3.5. Materials Development and Test Introduction Materials R&D for ...

Review of the ITER Technology R&D
3.5 Materials Development and Test
fatigue lifetime. Data on the radiation behaviour of Inconel 718 is sparse. Results of
irradiation tests show that irradiation up to 0.5 dpa will result in an increase of strength and in
a slight decrease of ductility that will not affect the component structural integrity and
lifetime [34].
Radiation induced stress relaxation is one of the critical issue for the bolts feasibility. Results
of stress relaxation investigations performed by the EU HT [34, 35, 36] show that for the
bolts of flexible cartridge fastenings where the irradiation dose is ~ 0.02-0.1 dpa, the stress
relaxation will be ~ 80-90% from initial pre-load to the end of life, so, the required pre-stress
should be ~ 800 MPa.
3.5.3 Plasma Facing Materials
The choice of plasma facing materials for different components is determined mainly by
plasma compatibility and erosion lifetime (ion erosion, thermal erosion during disruption and
transient events) [37].
- Beryllium has been chosen as the armour material for the first wall and port limiter,
- Tungsten has been selected as the material for the divertor baffle area with high
concentration of neutral particles due to lowest erosion rate,
- Carbon fibre composites (CFC) have been proposed for the lower part of the divertor
vertical target due to absence of melting during plasma disruptions.
The key issues related to the plasma-wall interaction phenomena have been discussed
elsewhere [37]. In this section the following R&D results will be highlighted:
- evaluation of the different armour material grades and the selection of the reference
grade for the ITER application,
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