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Title: Optimization of Outer Poloidal Field (PF) Coil Configurations for Inductive PF Coil-only Plasma Start-up on Spherical Tori

Abstract

The elimination of in-board ohmic heating solenoid is required for the spherical torus (ST) to function as an attractive fusion power plant. An in-board ohmic solenoid, along with the shielding needed for its insulation, increases the size and, hence, the cost of the plant. Here, we investigate using static as well as dynamic codes in ST geometries a solenoid-free start-up concept utilizing a set of out-board poloidal field coils. By using the static code, an optimization of coil positions as well as coil currents was performed to demonstrate that it is indeed possible to create a high quality multi-pole field null region while retaining significant flux (volt-seconds) needed for the subsequent current ramp-up. With the dynamic code that includes the effect of vacuum vessel eddy currents, we then showed that it is possible to maintain a large size field null region for several milliseconds in which sufficient ionization avalanche can develop in the applied toroidal electric field. Under the magnetic geometry typical of a next generation spherical torus experiment, it is shown that the well-known plasma breakdown conditions for conventional ohmic solenoid start-up of E(sub)TB(sub)T/B(sub)P {approx} (0.1-1) kV/m with V(sub)loop {approx} 6 V can be readily met while retaining significantmore » volt-seconds {approx} 4 V-S sufficient to generate multi-MA plasma current in STs.« less

Authors:
; ;
Publication Date:
Research Org.:
Princeton Plasma Physics Lab., Princeton, NJ (US)
Sponsoring Org.:
USDOE Office of Science (SC) (US)
OSTI Identifier:
827828
Report Number(s):
PPPL-3939
TRN: US0403708
DOE Contract Number:
AC02-76CH03073
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 9 Apr 2004
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; BREAKDOWN; EDDY CURRENTS; ELECTRIC CURRENTS; ELECTRIC FIELDS; GEOMETRY; HEATING; IONIZATION; OPTIMIZATION; PLASMA; POWER PLANTS; SHIELDING; SOLENOIDS; START-UP; OHMIC HEATING; SPHERICAL TORUS; SOLENOID-FREE START-UP; POLOIDAL FIELD COIL; INDUCTIVE

Citation Formats

Wonho Choe, Jayhyun Kim, and Masayuki Ono. Optimization of Outer Poloidal Field (PF) Coil Configurations for Inductive PF Coil-only Plasma Start-up on Spherical Tori. United States: N. p., 2004. Web. doi:10.2172/827828.
Wonho Choe, Jayhyun Kim, & Masayuki Ono. Optimization of Outer Poloidal Field (PF) Coil Configurations for Inductive PF Coil-only Plasma Start-up on Spherical Tori. United States. doi:10.2172/827828.
Wonho Choe, Jayhyun Kim, and Masayuki Ono. Fri . "Optimization of Outer Poloidal Field (PF) Coil Configurations for Inductive PF Coil-only Plasma Start-up on Spherical Tori". United States. doi:10.2172/827828. https://www.osti.gov/servlets/purl/827828.
@article{osti_827828,
title = {Optimization of Outer Poloidal Field (PF) Coil Configurations for Inductive PF Coil-only Plasma Start-up on Spherical Tori},
author = {Wonho Choe and Jayhyun Kim and Masayuki Ono},
abstractNote = {The elimination of in-board ohmic heating solenoid is required for the spherical torus (ST) to function as an attractive fusion power plant. An in-board ohmic solenoid, along with the shielding needed for its insulation, increases the size and, hence, the cost of the plant. Here, we investigate using static as well as dynamic codes in ST geometries a solenoid-free start-up concept utilizing a set of out-board poloidal field coils. By using the static code, an optimization of coil positions as well as coil currents was performed to demonstrate that it is indeed possible to create a high quality multi-pole field null region while retaining significant flux (volt-seconds) needed for the subsequent current ramp-up. With the dynamic code that includes the effect of vacuum vessel eddy currents, we then showed that it is possible to maintain a large size field null region for several milliseconds in which sufficient ionization avalanche can develop in the applied toroidal electric field. Under the magnetic geometry typical of a next generation spherical torus experiment, it is shown that the well-known plasma breakdown conditions for conventional ohmic solenoid start-up of E(sub)TB(sub)T/B(sub)P {approx} (0.1-1) kV/m with V(sub)loop {approx} 6 V can be readily met while retaining significant volt-seconds {approx} 4 V-S sufficient to generate multi-MA plasma current in STs.},
doi = {10.2172/827828},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Apr 09 00:00:00 EDT 2004},
month = {Fri Apr 09 00:00:00 EDT 2004}
}

Technical Report:

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  • Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs.
  • Coaxial Helicity Injection (CHI) has been successfully used in the National Spherical Torus Experiment (NSTX) for a demonstration of closed flux current generation without the use of the central solenoid. The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now producedmore » closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.« less
  • The basic features of a procedure for the optimization of the plasma scenario in an air core tokamak are presented. The method takes into account the eddy currents in the passive conducting structures. The problem is reduced to the synthesis of time-varying magnetic field. The solution of this inverse electromagnetic problem is carried out by means of an optimization procedure based on the receding horizon approach. The paper includes an example of application to the ITER tokamak.
  • The methods for finding poloidal and toroidal numbers of MHD oscillations from Mirnov coils are reviewed and modified. Examples of various MHD phenomena occurring during start-up on TFTR are illustrated. It is found that the MHD mode structure best fits a model with the toroidal correction included. A new algorithm which finds m,n numbers can accommodate toroidal effects which are manifested in the phase data. The algorithm can find m,n numbers with a given toroidal correction parameter lambda', (lambda' = 0 cylindrical). This algorithm is also used to find the optimal value of lambda' automatically, eliminating the need for ''guesswork.''more » The algorithm finds the best parameters to the fit much faster than more conventional computational techniques. 9 refs., 21 figs., 2 tabs.« less
  • This concept combines blanket and coil functions into a single component. The objectives of the concept are to: (1) provide design options, (2) simplify overall configuration, (3) enhance compactness, and (4) reduce costs. Some drawings of the system are given. (MOW)