S. Kaye

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...

Overview of the physics and engineering design of NSTX upgrade

The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5xa0s to 5–8xa0s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described.

Attainment of high confinement in neutral beam heated divertor discharges in the PDX tokamak

Abstract The PDX divertor configuration has recently been converted from an open to a closed geometry to inhibit the return of neutral gas from the divertor region to the main chamber. Since then, operation in a regime with high energy confinement in neutral beam heated discharges (ASDEX H-mode) has been routine over a wide range of operating conditions. These H-mode discharges are characterized by a sudden drop in divertor density and H α emission and a spontaneous rise in main chamber plasma density during neutral beam injection. The confinement time is found to scale nearly linearly with plasma current, but can be degraded due either to the presence of edge instabilities or heavy gas puffing. Detailed Thomson scattering temperature profiles show high values of T c near the plasma edge (∼ 450 eV) with sharp radial gradients (∼ 400 eV/cm) near the separatrix. Density profiles are broad and also exhibit steep gradients close to the separatrix.

Global energy confinement scaling for neutral-beam-heated tokamaks

A total of 677 representative discharges from seven neutral-beam-heated tokamaks have been used to study the parametric scaling of global energy confinement time. Contributions to this data base were from Asdex, DITE, D-III, ISX-B, PDX, PLT and TFR, and were taken from results of gettered, L-mode type discharges. Assuming a power law dependence of τE on the discharge parameters κ, Ip, Bt, e Ptot, a and R. standard multiple linear regression techniques were used in two steps to determine the scaling. The results indicate that the discharges used in the study are well described by the scaling .

IDEAL MHD STABILITY LIMITS OF LOW ASPECT RATIO TOKAMAK PLASMAS

The ideal magnetohydrodynamic (MHD) stability limits of low aspect ratio tokamak plasmas are computed numerically for plasmas with a range of cylindrical safety factors q*, normalized plasma pressures beta , elongations kappa and central safety factors q(0). Four distinct regimes are optimized, namely: (a) low-q* plasmas with q(0)=1.1 with and without a stabilizing wall, (b) low-q* plasmas with no wall and 1.1<q(0)<2, (c) high- beta , high bootstrap fraction plasmas at moderate kappa requiring a wall and edge current drive and (d) high- beta , very high bootstrap fraction plasmas with moderate to high kappa requiring a stabilizing wall but little external current drive. A stable equilibrium is found at an aspect ratio of A=1.4 and an elongation of kappa =3.0, with 99.3% of the current provided by the plasma pressure and beta =45%. Special attention is paid to the issues of numerical convergence and the proper definition of bootstrap current fraction

ITER L mode confinement database

This special topic describes the contents of an L mode database that has been compiled with data from Alcator C-Mod, ASDEX, DIII, DIII-D, FTU, JET, JFT-2M, JT-60, PBX-M, PDX, T-10, TEXTOR, TFTR and Tore Supra. The database consists of a total of 2938 entries, 1881 of which are in the L phase while 922 are ohmically heated only (ohmic). Each entry contains up to 95 descriptive parameters, including global and kinetic information, machine conditioning and configuration. The special topic presents a description of the database and the variables contained therein, and it also presents global and thermal scalings along with predictions for ITER. The L mode thermal confinement time scaling, determined from a subset of 1312 entries for which the τE,th are provided, is τE,th = 0.023Ip0.96BT0.03R1.83(R/a)0.06 κ0.64ne0.40Meff0.20P-0.73 in units of seconds, megamps, teslas, metres, -, -, 10-9 m-1

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...

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

Impurity levels and power loading in the pdx tokamak with high power neutral beam injection

Abstract The PDX tokamak provides an experimental facility for the direct comparison of various impurity control techniques under reactor-like conditions. Four neutral beam lines inject > 6 MW for 300 ms. Carbon rail limiter discharges have been used to test the effectiveness of perpendicular injection, but non-disruptive full power operation for > 100 ms is difficult without extensive conditioning. Initial tests of a toroidal bumper limiter indicate reduced power loading and roughly similar impurity levels compared to the carbon rail limiter discharges. Poloidal divertor discharges with up to 5 MW of injected power are cleaner than similar circular discharges, and the power is deposited in a remote divertor chamber. High density divertor operation indicates a reduction of impurity flow velocity in the divertor and enhanced recycling in the divertor region during neutral injection.

Advanced tokamak physics-status and prospects

Experimental and theoretical results from around the world point to the possibility of high confinement, high- beta , and high-bootstrap-fraction steady-state tokamak operating modes. These modes of operation, if fully developed and extended to steady-state, could lead to much less expensive tokamak demonstration power reactors and to a significantly reduced cost-of-electricity from fusion, as compared to projections based on low- beta N, pulsed operating modes. Present results have clear implications in the areas of particle control, plasma shaping, and current-profile control. Thus they have strongly influenced the design of the steady-state advanced tokamak TPX, which has the mission to combine the best results from present experiments and extend them to steady-state. These results also have important implications for follow-up tests in ITER, which have the goal of studying advanced-tokamak operation in an ignited plasma, as well as for the eventual configuration of an advanced-tokamak fusion reactor.