In a next step D/T fusion device like ITER, an intense neutron flux will be produced as a consequence of the nuclear fusion reactions. The effects of the neutron induced damage in the microstructure of the plasma-facing material (PFM) may significantly change the thermal properties and the mechanical properties as well as the behaviour of the swelling and the tritium retention in such materials. In addition, a peak heat flux as high as 20 MW m-2and a plasma flux of 1018-1020cm-2s-1are expected in the divertor zone during the normal operation of the reactor. The divertor materials have to withstand the neutron damage, the high heat fluxes and the high erosion caused by the interaction with the high flux plasma. The sputtered particles are co-deposited with plasma, which may contribute significantly to the total tritium inventory in the PFM. Furthermore, the interaction of steam with the sputtered particles (with usually high specific surfaces) could produce large amounts of hydrogen. All of the above topics represent critical issues for plasma performance, safety and economy, as they could limit the use of some PFM materials in next generation fusion devices. Therefore, substantial R&D effort is needed to elucidate the effects of the neutron induced damage on microstructure, erosion/deposition, tritium retention and dust formation, as well as on hydrogen production. In the framework of the European Fusion R&D program, an extensive effort on neutron effects of the material properties: namely, thermal conductivity, mechanical properties, dimensional stability, tritium trapping, erosion/deposition, co-deposition, dust formation/removal, chemical reactivity with steam and oxygen, outgassing, baking and tritium removal from PFM have been undertaken during the past several years. In this paper, the recent progress achieved within the European Fusion R&D program and contributions to the development of ITER PFMs are presented and critically discussed. © 2001 Elsevier Science B.V. All rights reserved.
All Science Journal Classification (ASJC) codes
- Civil and Structural Engineering
- Nuclear Energy and Engineering
- Materials Science(all)
- Mechanical Engineering
Wu, C. H., Alessandrini, C., Bonal, J. P., Davis, J. W., Haasz, A. A., Jacob, W., ... Würz, H. (2001). Progress of the European R&D on plasma-wall interactions, neutron effects and tritium removal in ITER plasma facing materials. Fusion Engineering and Design, 56-57, 179 - 187. https://doi.org/10.1016/S0920-3796(01)00255-1