The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power foo...The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer （SOL） width λq and heat spreading S,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip.Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of λq,L-mode =4.98 × Ip-0.68 and λq,H-mode =1.86 × Ip-1.08.Similar trends have also been demonstrated in the study of heat spreading with SL-mode =1.95 × Ip-0.542 and SH-mode =0.756 × Ip-0.872.In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor （CFETR）.展开更多
A detailed study of the divertor performance in EAST has been performed for both its double null （DN） and single null （SN） configurations. The results of application of the SOLPS （B2-Eirene） code package to the ...A detailed study of the divertor performance in EAST has been performed for both its double null （DN） and single null （SN） configurations. The results of application of the SOLPS （B2-Eirene） code package to the analysis of the EAST divertor are summarized. In this work, we concentrate on the effects of increased geometrical closure and of magnetic topology variation on the scrape-off layer （SOL） and divertor plasma behavior. The results of numerical predictions for the EAST divertor operational window are also described in this paper. A simple Core-SOL- Divertor （C-S-D） model was applied to investigate the possibility of extending plasma operational space of low hybrid current drive （LHCD） experiments for EAST.展开更多
The HIT-PSI is a linear plasma device built for physically simulating the high heat flux environment of future reactor divertors to test/develop advanced target plate materials.In this study,the geometry-modified SOLP...The HIT-PSI is a linear plasma device built for physically simulating the high heat flux environment of future reactor divertors to test/develop advanced target plate materials.In this study,the geometry-modified SOLPS-ITER program is employed to examine the effects of the magnetic field strength and neutral pressure in the device on the heat flux experienced by the target plate of the HIT-PSI device.The findings of the numerical simulation indicate a positive correlation between the magnetic field strength and the heat flux density.Conversely,there is a negative correlation observed between the heat flux density and the neutral pressure.When the magnetic field strength at the axis exceeds 1 tesla and the neutral pressure falls below 10 Pa,the HIT-PSI has the capability to attain a heat flux of 10 MW·m-2 at the target plate.The simulation results offer a valuable point of reference for subsequent experiments at HIT-PSI.展开更多
A major challenge facing the steady-state operation of tokamak fusion reactors is to develop a viable divertor solution with order-of-magnitude increase in power handling capability as compared with present experience...A major challenge facing the steady-state operation of tokamak fusion reactors is to develop a viable divertor solution with order-of-magnitude increase in power handling capability as compared with present experience.A recently developed divertor concept for this end has been tested recently on EAST tokamak through combining the effects of a closed divertor corner and E×B drifts.The E×B drifts in the divertor move particles towards the outer divertor corner area in the scrape-off layer for B×▽B directed away from the divertor,which can significantly enhance the particle concentration there,facilitating divertor detachment.In recent EAST experiments,the effects have been demonstrated where the lowest electron temperature at the divertor plate is obtained with strike point located close to the corner in the horizontal target and with B×▽B away from the divertor.These experimental results are in reasonable agreement with SOLPS-ITER simulations including drift effects,suggesting that the new divertor concept potentially provides a promising divertor solution for long-pulse operations of future tokamak fusion reactors with much higher power fluxes.展开更多
The edge transport code SOLPS5.0 is used to model edge plasmas in the experi- mental shots on JT-60U and the profiles of the transverse particle and heat transport coefficients D, Xe and Xi along the outer midplane ar...The edge transport code SOLPS5.0 is used to model edge plasmas in the experi- mental shots on JT-60U and the profiles of the transverse particle and heat transport coefficients D, Xe and Xi along the outer midplane are obtained by fitting the simulational results to the experimental data in L-mode shot 39090 and H-mode shots 37851, 37856. The modelling and fitting results show that within the pedestal region in H-mode shots 37851 and 37856 the radial particle transport coefficient D exhibits a significant drop, but, for L-mode shot 39090, the obvious drop in both D and Xe was not found.展开更多
