Article ID Journal Published Year Pages File Type
7987450 Nuclear Materials and Energy 2017 6 Pages PDF
Abstract
As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1 kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513 K for the FW and 623 K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.
Related Topics
Physical Sciences and Engineering Energy Nuclear Energy and Engineering
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