resistive wall mode theory of ideal plasmas, and a detailed calculation of the stability of a strongly coupled bean configuration using the NOVA-W linear stability code is presented. The experimental observations are in good agreement with the theoretical predictions

Initial physics results from the National Spherical Torus Experiment

The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a approximately equal to 1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of High Harmonic Fast Wave (HHFW) heating and current drive, 5 MW of Neutral Beam Injection (NBI) and Co-Axial Helicity Injection (CHI) for non-inductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high-bt and high-bootstrap current fraction, and to develop methods for non-inductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA and with a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 msec was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and bt =21% were produced.

Scaling of linear microtearing stability for a high collisionality National Spherical Torus Experiment discharge

Linear gyrokinetic simulations are performed based on a high collisionality NSTX discharge that is part of dimensionless confinement scaling studies. In this discharge, the microtearing mode is predicted to be unstable over a significant region of the plasma (r/a = 0.5–0.8), motivating comprehensive tests to verify the nature of the mode and how it scales with physical parameters. The mode is found to be destabilized with sufficient electron temperature gradient, collisionality, and beta, consistent with previous findings and simple theoretical expectations. Consistent with early slab theories, growth rates peak at a finite ratio of electron-ion collision frequency over mode frequency, νe/i/ω ∼ 1–6. Below this peak, the mode growth rate decreases with reduced collisionality, qualitatively consistent with global confinement observations. Also, in this region, increased effective ionic charge (Zeff) is found to be destabilizing. Experimental electron beta and temperature gradients are two to three times lar...

Liquid lithium divertor characteristics and plasma?material interactions in NSTX high-performance plasmas

Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The Liquid Lithium Divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. The LLD consisted of a copper-backed structure with a porous molybdenum front face. Nominal Li filling levels by the end of the run campaign exceeded the porosity void fraction by 150%. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m−2, no macroscopic ejection events were observed. In addition, no substrate line emission was observed after achieving lithium-melt temperatures indicating the lithium wicks and provides a protective coating on the molybdenum porous layer. Impurity emission from the divertor suggests that the plasma is interacting with oxygen-contaminated lithium whether diverted on the LLD or not. A database of LLD discharges is analysed to consider whether there is a net effect on the discharges over the range of total deposited lithium in the machine. Examination of H-97L indicates that performance was constant throughout the run, consistent with the hypothesis that it is the quality of the surface layers of the lithium that impact performance. The accumulation of impurities suggests a fully flowing liquid lithium system to obtain a steady-state PFC on timescales relevant to NSTX.