SYMPOSIA PAPER Published: 01 January 2000

Effect Of Irradiation On Structural Materials In Fusion Reactors


The ITER R&D program on the structural design criteria includes the effects of irradiation on material's properties. The work has started with the design rules for stainless steels at low temperatures (< 350°C) and at moderate irradiation doses (3 – 10 dpa), where degradation of ductility and toughness are of primary concern. It is now extended to copper alloys and multi-layer components, as well as to rules for high temperature properties. A similar work has also been initiated on ferritic / martensitic steels for the future Demonstration reactor, where the operating temperature range is 250–550°C and anticipated irradiation doses are well over 70 dpa.

This paper presents an overview of the above work. It details some of the recent irradiation data obtained for stainless steel type 316LN, precipitation strengthened CuCrZr alloy, oxide dispersion strengthened Cu-A125 alloy, and commercial and low activation grades of 7–10%Cr ferritic / martensitic steels. Since most of the fusion invessel components have to be fabricated using advanced manufacturing techniques, e.g. high temperature isostatic pressing (HIP) of steel/copper/beryllium layers in ITER primary modules, the paper also analyzes effects of such processes on the materials properties and design data.

Author Information

Tavassoli, F
CEA-CEREM, CEA/Saclay, Gif sur Yvette, France
Burlet, H
CEA-CEREM, CEA/Grenoble, Grenoble, France
Lind, A
Studsvik Material AB, NYKÖPING, Sweden
Singh, BN
Riso National Laboratory, Roskilde, Denmark
Tähtinen, S
VTT Manufacturing Technology, Materials and Structural Integrity, Finland
van Osch, E
Netherlands Energy Research Foundation ECN, ZG Petten, The Netherlands
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Developed by Committee: E10
Pages: 1093–1108
DOI: 10.1520/STP12455S
ISBN-EB: 978-0-8031-5419-3
ISBN-13: 978-0-8031-2852-1