Study On The Permeation And Retention Behavior Of Hydrogen Isotopes In Fusion Blanket Structural Materials

Reduced activation ferritic/martensitic(RAFM)steels and vanadium alloys are the main candidates for the structural materials of fusion blanket,however,there are concerns about the high hydrogen isotope permeability and inventory for both of these two kinds of materials.Hydrogen isotope permeation and retention behavior in the structural materials of fusion blanket relates directly to the economy and safety of the reactor and affects whether tritium self-sufficiency is practical.In this study,the hydrogen isotope permeation and retention parameters of the CLF-1 steel have been obtained by a gas-driven permeation(GDP)device and a thermal desorption spectrometry(TDS)device.Given the fact the service environment factors such as high energy neutron irradiation,tritium permeation barrier and plasma exposure could influence the hydrogen isotope transport behavior in blanket structural materials,the permeation and retention behavior of hydrogen isotopes in fusion blanket structural materials has been systematically investigated using high-energy heavy ion irradiation devices,linear plasma generators,the material and plasma evaluation system(MAPES)in EAST tokamak and various materials characterization methods.First,the hydrogen isotope permeation parameters of the CLF-1 steel have been obtained,which has been selected as the structrual materials for Chinese Test Blanket Module(TBM)of ITER.The effects of irradiation on deuterium permeation and retention behavior of the CLF-1 steel have been studied using 3.5 MeV He ions irradiation.Results show that the permeation parameters of the CLF-1 steel are similar to those of other RAFM steels such as F82H and EUROFER 97.Low dose of He ions irradiation has tiny effect on the deuterium permeability and diffusion coefficient of the CLF-1 steel,while high dose of helium ions irradiation reduces the permeability and diffusion coefficient and increases the retention significantly.This study can enrich the basic material database of the CLF-1 steel and supply basic data for the tritium transport calculation for ITER and CFETR.Second,based on the experimental data,a Fe-Cr-Al ferritic steel has been proposed as a candidate for tritium permeation barrier,the oxidation and deuterium permeation behavior of this steel has been investigated.After thermal oxidation in air at 800? for 90 h,an oxide layer is observed on the surface of the steel with a thickness of-150 nm-450 nm.The main composition of the oxide layer is confirmed to be Al2O3,and ?-Al2O3 has been detected in the oxide.GDP experiments show that the deuterium permeability of oxidized Fe-Cr-Al alloy is 2-3 orders of magnitude less than that of the CLF-1 steel.To evaluate the effects of irradiation on the permeation behavior of the oxidized Fe-Cr-Al steel,high-energy Au ions have been used to introduce irradiation damages into the oxide.The doppler broadening spectrometry of positron annihilation experiments shows that the density of vacancy-type defects increases with the increase in the irradiation dose.GDP experiments show that different irradiation doses result in different degrees of increase in deuterium permeability.This study provides reference to the development of the new tritium permeation barrier materials,and the investigation of irradiation effects on the permeation behavior of tritium permeation barriers fills the gaps in the relative research fields in China.Finally,the deuterium retention behavior of the F82H steel,V-5Cr-5Ti alloy after exposure to EAST plasmas has been investigated employing MAPES.The effect of tungsten armor has been included.In addition,surface modification of the F82H steel induced by helium plasma exposure has been studied and sequent effects on hydrogen retention have been also characterized.After exposure to EAST deuterium plasmas,four desorption peaks appear in the TDS spectra of the F82H steel,which is significantly different from that obtained after ion/plasma irradiation in the laboratory.After the same exposure,the deuterium inventory in V-4Cr-4Ti is much larger than that in the F82H steel.For the F82H steel and V-4Cr-4Ti samples with thin tungsten layer,after exposure in EAST,large area of W layer exfoliated,and big blisters could be seen between the remaining W layer and the substrate materials.The deuterium inventory of the samples with tungsten layer is smaller than that of the bare samples.After exposure to helium plasma,a lot of surface modifications are found,including tendril-like features in maze-like pattern,densely-distributed pinholes and voids with different sizes in the cross-section of the samples.The mechanism of the modifications has been investigated and the effects of helium plasma exposure on the hydrogen retention have been evaluated.This dissertation has evaluated the possibility to use bare blanket structural materials as plasma facing materials,and confirmed that the gap between tungsten and structural materials could be accumulation area for hydrogen isotopes by experiments.The data obtained can provide reference to the design of the first wall of the fusion blanket.In addition,the retention data of hydrogen isotopes,especially the data after plasma exposure in real tokamak environment,can enrich the database of fusion blanket structural materials.