document: The main drawback of thermo-oxidation in most actual devices and ITER is its limitation to maintenance periods, when the vessel walls can be heated up around 300–400 °C by hot helium injection through the cooling system [19,20], and also because of the required reconditioning of the walls before plasma operation to remove the absorbed oxygen [10]. However, the temperature achieved is not homogeneous over the vessel, as it is limited to the distance to the cooling tubes, and thus to the device design. The analysis of this study is a continuation of previous works done for the treatment of ITER carbon co-deposits [1–3], so the temperatures studied are in the range of 350 °C for divertor and 200–275 °C for main wall and remote parts. At present, due to budget restrains as well as due to tritium trapped in co-deposited carbon layers, ITER will not use carbon materials at the divertor strike points in spite of their excellent resilience against large heat loads. Nevertheless, many present experimental nuclear fusion devices (DIII-D, TCV, etc.) and new ones (JT-60SA, KSTAR, Wenderstein-7X) use carbon elements, so the removal of carbon co-deposits is still necessary for a better device operation—plasma density control, dust events, etc. The temperatures used in this work are not very different from the ones achievable in present devices, such that the results can be extrapolated to them. Moreover, even for ITER this study could be useful if carbon materials have to be eventually installed in the case that operation with tungsten tiles at the strike points is precluded by unexpected reasons.