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Title: A method for estimating tokamak poloidal field coil currents which incorporates engineering constraints

Abstract

This thesis describes the development of a design tool for the poloidal field magnet system of a tokamak. Specifically, an existing program for determining the poloidal field coil currents has been modified to: support the general case of asymmetric equilibria and coil sets, determine the coil currents subject to constraints on the maximum values of those currents, and determine the coil currents subject to limits on the forces those coils may carry. The equations representing the current limits and coil force limits are derived and an algorithm based on Newton's method is developed to determine a set of coil currents which satisfies those limits. The resulting program allows the designer to quickly determine whether or not a given coil set is capable of supporting a given equilibrium. 25 refs.

Authors:
Publication Date:
Research Org.:
Massachusetts Inst. of Tech., Cambridge, MA (USA). Plasma Fusion Center
Sponsoring Org.:
DOE/ER
OSTI Identifier:
6541751
Report Number(s):
DOE/ET/51013-T231; PFC/RR-90-8
ON: DE91001662; TRN: 91-000101
DOE Contract Number:
AC02-78ET51013
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; MAGNET COILS; DESIGN; ALGORITHMS; ASYMMETRY; CURRENT DENSITY; ELECTROMOTIVE FORCE; EQUATIONS; NEWTON METHOD; SUPERCONDUCTING COILS; TOKAMAK DEVICES; CLOSED PLASMA DEVICES; ELECTRIC COILS; ELECTRICAL EQUIPMENT; EQUIPMENT; ITERATIVE METHODS; MATHEMATICAL LOGIC; THERMONUCLEAR DEVICES; 700202* - Fusion Power Plant Technology- Magnet Coils & Fields

Citation Formats

Stewart, W.A. A method for estimating tokamak poloidal field coil currents which incorporates engineering constraints. United States: N. p., 1990. Web. doi:10.2172/6541751.
Stewart, W.A. A method for estimating tokamak poloidal field coil currents which incorporates engineering constraints. United States. doi:10.2172/6541751.
Stewart, W.A. Tue . "A method for estimating tokamak poloidal field coil currents which incorporates engineering constraints". United States. doi:10.2172/6541751. https://www.osti.gov/servlets/purl/6541751.
@article{osti_6541751,
title = {A method for estimating tokamak poloidal field coil currents which incorporates engineering constraints},
author = {Stewart, W.A.},
abstractNote = {This thesis describes the development of a design tool for the poloidal field magnet system of a tokamak. Specifically, an existing program for determining the poloidal field coil currents has been modified to: support the general case of asymmetric equilibria and coil sets, determine the coil currents subject to constraints on the maximum values of those currents, and determine the coil currents subject to limits on the forces those coils may carry. The equations representing the current limits and coil force limits are derived and an algorithm based on Newton's method is developed to determine a set of coil currents which satisfies those limits. The resulting program allows the designer to quickly determine whether or not a given coil set is capable of supporting a given equilibrium. 25 refs.},
doi = {10.2172/6541751},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue May 01 00:00:00 EDT 1990},
month = {Tue May 01 00:00:00 EDT 1990}
}

Technical Report:

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  • This paper proposes a new method for mapping the plasma operational space of a tokamak reactor. The operational space within poloidal field (PF) coil engineering constraints is defined by three parameters, {psi}, {beta}{sub p}, and I{sub p}, which are the magnetic flux supplied by the PF coil system, the poloidal beta, and the plasma current, respectively. It is also shown that the boundaries of the plasma operational space have a one-to-one correspondence with the PF coils. The design specifications of the PF coil system are thus related to the plasma operational space.
  • MIT has had the primary responsibility for the PF coil system (WBS G) during the CIT conceptual design. The purpose of this report is to summarize the status of that subsystem design as of April 15, 1986. Coil locations and currents have been defined in a manner consistent with physics requirements (as of February 21, 1986) and system code selections of radial build dimensions, volt-second, and pulse time requirements (February 12, 1986). Plasma equilibria have been determined for limiter and divertor cases and have driven the design. A design summary is given in Section 2.0 while more detailed discussions aremore » contained in Sections 3.0 through 6.0.« less
  • A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the integrated-blanket-coil (IBC) concept, is applied to the poloidal field and blanket systems of a Tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to less than or equal to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, MHD-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference casemore » examined in this paper, the MHD-induced pressure drop was estimated to be approx. 1/3 MPa and the associated primary membrane stress was estimated to be approx. 47 MPa. The preliminary analyses presented in this paper indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coils functions in a single component.« less
  • This paper presents a method for obtaining the locations and currents of the poloidal coil systems for a tokamak, given an desirable magnetohydrodynamic equilibrium for the device. The method involves a simultaneous minimization of the match to the desired poloidal field and the stored energy in the coils, subject to the constraints necessary to achieve decoupling of the equilibrium and inductive-current-drive (ohmic-heating) systems and to achieve a given coupling of the current-drive system with the plasma. A compendium of mutual and self-inductance formulas as they apply to tokamak systems is presented, as well as examples of how the method hasmore » been used in the design of several tokamaks. Finally, a user manual for a computer code that implements this method is provided. 14 refs., 11 figs., 1 tab.« less
  • Irregularities in the winding or alignment of poloidal or toroidal magnetic field coils in tokamaks produce resonant low m, n = 1 static error fields. Otherwise stable discharges can become nonlinearly unstable, and locked modes can occur with subsequent disruption when subjected to modest m = 2, n = 1 external perturbations. Using both theory and the results of error field/locked mode experiments on DIII-D and other tokamaks, the critical m = 2, n = 1 applied error field for locked mode instability in TPX is calculated for discharges with ohmic, neutral beam, or rf heating. Ohmic discharges axe predictedmore » to be most sensitive, but even co-injected neutral beam discharges (at [beta][sub N] = 3) in TPX will require keeping the relative 2, 1 error field (B[sub r21]/B[sub T]) below 2 [times] 10[sup [minus]4]. The error fields resulting from as-built'' alignment irregularities of various poloidal field coils are computed. Coils if well-designed must be positioned to within 3 mm with respect to the toroidal field to keep the total 2,1 error field within limits. Failing this, a set of prototype correction coils is analyzed for use in bringing 2,1 error field down to a tolerable level.« less