Thermal-hydraulic analysis of a high-pressure helium-cooled shield/blanket for ITER

Author

Bourque, R.F. ; Wong, C.P.C.

Author_Institution

Gen. Atomics, San Diego, CA, USA

Volume

1

fYear

1993

fDate

11-15 Oct 1993

Firstpage

273

Abstract

A helium-cooled blanket/shield for ITER is presented that can provide high-grade heat and tritium self-sufficiency. It consists of narrow and relatively simple canisters filled with static liquid metal which is cooled by high pressure helium flowing through small double-walled tubes immersed in the liquid metal. The gaps between the tubes are also filled with static liquid metal. There are therefore three barriers between high-pressure helium and vacuum. Thermal-hydraulic analyses are presented that show the concept to be viable with both ferritic steel and vanadium alloy and with lithium and NaK liquid metal

Keywords

Tokamak devices; fusion reactor blankets; fusion reactor design; ITER; Li liquid metal; NaK liquid metal; V alloy; canisters; double-walled tubes; ferritic steel; high-grade heat; high-pressure He-cooled shield/blanket; liquid metal immersion; static liquid metal; thermal-hydraulic analysis; tritium self-sufficiency; Coolants; Cooling; Heat transfer; Helium; Impurities; Iron alloys; Lithium; Plasma welding; Steel; Temperature control;

fLanguage

English

Publisher

ieee

Conference_Titel

Fusion Engineering, 1993., 15th IEEE/NPSS Symposium on

Conference_Location

Hyannis, MA

Print_ISBN

0-7803-1412-3

Type

conf

DOI

10.1109/FUSION.1993.518329

Filename

518329