Thermal hydraulic analysis of the TPX plasma facing components
Conference
·
OSTI ID:465005
- General Atomics, San Diego, CA (United States)
The purpose of the Tokamak Physics Experiment (TPX) is to develop and demonstrate steady state tokamak operating modes that can be extrapolated to reactor conditions. TPX will have a double null divertor with an option to operate in a single null mode. The maximum input power will be 45 MW and the pulse length will be 1,000 s. The major and minor radii will be 2.25 m and 0.5 m respectively. The material of plasma facing components (PFCs) will be carbon fiber composite (CFC). The plasma facing components (PFC) cooling will be provided by water at an inlet pressure of 2 MPa and inlet temperature of 50 C. The heat flux on the PFCs will be less than 0.2 MW/m{sup 2} on line of sight shields to 7.5 MW/m{sup 2} on divertor surfaces. The maximum allowable temperature on the divertor surface is 1,400 C and 600 C on all other PFCs. The attachment method, the type of CFC, the coolant flow velocity and the type of coolant channel is chosen based on the surface heat flux. In areas of highest heat flux, heat transfer augmentation will be used to obtain a safety margin of at least 2 on critical heat flux. An extensive thermal flow analysis has been performed to calculate the temperatures and pressure drops in the PFCs. A number of R and D programs are also in progress to verify the analysis and to obtain additional data when required. The total coolant flow rate requirement is estimated to be about 50 m{sup 3}/min (12,000 gpm) and the maximum pressure drop is estimated to be less than 1 MPa.
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC02-76CH03073
- OSTI ID:
- 465005
- Report Number(s):
- CONF-950905--; ISBN 0-7803-2969-4
- Country of Publication:
- United States
- Language:
- English
Similar Records
THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR
The TPX vacuum vessel and in-vessel components
High-heat-flux testing of helium-cooled heat exchangers for fusion applications
Conference
·
Sun Oct 01 00:00:00 EDT 2000
·
OSTI ID:808789
The TPX vacuum vessel and in-vessel components
Journal Article
·
Mon Oct 31 23:00:00 EST 1994
· Fusion Technology
·
OSTI ID:41429
High-heat-flux testing of helium-cooled heat exchangers for fusion applications
Journal Article
·
Mon Jul 01 00:00:00 EDT 1996
· Fusion Technology
·
OSTI ID:374450