Design assessment of ITER port plug plasma facing material options

The objective of this computational engineering design study is to assess whether the ITER confined plasma will be adversely affected if the diagnostic port plug and tritium breeding module plasma-facing surfaces are left as bare stainless steel or armoured with W, rather than with Be as on the rest of the main chamber first wall. The OSM-EIRENE-DIVIMP code package is employed to determine the 2D steady-state impurity distribution, after benchmarking the OSM plasma calculations against reference SOLPS simulations. For far-SOL transport, the computational domain is extended to explicitly include plasma contact with the wall. To sample a large area of the foreseen ITER parameter space, a range of boundary plasmas are assigned via OSM based on observed experimental trends, including radial decay lengths, parallel flows, and pedestal profiles. Taking core impurity limits from ASTRA simulations, the results indicate that Be cladding of the port plugs under consideration is not required, with the proviso that neutral particle injection directly in front of the ports is avoided. (C) 2011 EURATOM. Published by Elsevier B.V. All rights reserved.