Experimental And Modeling Study Of Divertor Power Footprint Width

The tokamak magnetically confined fusion is considered to be the most promising way for human to gain the ability to peacefully utilize the fusion energy.For future tokamaks,one of the biggest challenges is the exhaust of fusion energy from the core plasmas efficiently and safely.The divertor structure,as the main plasma facing component,would definitely have to face a significant amount of heat flux from the core plasmas.Although present tokamaks like EAST have upgraded their divertors from the graphite structures to the tungsten structures with better corrosion resistance from the heat load.But for future tokamaks,the heat load on the divertors will be way more than that of the present tokamaks.The erosion of divertor target materials seems inevitable due to the high temperature plasma flow to the targets,so finding ways to mitigate the heat load on divertor targets will be the key to the prosperity of tokamak fusion.There are several ways to deal with the excessive heat load issue on divertor target One way is to increase the scrape-off layer(SOL)and divertor plasma radiation by active impurity seeding,thus a huge amount of energy is been radiated and distributed evenly on the first wall and divertor target before reaching the target But this way might have to sacrifice the core plasma confinement since some of the impurity particles may enter into the core plasmas and give rise to some energy loss of the core plasmas.Another way to mitigate the heat load on divertor target is to expand the plasma-wetted area.One way to expand the plasma-wetted area is by using advanced divertor structures like snow-flake divertor.This way would make the fusion energy too uneconomical since the coils used to generate the advanced divertor configuration at the divertor region are very expensive.Another way to expand the plasma-wetted area is to broaden the divertor power footprint width(also known as the heat flux width or the SOL width).The main topics of our thesis is the studies of the effects of plasma parameters like plasma current and heating scheme on the EAST divertor power footprint width and the prediction simulation of heat flux width for China fusion engineering test reactor(CFETR).The EAST experimental data is based on the divertor Langmuir probes on EAST and the modeling part is based on the two major edge plasma simulation codes SOLPS and BOUT++.The major contents of the thesis are listed below:(1)Experimental and modeling study of the effect of plasma current on the EAST divertor power footprint width.Experimentally,we found that the heat flux width in EAST has strong negative scaling on the plasma current in LHW and NB heated H-mode and L-mode plasmas,which are in consistent with the results on a variety of other tokamaks worldwide.Simulations of the effect of plasma current on the SOL width show that the negative scaling of SOL width on plasma current may be due to the change of connection length.The connection length is negatively correlated to the plasma current.While the longer the connection length is,the longer time the particle will have for radial transport in the SOL and the wider the SOL width will be,which result in the negative scaling of SOL width on plasma current.The simulation has in some way,explained the potential reason to the negative scaling of SOL width on plasma current.(2)Experimental and modeling study of the effect of heating scheme on the EAST SOL width Experimental,we found that the heat flux width in the lower hybrid wave(LHW)heated plasmas is about twice of that in the neutral beam injection(NBI)heated plasmas.More detailed study shows that the SOL width increases with the increase of LHW power ratio.In order to figure out this issue,two steady-state H-mode discharges,one heated by LHW and another heated by NBI,were chosen for the simulation.We found that drifts are the dominant factor in the edge plasma transport for both discharges but the background turbulence tends to be stronger in LHW heated discharge than that in NBI heated discharge,which may be the reason to the larger SOL width in LHW heated discharges.(3)Modeling study of the effect of radial transport on the EAST SOL width.SOLPS and BOUT++transport code were used to simulate the effects of radial particle and heat transport coefficients on the SOL width in EAST.We found that the SOL width increases with the increase of both particle and heat transport coefficients,which is in consistent with the theoretical scaling.(4)Experimental and modeling study of plasma operating regime on the EAST divertor particle flux width.We compared the particle flux width from EAST grassy ELMy and type-Ⅰ ELMy discharges with that from the radial transport coefficient scan simulation of BOUT++transport code.We found that plasmas under different operating regimes have different particle flux width is due to the different intensities of radial background turbulence transport.(5)Prediction of divertor heat flux width for CFETR steady-state discharge through BOUT++ and SOLPS simulations were carried out under the CFETR preliminary divertor structure and configuration.BOUT++ simulation results show that the particle flux width is more than twice of the heat flux width.The heat flux width from BOUT++is about 2.56 mm while it is about 6.01 mm from SOLPS simulation with Ar seeding.Although the heat flux width results from BOUT++and SOLPS failed to agree with each other,both results show that the heat flux width for CFETR is probably much larger than the prediction of the extrapolation from existing experimental scaling.(6)Through the comparison of divertor heat flux width from BOUT++tansport and turbulence codes with that fromthe prediction of heuristic drift-based model,we found that the edge plasma radial transport for CFETR is not dominated by drifts anymore,but is instead dominated by the background turbulence transport,this also explained the reason why the heat flux width of CFETR couldn’t be well predicted by the existing experimental scaling.