Under a carbon-based divertor operation of the International Tokamak Experimental Reactor, tritium retention by codeposits in hard-to-reach areas represents an important concern, and an urgent need exists for cleaning techniques suitable for the in situ removal of these codeposits. The present design of divertor modules includes castellated structures, with narrow gaps prone to accumulate material. On the other hand, glow-discharge cleaning in He/O-2 mixtures is today one of the main candidates for suppressing tritiated deposits in plasma facing components. In the present work, the effectiveness of this technique for removing a-C: H films in gaps is addressed. The authors show that a higher-than-expected efficiency, as compared to the penetration ability of plasma ions, can be obtained. Optimization of the plasma parameters for complete oxidation of the films has also been performed. Erosion rates above 3 nm/min at room temperature can easily be achieved in fully exposed films deposited in laboratory plasmas. (c) 2007 American Vacuum Society.