MHD equilibrium methods for ITER (International Thermonuclear Experimental Reactor) PF (poloidal field) coil design and systems analysis

Technical Report · Wed Mar 01 00:00:00 EST 1989

DOI:https://doi.org/10.2172/6244555· OSTI ID:6244555

Strickler, D J; Galambos, J D; Peng, Y K.M.

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.

• Research Organization: Oak Ridge National Lab., TN (USA). Fusion Engineering Design Center
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6244555
• Report Number(s): ORNL/FEDC-88/7; ON: DE89012902
• Resource Relation: Other Information: Portions of this document are illegible in microfiche products
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
POLOIDAL FIELD DIVERTORS
DESIGN
TOKAMAK TYPE REACTORS
SYSTEMS ANALYSIS
CALCULATION METHODS
COST
MHD EQUILIBRIUM
OPTIMIZATION
PERFORMANCE
DIVERTORS
EQUILIBRIUM
THERMONUCLEAR REACTORS
700202* - Fusion Power Plant Technology- Magnet Coils & Fields