Yueng Kay Martin Peng

Exploration of Spherical Torus Physics in the NSTX Device

The National Spherical Torus Experiment (NSTX) is being built at the Princeton Plasma Physics Laboratory to test the fusion physics principles for the Spherical Torus (ST) concept at the MA level. The NSTX nominal plasma parameters are R {sub 0} = 85 cm, a = 67 cm, R/a greater than or equal to 1.26, B {sub T} = 3 kG, I {sub p} = 1 MA, q {sub 95} = 14, elongation {kappa} less than or equal to 2.2, triangularity {delta} less than or equal to 0.5, and plasma pulse length of up to 5 sec. The plasma heating/current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up, as well as a dispersive scrape-off layer for heat and particle flux handling.

Features of Spherical Torus Plasmas

The spherical torus is a very small aspect ratio (A 2 are characterized by high toroidal beta (βt > 0.2), low poloidal beta (βp 1.5), and strong magnetic helical pitch (Θ comparable to F). A large near-omnigeneous region is seen in the large major radius, bad curvature region of the plasma in comparison with the conventional tokamaks. These features combine to engender the spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost. Because of its strong paramagnetism and helical pitch, the spherical torus plasma shares some of the desirable features of spheromak and reversed-field pinch (RFP) plasmas, but with tokamak-like confinement and safety factor q. The general class of spherical tori, which includes the spherical tokamak (q>1), the spherical pinch (1>q>0), and the spherical RFP (q<0), have magnetic field configurations unique in comparison with conventional tokamaks and RFPs.

Non-inductive current generation in NSTX using coaxial helicity injection

Coaxial helicity injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges, which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration than any CHI discharges previously produced in a spheromak or a spherical torus.

β-Limiting MHD instabilities in improved-performance NSTX spherical torus plasmas

Global magnetohydrodynamic (MHD) stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized β and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable β. As a result of these improvements, peak β values have reached (not simultaneously) βT = 35%, βN = 6.4, βN = 4.5, βN/li = 10, and βP = 1.4. High βP operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 s with sustained periods of βN≈6 above the ideal no-wall limit and near the with-wall limit. Details of the β-limit scalings and β-limiting instabilities in various operating regimes are described.

The physics of spherical torus plasmas

Broad and important progress in plasma tests, theory, new experiments, and future visions of the spherical torus (ST, or very low aspect ratio tokamaks) have recently emerged. These have substantially improved our understanding of the potential properties of the ST plasmas, since the preliminary calculation of the ST magnetohydrodynamic equilibria more than a decade ago. Exciting data have been obtained from concept exploration level ST experiments of modest capabilities (with major radii up to 35 cm), making important scientific contributions to toroidal confinement in general. The results have helped approval and construction of new and/or more powerful ST experiments, and stimulated an increasing number of theoretical calculations of interest to magnetic fusion energy. Utilizing the broad knowledge base from the successful tokamak and advanced tokamak research, a wide range of new ST physics features has been suggested. These properties of the ST plasma will be tested at the 1 MA level with major radiu...

High-pressure flux-conserving tokamak equilibria

Magnetohydrodynamic (MHD) tokamak equilibria are found with values of β up to 20% and realistic MHD safety factor values (e.g. q(axis) = 1 and q(edge) = 4.8) for tokamaks with aspect ratio A = 4 and D-shaped cross-section. If such equilibria can be attained experimentally, they will be very attractive for decreasing the projected costs of tokamak power reactors. In the flux-conserving tokamak (FCT) model, where rapid heating is applied to an already relatively hot plasma, these high-β equilibria are achievable. The quasi-static evolution of FCT equilibria as β increases is studied. An operating window is found in the pressure profile width wp at the half-height: for high β, the values of wp lie between 0.40 and 0.55 times the minor plasma diameter. Within this window, plasma current and poloidal β increase with β. For fixed plasma boundary, significant poloidal surface currents are induced, but these can be eliminated by small increases in the plasma minor radius, the pressure profile width, or the vacuum toroidal field.

Evolution of flux-conserving tokamak equilibria with preprogrammed cross sections

Abstract The evolution of MHD equilibria toward high β is modeled by magnetic flux conservation with a given q ( ψ ) and by single fluid particle and energy balances which determine p ( ψ , t ). These one-dimensional flux surface averaged equations, written with magnetic flux ψ as the independent variable, are coupled to the two-dimensional MHD equilibrium equation through ψ, p ( ψ , t ), and q ( ψ ). The location and evolution of the plasma cross section boundary are precisely specified through the use of a fixed boundary equilibrium technique. In moving boundary studies (e.g., plasma compression) the resulting system of equations is advanced in time from an initial state by a procedure which utilizes two nested predictor-corrector loops together with an implicit time-stepping technique. The inner predictor-corrector loop advances the transport equations subject to a given equilibrium configuration while the outer loop evolves the equilibrium. For fixed plasma boundaries this procedure is modified for greater computational speed. These techniques provide satisfactory numerical convergence together with complete consistency between the coupled one-dimensional system of equations and the two-dimensional equilibrium. This method can be applied to the study of equilibrium evolution involving dramatic changes of plasma position, shape, and profiles while prescribing the evolution of the plasma boundary. As such, it can serve as a useful tool in the design of poloidal field systems or as a source of equilibria in high-β MHD stability studies. As an example, the compressional scaling laws of Furth and Yoshikawa are found to be modified for small aspect ratio.

H-mode research in NSTX

H-modes are routinely obtained in the National Spherical Torus Experiment (NSTX) and have become a standard operational scenario. L–H transitions triggered by NBI heating have been obtained over a wide parameter range in Ip, Bt, and e in either lower-single-null (LSN) or double-null (DN) diverted discharges. Edge localized modes are observed in both configurations but the characteristics differ between DN and LSN, which also have different triangularities (δ). An H-mode duration of 500 ms was obtained in LSN, with a total pulse length of ~1 s. Preliminary power threshold studies indicate that the L–H threshold is between 600 kW and 1.2 MW, depending on the target parameters. Gas injector fuelling from the centre stack (i.e. the high toroidal field side) has enabled routine H-mode access, and comparisons with low-field side (LFS) fuelled H-mode discharges show that the LFS fuelling delays the L–H transition and alters the pre-transition plasma profiles. Gas puff imaging and reflectometry show that the H-mode edge is usually more quiescent than the L-mode edge. Divertor infrared camera measurements indicate up to 70% of available power flows to the divertor targets in quiescent H-mode discharges.

Fusion Nuclear Science Facility (FNSF) Before Upgrade to Component Test Facility (CTF)

Abstract The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

TSC plasma halo simulation of a DIII-D vertical displacement episode

A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench and current quench. With a suitable halo surrounding the main plasma, the TSC predictions are in good agreement with experimental results for the plasma current decay, plasma trajectory, toroidal and poloidal vessel currents, and for the magnetic probe and flux loop values for the entire VDE. Simulations with no plasma halo yield much faster vertical motion and significantly worse agreement with magnetic and flux loop data than do halo simulations