skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: TPX Poloidal Field (PF) power systems simulation

Abstract

This paper describes the modeling and simulation of the PF power system for the Tokamak Physics Experiment (TPX), which is required to supply pulsed DC current to the Poloidal Field (PF) superconducting coil system. An analytical model was developed to simulate the dynamics of the PF power system for any PF current scenario and thereby provide the basis for selection of PF circuit topology, in support of the major design goal of optimizing the use of the existing Tokamak Fusion Test Reactor (TFTR) facilities at the Princeton Plasma Physics Lab (PPPL).

Authors:
;  [1];  [2]
  1. Princeton Univ., NJ (United States). Plasma Physics Lab.
  2. Ebasco Services, Inc., New York, NY (United States)
Publication Date:
Research Org.:
Princeton Univ., NJ (United States). Plasma Physics Lab.
Sponsoring Org.:
USDOE, Washington, DC (United States)
OSTI Identifier:
10192323
Report Number(s):
PPPL-CFP-2966; CONF-931018-29
ON: DE94002226; TRN: 93:025632
DOE Contract Number:
AC02-76CH03073
Resource Type:
Conference
Resource Relation:
Conference: Symposium on fusion engineering,Hyannis, MA (United States),11-15 Oct 1993; Other Information: PBD: [1993]
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; TOKAMAK DEVICES; POWER SYSTEMS; SIMULATION; SUPERCONDUCTING COILS; THERMAL ANALYSIS; OPTIMIZATION; 700440; POWER SUPPLIES, ENERGY STORAGE

Citation Formats

Lu, E., Bronner, G., and Neumeyer, C.. TPX Poloidal Field (PF) power systems simulation. United States: N. p., 1993. Web.
Lu, E., Bronner, G., & Neumeyer, C.. TPX Poloidal Field (PF) power systems simulation. United States.
Lu, E., Bronner, G., and Neumeyer, C.. Mon . "TPX Poloidal Field (PF) power systems simulation". United States. doi:. https://www.osti.gov/servlets/purl/10192323.
@article{osti_10192323,
title = {TPX Poloidal Field (PF) power systems simulation},
author = {Lu, E. and Bronner, G. and Neumeyer, C.},
abstractNote = {This paper describes the modeling and simulation of the PF power system for the Tokamak Physics Experiment (TPX), which is required to supply pulsed DC current to the Poloidal Field (PF) superconducting coil system. An analytical model was developed to simulate the dynamics of the PF power system for any PF current scenario and thereby provide the basis for selection of PF circuit topology, in support of the major design goal of optimizing the use of the existing Tokamak Fusion Test Reactor (TFTR) facilities at the Princeton Plasma Physics Lab (PPPL).},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Nov 01 00:00:00 EST 1993},
month = {Mon Nov 01 00:00:00 EST 1993}
}

Conference:
Other availability
Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that hold this conference proceeding.

Save / Share:
  • The poloidal-field (PF) coil set for the Tokamak Ignition/Burn Engineering Reactor (TIBER-II) consists of 24 solenoid modules, 16 of which are stacked inside the toroidal-field (TF) system at the center of the machine. These central solenoid modules operate at high-current densities, and maximum fields at the windings approach 14 T. Although TIBER-II is designed for steady-state operation with noninductive current drive, other operating scenarios are also considered. In the pulsed or inductive mode, PF coil currents are ramped to induce plasma current. In this mode, peak fields approaching 14 T appear on the central solenoid modules at the ends ofmore » the stack; the required current densities in these modules approach 40 A . mm/sup 2/. The central solenoid modules are layer wound using cable-in-conduit conductor (CICC) with (NbTi)/sub 3/Sn composite strands for improved high-field performance. Layer winding permits grading the conductor for maximum overall winding-pack current density and also results in less wasted space in the radial build of the machine. Cooling connections may be made at each layer of a module as needed. Current leads to the modules are routed through the high-field central bore. The central solenoid modules can easily support the centering load of the PF system, reducing the overall radial build of the machine and greatly increasing the limit on the number of pulse cycles imposed by fatigue considerations in the central solenoid. 5 refs., 3 figs., 2 tabs.« less
  • The use of deuterium-tritium fuel in the Compact Ignition Tokamak will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion devices auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. In-vessel remote maintenance included replacement of divertor and first wall hardware, faraday shields, and for an in-vessel inspection system. Provision for remote replacement of a vacuum vessel sector, toroidal field coil or poloidal field ring coil was not included in the project baseline. As a result ofmore » recent coil failures experienced at a number of facilities, the CIT project decided to reconsider the question of remote recovery from a coil failure and, in January of 1990, initiated a coil replacement study. This study focused on the technical requirements and impact on fusion machine design associated with remote recovery from any coil failure.« less
  • The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this paper, VEQ and its role in the TSC analysis of the CIT PF system are described. 7 refs., 5 figs., 6 tabs.
  • The plasma shaping flexibility of the Compact Ignition Tokamak (CIT) poloidal field (PF) coil set is demonstrated through MHD equilibrium calculations of optimal PF coil current distributions and their variation with poloidal beta, internal inductance, plasma 95% elongation, and 95% triangularity. Calculations of the magnetic stored energy are used to compare solutions associated with various plasma parameters. The Control Matrix (CM) equilibrium code, together with the nonlinear equation and numerical optimization software packages HYBRD, and VMCON, respectively, are used to find equilibrium coil current distributions for fixed divertor geometry, volt-seconds, and plasma profiles in order to isolate the dependence onmore » individual parameters. A reference equilibrium and coil current distribution are chosen, and correction currents dI are determined using the CM equilibrium method to obtain other specified plasma shapes. The reference equilibrium is the {kappa} = 2 divertor at beginning of flattop (BOFT) with a minimum stored energy solution for the coil current distribution. The pressure profile function is fixed.« less
  • The Texas Experimental Tokamak (TEXT) has been operational since November of 1980. It has been running consistently since August 1981 with a 300 kA plasma current flat top. Plasma parameters from a typical shot are shown. Start-up and operation of the machine is described in a separate paper. All units are now on line and operating within design parameters. Initial testing and changes necessary to assure specified performance is described. Discharge cleaning is employed in order to minimize vacuum system contamination. The optimized characteristics of this process are described.