26 research outputs found
Exploration of the equilibrium operating space for NSTX-Upgrade
This paper explores a range of high-performance equilibrium scenarios available in the NSTX-Upgrade device [J.E. Menard, submitted for publication to Nuclear Fusion]. NSTX-Upgrade is a substantial upgrade to the existing NSTX device [M. Ono, et al., Nuclear Fusion 40, 557 (2000)], with significantly higher toroidal field and solenoid capabilities, and three additional neutral beam sources with significantly larger current drive efficiency. Equilibria are computed with freeboundary TRANSP, allowing a self consistent calculation of the non-inductive current drive sources, the plasma equilibrium, and poloidal field coil current, using the realistic device geometry. The thermal profiles are taken from a variety of existing NSTX discharges, and different assumptions for the thermal confinement scalings are utilized. The no-wall and idealwall n=1 stability limits are computed with the DCON code. The central and minimum safety factors are quite sensitive to many parameters: they generally increases with large outer plasmawall gaps and higher density, but can have either trend with the confinement enhancement factor. In scenarios with strong central beam current drive, the inclusion of non-classical fast ion diffusion raises qmin, decreases the pressure peaking, and generally improves the global stability, at the expense of a reduction in the non-inductive current drive fraction; cases with less beam current drive are largely insensitive to additional fast ion diffusion. The non-inductive current level is quite sensitive to the underlying confinement and profile assumptions. For instance, for BT=1.0 T and Pinj=12.6 MW, the non-inductive current level varies from 875 kA with ITER-98y,2 thermal confinement scaling and narrow thermal profiles to 1325 kA for an ST specific scaling expression and broad profiles. This sensitivity should facilitate the determination of the correct scaling of transport with current and field to use for future fully non-inductive ST devices. Scenarios are presented which can be sustained for 8-10 seconds, or (20-30)τCR, at βN=3.8-4.5, facilitating, for instance, the study of disruption avoidance for very long pulse. Scenarios have been documented which can operate with βT~25% and equilibrated qmin>1. The value of qmin can be controlled at either fixed non-inductive fraction of 100% or fixed plasma current, by varying which beam sources are used, opening the possibility for feedback qmin control. In terms of quantities like collisionality, neutron emission, non-inductive fraction, or stored energy, these scenarios represent a significant performance extension compared to NSTX and other present spherical torii
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Compact DT fusion spherical tori at modest fields
A spherical torus is obtained by retaining only the indispensable components on the inboard side of a tokamak plasma, such as a cooled, normal conductor that carries current to produce a toroidal magnetic field. The resulting device features an exceptionally small aspect ratio (typically 2-to-1 elongation), and ramp-up and maintenance of the plasma current primarily by noninductive means. The tokamak plasma takes on the spherical shape with a modest hole through the center, suggesting the name of spherical torus. This paper reviews the initial assessments of near-term DT fusion devices based on the spherical torus concept
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Equilibrium field coils and free-boundary equilibrium considerations for TNS
The Next Step (TNS) tokamak is expected to have a D-shaped plasma to permit MHD stable operation with volume averaged-beta, anti..beta.. up to 10 percent. By following a procedure similar to the method of virtual casing, external coil arrangements were produced that reproduce the ideal vacuum vertical field B/sub z//sup ID/ for a D-shaped plasma to within a few percent root-mean-squared deviation. A typical coil system and a free-boundary equilibrium are shown. It is seen that D-shaped equilibria can be with the external coils. Again, three sets of coils are sufficient for centering and shaping the plasma for the range of anti..beta.. values of interest. It is concluded that placing the equilibrium field coils external to the toroidal field coils has the potential of substantially reducing the cost and complexity of the D-shaped TNS device. (MHR
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Two-point model for divertor transport
Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime
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System studies for quasi-steady-state advanced physics tokamak
Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated
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Hybrid equilibrium field coils for the ORNL TNS
In this study, we make a comparative study of the power supplies required by interior and exterior (to the toroidal field (TF) coils) equilibrium field coils that are separately appropriate for high-..beta.., D-shaped plasmas in TNS. It is shown that the interior coils need power supplies that are an order of magnitude below those required by the exterior coils (while the latter case is much less difficult to build than the former). A hybrid EF coil concept is proposed that combines the interior and the exterior coils to retain their advantages in avoiding large interior coils while lowering the power supplied to the exterior coils by an order of magnitude
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Injection scenarios for TNS
Neutral beam injection heating is the most technically advanced and experimentally proven form of auxiliary Tokamak heating. In TNS, due to its large size and density, the major problem is a lack of beam penetration. An attempt is made to determine the scope of the problem and outline possible strategies for dealing with it. Some of these strategies are: raise the injection energy, injection vertically into a locally produced minimum in the toroidal magnetic field so that the injected ions are trapped in this well and anti B x nablaB drift into the center of the machine, and using a low density startup and taking advantage of ..cap alpha..-particle heating and flux surface shifts will allow 150 keV deuteron beams to have adequate protection. (MHR
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Advantages of iron core in a tokamak
A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti ..beta.. (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated
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Plasma engineering innovations for the ORNL TNS reactor
Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma ..beta.. values up to 10 percent and averaged densities between 0.6 x 10/sup 14/ cm/sup -3/ and 2.5 x 10/sup 14/ cm/sup -3/. Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors