L. Dudek

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.

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.

TOKAMAK FUSION TEST REACTOR POLOIDAL ROTATION DIAGNOSTIC (INVITED)

A new spectroscopic diagnostic was developed to measure poloidal velocity profiles of Tokamak fusion test reactor (TFTR) plasmas. Carbon poloidal velocities were measured using the Doppler shift of the Cu2009VI 5291 A impurity line of both intrinsic emission and charge exchange emission from neutral beams. Poloidal velocities are typically small (vθ⩽104u2009m/s) requiring small wavelength shifts (Δλ⩽0.2u2009A) to be measured. However, the high central ion temperatures in TFTR required the use of a low dispersion spectrometer to view the entire linewidth (full width at half maximum ⩽25 A). A very high throughput spectrometer/detector system was assembled to achieve the necessary precision in vθ. Statistical errors in the chord-averaged poloidal velocity less than 100 m/s have been obtained. The short focal length spectrometer features f/1.8 input optics, a transmission grating, and refractive optics. A thinned back-illuminated charge coupled device detector provided a high quantum efficiency (QE=75%). The diagnostic h...

Engineering overview of the National Spherical Torus Experiment (NSTX)

The NSTX project will provide a national facility for the study of plasma confinement, heating, and current drive in a low aspect ratio, spherical torus (ST) configuration, the ST configuration is an alternate confinement concept which is characterized by high /spl beta/, high elongation, high bootstrap fraction, and low B/sub T/ compared to conventional tokamaks, NSTX is the next step ST experiment following smaller experiments such as the PPPL CDX-U (Current Drive Experiment, Upgrade), the START (Small Tight Aspect Ratio Tokamak) at Culham, and the HIT (Helicity Injected Tokamak) at University of Washington, and is similar in scale to the MAST (Meg-Amp Spherical Tokamak) machine now under construction at Culham. This paper provides a description of the mission and gives an overview of the engineering features of the design of the machine and facility and discusses some of the key design solutions.

Engineering design of the National Spherical Torus Experiment

NSTX is a proof-of-principle experiment aimed at exploring the physics of the ‘spherical torus’ (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, among other advantages. The low aspect ratio (R:a, typically 1.2‐2 in ST designs compared to 4‐5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called ‘center stack’ in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.

TFTR neutral beam injected power measurement

Energy flow within TFTR neutral beamlines is measured with a waterflow calorimetry system capable of simultaneously measuring the energy deposited within four heating beamlines (three ion sources each), or of measuring the energy deposited in a separate neutral beam test stand. Of the energy extracted from the ion source on the well‐instrumented test stand, 99.5±3.5% can be accounted for. When the ion deflection magnet is energized, however, 6.5% of the extracted energy is lost. This loss is attributed to a spray of devious particles onto unmonitored surfaces. A 30% discrepancy is also observed between energy measurements on the internal beamline calorimeter and energy measurements on a calorimeter located in the test stand target chamber. Particle reflection from the flat plate calorimeter in the target chamber, which the incident beam strikes at a near‐grazing angle of 12°, is the primary loss of this energy. A slight improvement in energy accountability is observed as the beam pulse length is increased...

Progress toward commissioning and plasma operation in NSTX-U

The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5–10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2–3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3–6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymorexa0» explored.«xa0less

Operation of a TFTR ion source with a ground potential gas feed into the neutralizer

TFTR long pulse ion sources have been operated with gas fed only into the neutralizer. Gas for the plasma generator entered through the accelerator rather than directly into the arc chamber. This modification has been proposed for tritium beam operation to locate control electronics at ground potential and to simplify tritium plumbing. Source operation with this configuration and with the nominal gas system that feeds gas into both the ion source and the center of the neutralizer are compared. Comparison is based upon accelerator grid currents, beam composition, and neutral power delivered to the calorimeter. Charge exchange in the accelerator can be a significant loss mechanism in both systems at high throughput. A suitable operating point with the proposed system was found that requires 30% less gas than used presently. The extracted D+, D+2, and D+3 fractions of the beam were found to be a function of the gas throughput; at similar throughputs, the two gas feed systems produced similar extracted ion fr...

Cryosorption of helium on argon frost in Tokamak Fusion Test Reactor neutral beamlines

Helium pumping on argon frost has been investigated on Tokamak Fusion Test Reactor (TFTR) neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are ∼25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, >20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a twofold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 Torru2009l of helium, the beamline base pressure rose to 2×10−5 Torr from an initial value of 10−8 Torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 Torru2009l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity.