Tritium retention in next step devices and the requirements for mitigation and removal techniques
Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required.
Bibliographic Reference: An article published in: Plasma Physics and Controlled Fusion 48 (2006), pp. B189-B199
Availability: This article can be accessed online by subscribers, and can be ordered online by non-subscribers, at: http://dx.doi.org/10.1088/0741-3335/48/12B/S18
Record Number: 200719128 / Last updated on: 2007-04-11
Original language: en
Available languages: en