Speaker
Description
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 (RAFM) steels have been investigated; however, it is supposed that the surface of the RAFM steel will be oxidized under an actual DEMO reactor condition at a tritium recovery system, and then the permeation behavior will be changed. In this study, deuterium permeation through the RAFM steels heat-treated under simulated environment conditions has been investigated for more precise prediction of tritium loss at DEMO reactor blankets.
RAFM steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen for 100 h at 300, 400 and 500 ºC to simulate a DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. Deuterium permeation flux through the sample was detected by a quadrupole mass spectrometer with the driving pressure of 10.0‒80.0 kPa in the temperature range of 250‒550 ºC.
An iron oxide layer was formed on the F82H surface after the heat treatment. From the results of grazing incidence X-ray diffraction analysis, the major composition of the surface oxidation layer was Fe2O3. The oxide layers became thicker as the heat-treatment temperature increased. The thickness of the oxide layers of the samples heat-treated at 300, 400 and 500 ºC was approximately 50 nm, 130 nm and 5 μm, respectively. In the results of permeation tests, the iron oxide layers of all the samples decreased the deuterium permeation down to only 25%, suggesting that the structure of the iron oxide layers were not tight; therefore, the layers naturally formed under the DEMO blanket condition will not be efficient as tritium permeation barriers.