A simulation on HCSB-DEMO （helium-cooled solid breeder, HCSB） edge plasma, by using 2D edge plasma transport code SOLPS5.0, is presented. There is 400 MW heat power crossing CIB （core interface boundary）. The heat...A simulation on HCSB-DEMO （helium-cooled solid breeder, HCSB） edge plasma, by using 2D edge plasma transport code SOLPS5.0, is presented. There is 400 MW heat power crossing CIB （core interface boundary）. The heat flux profiles and peak flux at the divertor targets with different boundary densities axe investigated. It is indicated that the HCSB-DEMO divertor should operate at a proper upstream density in order to avoid a high heat flux at the divertor targets. When the upstream density is 0.63x 102~ m-3, the peak heat flux at the divertor targets will be above 17 MW/m2. The cross-field transport, the power crossing CIB and the power fraction taken by electrons and ions and SOL （scrape-off-layer） thickness are analyzed as unknown parameters. It is shown that the peak flux at the divertor target is very sensitive to these parameters. The simulation results will be used in the divertor design for HCSB-DEMO.展开更多
One of the critical issues to be solved for HL-2M is the power exhaust.Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern,which in turn has a strong eff...One of the critical issues to be solved for HL-2M is the power exhaust.Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern,which in turn has a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region.The numerical simulation SOLPS5.0 package is used to design and explore the divertor target plates for HL-2M.We start with the choice of a proper target plate geometry,which has a smaller incidence angle in the permissible space,and then discuss the method of gas puffing to reduce the heat flux density on the target and the effects of gas puffing on the divertor plasma performance.展开更多
A comparative study of SN （single null）, DDN （disconnected double-null） and DN （double null） diverters for HCSB-DEMO （helium-cooled solid breeder, HCSB） is reported in this paper by using the 2D edge plasma tr...A comparative study of SN （single null）, DDN （disconnected double-null） and DN （double null） diverters for HCSB-DEMO （helium-cooled solid breeder, HCSB） is reported in this paper by using the 2D edge plasma transport code SOLPS5.0. There is a heat power of 400 MW crossing CIB （core interface boundary）. The peak heat flux at targets with different upstream densities is investigated. It is indicated that the peak heat flux at the outer target with a SN diverter is lower than that at the outer-down target with a DDN diverter under the same upstream density, but is higher than that at the outer target with a DN diverter. The diverter should operate at a proper upstremn density to avoid strong high heat flux at the targets. The peak heat flux at the targets and first wail are sensitive to the SOL （scrape-off-layer） grid thickness. The simulated results wilt provide data for the design of diverter in HCSB-DEMO.展开更多
One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in tu...One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Pack- age is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the di- vertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.展开更多
A low density plasma edge of small size divertor tokamak has been modeling by “B2SOLPS0.5.2 D” fluid transport code. The results of modeling are: 1) Formation of the strong “ITB” has detected more reliable with di...A low density plasma edge of small size divertor tokamak has been modeling by “B2SOLPS0.5.2 D” fluid transport code. The results of modeling are: 1) Formation of the strong “ITB” has detected more reliable with discovery that, low density plasma is necessary and important condition for it to form. 2) Reduction of plasma density play significantly role in the formation of the strong ITB as global parameter, possibly through change in the steep density gradient which stabilize “ITG” mode. 3) The radial electric field of small size divertor tokamak plasma edge is plasma density dependence and maximum radial electric field shear is found at low plasma density. 4) In the “NBI” discharge the toroidal (parallel) velocity at low plasma density is co-current and upward direction. 5) The structure of plasma pressure and radial electric field in quiescent H-mode are obtained.展开更多
