رفتار نفوذ دوتریوم از فولاد ضد زنگ 430
Abstract
1- Introduction
2- Experimental method
3- Results and discussion
4- Conclusions
References
Abstract
The deuterium permeation behavior of palladium-coated 316L stainless steel and 430 stainless steel was studied by the gas permeability testing device in this study. The results show that the deuterium permeability of the palladium-coated 316L stainless steel tested at 350 °C∼650 °C was very close to the Forcey’s results. This way, the conformity to some other reported results has been verified for the gas permeability testing device. The deuterium permeability for the oxidized 430 stainless steel was compared with that for the palladium-coated 430 stainless steel at temperature range of 350 °C–600 °C. The deuterium permeability for the oxidized 430 stainless steel was reduced by one order of magnitude compared with that for the palladium-coated 430 stainless steel. However, the activation energy of deuterium permeation as gas form for the oxidized 430 stainless steel was almost the same as that for unoxidized 430 stainless steel.
Introduction
Zirconium hydride is used as the reactor moderator in the thermionic space nuclear power and it easily decomposes hydrogen in the operating environment. The released hydrogen penetrates into the core through the relevant structural components, affecting the physical parameters of the reactor [1]. Compared with 316L austenitic stainless steel, 430 ferritic stainless steel has the advantages of good thermal conductivity, low thermal expansion coefficient and low sensitivity to stress corrosion cracking. It is selected as one of the alternative reactor core materials for structural components of the thermionic space nuclear power. Therefore, investigating the hydrogen permeation behavior of 430 ferritic stainless steel has important engineering application value for the development of thermionic space nuclear power. A lot of research has been carried out on the preparation and properties evaluation of hydrogen isotope permeation barrier coatings for the fusion reactor. However, 316L stainless steel or low-activity ferritic/martensitic steel was used as the substrate material in most of these studies [2–15]. There are few reports on the hydrogen isotope permeation behavior of 430 ferritic stainless steel. In the present study, the deuterium permeability of palladium-plated 316L stainless steel was first tested and compared with the reported values to verify the conformity to other reported results for the gas permeability testing device. Then the deuterium permeation behavior of the palladium-plated 430 stainless steel as well as that with natural oxide film was investigated.