Overview of co-deposition and fuel inventory in castellated divertor structures at JET
2007 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 367, 1432-1437 p.Article in journal (Refereed) Published
The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 x 10(15) cm(-2). Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.
Place, publisher, year, edition, pages
Elsevier, 2007. Vol. 367, 1432-1437 p.
DEUTERIUM; CARBON; PENETRATION; DEPOSITION; RETENTION; PLASMA; WALL
Engineering and Technology
IdentifiersURN: urn:nbn:se:kth:diva-25596DOI: 10.1016/j.jnucmat.2007.04.007ISI: 000249083500096ScopusID: 2-s2.0-34447500289OAI: oai:DiVA.org:kth-25596DiVA: diva2:370503
12th International Conference on Fusion Reactor Materials (ICFRM-12) Santa Barbara, CA, DEC 07-12, 2005
QC 201011172012-01-182010-10-272012-02-22Bibliographically approved