Estimation of the contribution of gaps to tritium retention in the divertor of ITER

D. Matveev, A. Kirschner, K. Schmid, A. Litnovsky, D. Borodin, M. Komm, G. Van Oost, U. Samm

Research output: Contribution to journalConference articlepeer-review

4 Scopus citations


An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s-1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas.

Original languageEnglish
Article number014063
JournalPhysica Scripta
StatePublished - 2014
Externally publishedYes
Event14th International Conference on Plasma-Facing Materials and Components for Fusion Applications, PFMC 2013 - Julich, Germany
Duration: 13 May 201317 May 2013


  • ITER
  • castellated surfaces
  • divertor
  • gaps
  • impurity deposition
  • tritium retention


Dive into the research topics of 'Estimation of the contribution of gaps to tritium retention in the divertor of ITER'. Together they form a unique fingerprint.

Cite this