Study On The Interaction Of Tungsten And Its Alloys With Deuterium Plasma With Low Energy And High Beam Current

Deuterium(D)-tritium(T)thermonuclear fusion has the potential to realize sustainable energy production on a large scale.The International Thermonuclear Experimental reactor(ITER)and the China Fusion Engineering Experimental reactor(CFETR)under construction are magnetically confined plasma fusion devices to verify the feasibility of deuterium-tritium fusion energy.An important component of ITER and CFETR is the divertor,which removes helium and impurities from the fusion reaction from the plasma beam.The divertor must be able to withstand heat flux of 10 MW·m-2 and particle bombardment of up to 1024 m-2 s-1.Tungsten(W)is one of the best candidate materials for divertor and first wall because of its high hydrogen sputtering threshold and good thermal performance.The study of the interaction between tungsten and its alloys and plasma and the quantitative measurement of hydrogen isotope retention are of great significance for the research and development of plasma wall-oriented materials and the safe operation of the device.In this work,Comprehensive ECR Plasma for Tritium(CEPT),developed by our research group was used to test the radiation resistance and deuterium retention of pure tungsten prepared by different processes under the irradiation conditions of low energy(25 eV/D)and high beam current(3.71 ×1021 D·m-2·s-1).The surface radiation damage evolution and the change of deuterium retention of pure tungsten under long-term continuous irradiation were studied.The effects of Y2O3 doping on the radiation resistance and deuterium retention of pure W were investigated.The results show that:(1)The surface morphology and deuterium retention behavior of three kinds of W materials prepared by different processes-chemical vapor deposition(CVD-W),powder metallurgy(cold rolling)(Anti-W)and powder metallurgy(hot rolling)(Xia-W)are obviously different under the same deuterium plasma irradiation parameters(matrix temperature 370 K,irradiation dose 1.3× 1025 D/m2).After irradiation of three kinds of W materials,there are a large number of spherical blisters on the surface of Anti-W,a small amount of platform blisters on the surface of Xia-W,and almost no blistering on the surface of CVD-W.The X-ray diffraction(XRD)peaks of the irradiated materials all shifted to a small angle,and the changes of lattice constant and lattice expansion from large to small were in the following order:CVD-W(0.54%),Xia-W(0.37%)and Anti-W(0.28%).The(TDS)measurement of thermal desorption spectra showed that among the three samples,Anti-W had the highest deuterium retention(9.11 × 1020/m2),followed by Xia-W(1.44 × 1020/m2),and CVD-W had the lowest deuterium retention(0.41 × 1020/m2).It can be seen that the lattice expansion and the reduction of deuterium retention alleviate the macroscopic foaming behavior on the sample surface to some extent.The deutero-thermal desorption activation energies of Anti-W,CVD-W and Xia-W are 1.57 eV,0.88 eV and 0.53 eV,respectively.It is further revealed that deuterium may exist in vacancies and deuterium bubbles in Anti-W,in grain boundaries/dislocations in CVD-W,and mainly in lattice gaps in Xia-W.Compared with the comprehensive properties,the CVD-W prepared by chemical vapor deposition has good radiation resistance,and the deuterium retention is much lower than that of pure tungsten prepared by traditional powder metallurgy process.(2)When the deuterium ion is irradiated continuously for a long time,the deuterium retention in Anti-W will continue to increase with the increase of radiation dose(8h,the retention is 1.89 ×1021/m2),but the deuterium desorption temperature increases at first and then decreases(up to 850K).Nuclear reaction recoil(NRA)measurements show that deuterium atoms are mainly enriched at 1500 nm from the sample surface,and deuterium enrichment in tungsten leads to surface blistering,which is related to tungsten lattice expansion,which decreases when the surface blistering is obvious.After long-time irradiation,there are mainly two kinds of deuterium traps in tungsten:vacancy is dominant in 1.33 ×1025 D·m-2 dose range,thermal desorption activation energy is 1.09-1.16 eV;in 1.33-21.36 ×1025 D·m-2 dose range is mainly vacancy clusters,thermal desorption activation energy is 1.67-1.77 eV.Long-time irradiation is the actual working condition of the first wall material.Under the synergistic action of irradiation and radiation-induced sample temperature rise,vacancies will gather into vacancy clusters and become traps for further trapping deuterium.The continuous increase of deuterium molecular pressure in the vacancy clusters will eventually lead to the rupture of the bubbles on the sample surface,resulting in tungsten dust entering the plasma discharge area and causing the plasma to cool down,so it is necessary to take appropriate measures to prevent the blistering behavior on the W surface or remove the hydrogen isotopes retained in W in time.(3)Under the same deuterium plasma irradiation parameters,a small amount of spherical bubbles(about 2μm)appeared on the surface of W-Y2O3 and accompanied by oval expansion(～2 × 10 μm),)while the surface of pure W was covered with a large number of small spherical bubbles(～100 nm).The lattice expansion rate of W-Y2O3 is 0.35%,which is higher than that of pure W(0.19%).The deuterium retention in W-Y2O3 is 5.23 × 1020 D/m2,which is slightly higher than that in pure W(2.10 ×1021 D/m2),and the thermal desorption temperature range of deuterium in W-Y2O3 is 450 K-850 K.There are two kinds of deuterium capture sites in both W-Y2O3 and pure W,in which the thermal desorption activation energy of deuterium capture site at low temperature is 0.39 eV,which is lower than that of pure W(1.73 eV)and higher than that of pure W(1.57 eV).Y2O3 doping improves the radiation resistance of tungsten to some extent,but the deuterium retention increases,the desorption temperature extends to high temperature,and the difficulty of deuterium removal increases.