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Accueil > Séminaires > Séminaires passés > Séminaires de 2009 > Materials for DEMO : Status, Achievements & Future Objectives

Jean-Louis Boutard

Materials for DEMO : Status, Achievements & Future Objectives

Mardi 06 janvier 2009, à 16h00

The decision of constructing ITER has opened the perspective for DEMO, a fusion reactor demonstrating the feasibility of the thermo-nuclear fusion energy production. The selected D-T fusion reaction releases one 14.03 MeV neutron and one 3.56 MeV helium. Elements of design of the main in-vessel components of a fusion power plant, i.e. tritium-breeding blanket, divertor and first wall, will be presented. The structural materials for these components will have to withstand high doses 100 dpa and production of transmutation elements such as He ( 10 appm/dpa) and H ( 45 appm/dpa) induced by the 14.03 MeV neutrons. In addition the divertor will have to undergo high heat fluxes 10 MW/m2.

The irradiation by the 14.03 MeV neutrons will affect the materials at the atomic scale : (i) the crystalline structure is locally destroyed by displacement cascades, (ii) the chemical bonds are strained by He and H transmutation products, and (iii) radiation induces microstructure changes controlled by point defects and impurities diffusion. In the absence of an intense 14.03 MeV neutron source various irradiation techniques are used : (i) alpha particles implantation, (ii) irradiation in fast neutron spectrum or mixed spallation-neutron spectrum, (iii) ion beam irradiation in dual or triple beam configuration, to assess the radiation effects on the in-service properties in the future fusion reactors. The main issues concerning the relevance of these techniques to simulate 14.03 MeV neutron radiation effects will be discussed.

For Tritium–Breeding Blankets, Reduced Activation (RA) 9 % Cr ferritic martensitic steels for temperatures up to 550 0C and Oxide Dispersion Strengthened (ODS) ferritic steels up to 750 0C have been selected on the basis of their well known metallurgy and high resistance to neutron irradiation in fast reactors. For the divertor, high thermal conductivity W-alloys are to be used as structure and protection to withstand the high heat flux of 10 MW/m2. The most significant experimental results about point defect & He accumulation and phase stability, which control the hardening and embrittlement of ferritic martensitic steels, will be presented. The issues concerning the initial fracture toughness and in–service phase stability of the W-alloys will be underlined, and, development of improved alloys presented.

Qualification of these materials should be carried out in the future International Fusion Material Irradiation Facility (IFMIF) based on D-Li reaction producing a neutron spectrum very similar to the D-T fusion one. The main characteristics of IFMIF will be presented.
Physical modelling of radiation effects is also an important part of the fusion materials development, (i) to understand & inter-correlate the results obtained under the various damage spectra, (ii) to extrapolate these data to the fusion neutron spectrum, (iii) to provide guideline to develop materials with improved heat and radiation resistance, and (iv) to optimise the IFMIF programme and extrapolate its future data to the DEMO conditions. Example of physical prediction of the energetics of point defects and He in Fe & Fe-C based on DFT calculations will be given as well as kinetic modelling of damage recovery and He – desorption in Fe & Fe-C.

Since January 2008, the EFDA programme is organised in a Fusion Materials Topical Group encompassing four research projects : (i) Radiation effects modelling and experimental validation, (ii) W-alloy development (iii) ODS ferritic steel development and (iv) SiCf/SiC composite for structural applications. Seventeen laboratories among the twenty-two within the Fusion EURTOM are contributing to the Topical Group.