In the context of future energy supply, fusion energy promises an economical and environmentally acceptable energy production. It is based on abundant and equally spread fuel, offers high intrinsic safety and does not generate long-lived nuclear waste.
Producing electricity, however, out of a fusion reaction is far from easy, and important technological challenges are still to be solved. This difficulty explains the long time scale along which the development of fusion is being pursued, as well as the high development cost required, in particular for the construction and operation of ITER. Still, the international community has clearly expressed its full commitment to this enterprise. The effort is worthwhile, due to the critical issue at stake: guaranteeing safe energy supply in the long term.
Belgium was part of it from the start, and should remain at the foreground. The construction of ITER offers new opportunities, not only for scientific and R&D institutes, but also for the Belgian industrial capabilities.
More information: 
White Paper on Fusion (11 MB)
Material research represents a crucial issue for the assessment of fusion as a future source of energy. For many aspects of the design, it is indeed material technology that will dictate the most viable concept for the commercial power plant.
Structural materials need to show a superior mechanical and chemical behaviour to guarantee the safe operation of the reactor during its whole lifetime, while retaining low activation characteristics to minimise the environmental impact of the produced waste.
In addition, ITER, DEMO and other next step fusion devices strongly depend on reliable wall components that can sustain the intense particle and heat fluxes from the plasma.
In the field of structural materials, specific efforts have been focused for the last twenty years in Europe, Japan and the US, on developing suitable Reduced Activation Ferritic/Martensitic (RAFM) steels. EUROFER97 has now emerged in Europe as the reference material for the design of DEMO (the prototype reactor that will demonstrate the feasibility of fusion energy production). While the final assessment of this material under the actual reactor conditions will only occur in dedicated machines (ITER, IFMIF), it is presently of primary importance to develop the scientific understanding of the mechanisms that control the physical, mechanical and chemical behaviour of such a material under radiation.
An integrated approach is followed at SCK•CEN to tackle this objective. It combines a series of irradiation campaigns, involving the base alloy, ODS (Oxide Dispersion Strengthened) alternatives and joint samples, with a full characterisation of the mechanical properties and the corrosion behaviour (in water and lithium-lead environments) of EUROFER97 in the unirradiated and irradiated conditions. This experimental work on the real steel is coupled with a more fundamental modelling effort aiming at understanding and predicting the evolution of radiation damage from the atomic scale.
In the field of plasma wall interaction processes, the induced particle fluxes will finally degrade the plasma facing components with respect to their thermal and mechanical properties. On top of that, wall erosion is another critical issue that has significant impact on the lifetime of PFM and on the contamination of the fusion plasma. The eroded material will be transported in the form of dust and will be redeposited in the divertor areas. Deposited material exhibits different mechanical and thermal properties and has its effect on tritium retention and erosion, where mixing of materials could cause problems related to the alloy formation. These materials will also be subject to an intense neutron flux, resulting in the further degradation (embrittlement; bulk production of H, He gas atoms) of their characteristics.
The PFM in ITER are primarily based on beryllium for the first wall, carbon fibre composites and tungsten for the divertor and copper-based alloys as a heat sink. For DEMO, advanced tungsten en tungsten alloys will not only be used for the armour, but also as a structural material in the divertor.
The present R&D at SCK•CEN aims at the characterisation of existing and new developed materials, coatings and joints with respect to their thermal and mechanical properties in a wide temperature range (overheating, embrittlement, fatigue, creep, swelling, erosion, corrosion, tritium and helium retention) before and after neutron irradiation. Basic understanding of the PFM (mainly tungsten) under the severe thermal loads and neutron irradiation are being investigated by microstructural examinations.
Contact person(s)Massaut Vincent , Uytdenhouwen Inge , Lucon Enrico , Chaouadi Rachid , Malerba Lorenzo , Terentyev Dmitry