Oxide layer formation in reduced activation ferritic steel F82H under DEMO reactor blanket condition

Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic steels have been investigated; however, oxidation of the steel surface is expected under an actual DEMO reactor condition, and then the tritium permeation behavior will be changed. In this study, deuterium permeation through the steels heat-treated under simulated environment conditions has been investigated for more precise predictions of tritium loss at DEMO reactor blankets. Reduced activation ferritic/martensitic steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen at 300, 400 and 500 °C for 100 and 200 h to simulate a solid breeder DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. An iron oxide layer was formed on the sample surface, and the thickness of the layer was 50 nm‒12 μm. The oxide layer on the sample surface heat-treated at 500 °C for 100 h decreased deuterium permeation by a factor of 5. After the permeation tests, dissipation of the oxide layers was confirmed.