| Deuterium(D)-tritium(T)nuclear fusion reaction is the easiest one to be realized on the earth.In fusion reactors,a large amount of D,T plasmas bombard the plasmafacing components directly,inducing the penetrate of D and T atoms through the fusion materials.The T permeation behavior will not only interfere with the evaluation of T production in the test blanket module(TBM),but also generate a large amount of radioactive coolants such as tritiated water or T-helium(He).Therefore,it is important to determine the performance changes of the in-vessel components and materials under various service conditions and analyze their effects on the T permeation behavior.It is useful for accurately evaluating the T production rate of the TBM and the T loss in invessel components.In this thesis,the laboratory plasma-driven permeation facility is used to analyze the mechanism of the evolution of surface conditions on the permeation behavior of hydrogen isotopes in fusion reactor materials and key components during service is carried out,combined with scanning electron microscopy,transmission electron microscopy,X-ray diffraction,X-ray photoelectron spectroscopy,elastic recoil analysis and spectral analysis.The fusion materials involved in this paper include low-activation ferrite/martensitic low-activation(RAFM)steels(such as CLF-1 steel and CLAM steel),fifth subgroup metals such as vanadium(V),and copper alloys(CuCrZr).In-vessel components are mainly aimed at divertor monoblocks and divertor membrane pump systems.Firstly,based on the working condition of plasma exposure in fusion reactor,RAFM steel is used as the object material and the damage of D and He plasma exposure on RAFM steel is studied in this thesis.The mechanism of different plasma on the surface morphology and steady-state D penetration of the samples is explored.The experimental results show that the surface morphology of RAFM steel has various convex nanostructures such as tentacles,labyrinths,corals,and cones after being exposed by D or He plasma.In addition,the He plasma exposure caused a large number of He bubbles in the range of～100 nm under sample surface.The surface roughness increases with the increase of the plasma fluence,and the steady-state D permeation flux decreases with the increase of the plasma fluence.The surface morphology has a direct impact on the incident flux of D plasma.It can be calculated by the SRIM code that the incident angle of the ions bombarding the surface decreases,the reflection amount of D ions on the bombarding surface increases,and the penetration amount decreases.It can be seen from the simulation using TMAP code that the thickness of the near-surface He irradiation damage layer is very thin compared to the whole sample,and the slowing of the diffusion rate of D atoms in the damaged layer has no significant effect on the steady-state D penetration of the whole sample.But the He bubbles could occupy the intragranular and grain boundaries,and hinder the diffusion of uncaptured D atoms into the interior.The complex undulating topography of the surface and the He bubble layer both inhibit the penetration of D atoms.Secondly,in view of the common oxidation behavior in the reactor,this thesis attempts to understand the influence mechanism of surface oxidation on hydrogen isotope permeation.Two control experiments of oxidized sample and unoxidized sample,single RAFM steel and RAFM-FeCrAl welded composite structural steel were set up to study the effect of surface oxidation on the D plasma-driven permeation.Experiments show that the existence of the oxide layer on the downstream surface of the RAFM steel has no significant effect on the D penetration through the RAFM steel.Under the same oxidation conditions,the hydrogen isotope penetration of the oxidized RAFM-FeCrAl is 3-4 lower than that of the unoxidized RAFM steel sample.The reason is that the oxide layer formed on RAFM steel has low density and a large number of holes,while FeCrAl is easy to form a dense oxide layer.So,whether the oxide layer can play a good role in inhibiting penetration depends on the density of oxide layer.If the structural material faces the cooling water surface to form a complete and dense oxide layer,it can also be a good tritium barrier material.Exploring the optimal oxidation process of RAFM steel or preparing RAFM-FeCrAl type composite structural steel can provide engineering application technology for the preparation of tritium barrier components in the future.Finally,the D permeation of two real component modules,the divertor monoblock and the divertor membrane pump,are tested in this thesis.The effects of the module surface conditions on the D plasma-driven permeation are discussed.The tungstencopper monoblock/plate design used in the plasma-facing component of divertor.There is a 0.4mm gap between the tungsten tiles on the module.The structural material and heat sink material at the bottom of the gap can not be bombarded by plasmas directy covered with tungsten tiles,but the penetration amount of W/CuCrZr and W/CLF-1 covered with tungsten tiles was found to be comparable to that of the bare CuCrZr and CLF-1 not covered with tungsten tiles in this thesis.It is speculated that the magnetic field-confined neutral particles are the main driving factor for initiating the permeation of the W/CuCrZr and W/CLF-1 module samples.The impurities between the gaps of the tungsten tile also have an important influence on the deuterium permeation behavior of the module sample.When the gaps contain silicon impurities,the deuterium permeation speed of the module sample is very slow and cannot be balanced for a long time.However,the amount of deuterium penetration can quickly reach equilibrium in module sample with clean gaps.For membrane pumps with high hydrogen permeability,the presence of a non-metallic impurity layer on the surface of the metal membrane can greatly increase the amount of hydrogen isotope atom/plasma-driven permeation.The experimental results show that there are differences in the resistance to plasma exposure of the three materials of the fifth subgroup,vanadium,niobium and tantalum,among which the niobium film has the best radiation resistance and is a potential separation material.In summary,this thesis(1)performs a real-time study on the effect of He plasma exposure on D plasma-driven permeation through RAFM steel.The influence of the surface nanostructures and He bubbles generated by the low-energy He plasma exposure on the D plasma-driven permeation has been estimated by combined with theoretical calculations to quantitatively analyze the low-energy He plasma exposure.(2)The plasma-driven permeation of RAFM-FeCrAl is performed.The inhibitory effect of oxide layer on downstream surface has been clarified.The thesis confirmed the prospect of the composite structural steel for the preparation of tritium barrier components;(3)The influence of the impurities between the gaps of the tungsten tile on the D penetration behavior of the module is found.A new idea is given for suppressing the short-cut penetration of tritium in the gaps of fusion reactors in the future.These findings have good reference value for the component design of fusion reactors. |
