Toroidal field coil replacement study for the International Thermonuclear Experimental Reactor (ITER)

Technical Report · Thu Aug 01 04:00:00 EDT 1991

OSTI ID:10142070

The objective of this study is to develop an estimate of the time required to replace a failed toroidal field (TF) coil in the International Thermonuclear Experimental Reactor (ITER). This estimate is based on the current ITER design which resulted from the Conceptual Design Activities began in spring of 1988 and ending in December 1990. Although the TF Coils in ITER are designated permanent components and expected to last the life of the plant, the history of failures of large coils in fusion devices has prompted an assessment of the down time that might be required for replacement of a failed TF coil in ITER. The difficulty of replacement of a TF Coil in ITER is compounded by the large physical size of ITER, and the radiation fields which necessitate the use of remote handling equipment for most maintenance activities. Since 10 out of 16 TF coils are adjacent to either a Neutral Beam Injector (NBI) port or a remote handling (RH) port, it is probable that a failed TF Coil will be in one of these inconvenient locations. For this study, TF coil number 3 is assumed to fail and its replacement will be evaluated as a typical case. TF coil number 3 is located between a remote handling port ({number_sign}3) and a port allocated to the lower hybrid ({number_sign}4).

• Research Organization: Lawrence Livermore National Lab., CA (United States); Bechtel Corp., San Francisco, CA (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: W-7405-ENG-48
• OSTI ID: 10142070
• Report Number(s): UCRL-CR--109602; ITER/US--91/EN-ME-1; ON: DE92013010
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
700430
ITER TOKAMAK
MAGNET COILS AND FIELDS
MAINTENANCE
REMOTE HANDLING
REMOTE HANDLING EQUIPMENT
REMOVAL
TOROIDAL FIELD DIVERTORS