Selection of plasma facing materials for ITER

Author

Ulrickson, M. ; Barabash, V. ; Chiocchio, S. ; Federici, G. ; Janeschitz, G. ; Matera, R. ; Akiba, M. ; Vieider, G. ; Wu, C. ; Mazul, I.

Author_Institution

Sandia Nat. Labs., Albuquerque, NM, USA

Volume

1

fYear

1995

fDate

30 Sep-5 Oct 1995

Firstpage

394

Abstract

ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m² for normal operation with transients to 15 MW/m² for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final selection of plasma facing materials for the ITER divertor will not be made until the end of the EDA

Keywords

fusion reactor design; fusion reactor materials; fusion reactors; 10 s; Be; C fiber composites; DT; DT fuel; Engineering Design Activity; ITER; W; divertor components; fusion reactors; heat removal structures; long pulse operation; neutron environment; normal operation; plasma facing materials; steady state; steady state heat flux; tokamak; Composite materials; Design engineering; Electronic design automation and methodology; Fuels; Heat engines; Neutrons; Organic materials; Plasma materials processing; Steady-state; Tokamaks;

fLanguage

English

Publisher

ieee

Conference_Titel

Fusion Engineering, 1995. SOFE '95. Seeking a New Energy Era., 16th IEEE/NPSS Symposium

Conference_Location

Champaign, IL

Print_ISBN

0-7803-2969-4

Type

conf

DOI

10.1109/FUSION.1995.534250

Filename

534250