Asymmetries between the divertor legs of small size divertor (SSD) tokamak plasma edge are noticed to reverse when the direction of toroidal magnetic field is reversed. In the present paper the small size divertor tok...Asymmetries between the divertor legs of small size divertor (SSD) tokamak plasma edge are noticed to reverse when the direction of toroidal magnetic field is reversed. In the present paper the small size divertor tokamak plasma edge under effect of toroidal magnetic field reversal is simulated by B2SOLPS0.5.2D fluid transport code. The simulation demonstrate the following results: 1) Parallel (toroidal) flow flux and Mach number up to 0.6 at higher plasma density reverse with reverse toroidal magnetic direction in the edge plasma of small size divertor tokamak. 2) The radial electric field is toroidal magnetic direction independence in edge plasma of small size divertor tokamak. 3) For normal and reverse toroidal magnetic field, the strong ITB is located between the positions of the maximum and minimum values of the radial electric field shear. 4) Simulation result shows that, the structure of radial electric field at high field side (HFS) and low field side (LFS) is different. This difference result from the change in the parallel flux flows in the scrape off layer (SOL) to plasma core through separatrix. 5) At a region of strong radial electric field shear, a large reduction of poloidal rotation was observed. 6) The poloidal rotation is toroidal magnetic field direction dependence.展开更多
The effect of poloidal E × B and diamagnetic drifts in edge plasma of Small Size Divertor (SSD) Tokamak is studied with two-dimensional B2SO- LPS-0.5.2D fluid transport code. The simulation results show the follo...The effect of poloidal E × B and diamagnetic drifts in edge plasma of Small Size Divertor (SSD) Tokamak is studied with two-dimensional B2SO- LPS-0.5.2D fluid transport code. The simulation results show the following: 1) For normal toroidal magnetic field, the increasing of core plasma density leads to large divertor asymmetries due to poloidal E × B and diamagnetic drifts. 2) Switching on the E × B and diamagnetic drifts leads to large change in poloidal distribution of radial electric field and induced counter-clockwise circulation (flow) around the x-point. 3) Switching on the E × B and diamagnetic drifts leads to the structure of poloidal distribution of radial electric field is nonmonotonic which responsible for negative spikes. 4) Switching on the E × B and diamagnetic drifts in vicinity of separatrix leads to the structure of poloidal distribution of radial electric field that has viscous layer. 5) Switching on the E × B and diamagnetic drifts results in torque generation. This torque spins up the toroidal rotation. 6) The E × B drift velocity depends on the plasma temperature heating and doesn't depend on plasma density.展开更多
An investigation into tungsten(W)impurity behaviors with the update of the EAST lower W divertor for H-mode has been carried out using SOLPS-ITER.This work aims to study the effect of external neon(Ne)impurity seeding...An investigation into tungsten(W)impurity behaviors with the update of the EAST lower W divertor for H-mode has been carried out using SOLPS-ITER.This work aims to study the effect of external neon(Ne)impurity seeding on W impurity sputtering with the bundled charge state model.As the Ne seeding rate increases,plasma parameters,W concentration(C_(W)),and eroded W flux(Γ_(W)^(Ero))at both targets are compared and analyzed between the highly resolved bundled model‘jett’and the full W charge state model.The results indicate that‘jett’can produce divertor behaviors essentially in agreement with the full W charge state model.The bundled scheme with high resolution in low W charge states(<W^(20+))has no obvious effect on the Ne impurity distribution and thus little effect on W sputtering by Ne.Meanwhile,parametric scans of radial particle and thermal transport diffusivities(D_(⊥)andχ_(e,i))in the SOL are simulated using the‘jett’bundled model.The results indicate that the transport diffusivity variations have significant influences on the divertor parameters,especially for W impurity sputtering.展开更多
To understand the effect of injected deuterium(D)pellets on background plasma,the ablation of D pellets and the transport of D species in both atomic and ionic states in the EAST device are simulated using a modified ...To understand the effect of injected deuterium(D)pellets on background plasma,the ablation of D pellets and the transport of D species in both atomic and ionic states in the EAST device are simulated using a modified dynamic neutral gas shield model combined with the edge plasma code SOLPS-ITER.The simulation results show that there is a phenomenon of obvious atomic deposition in the scrape-off layer(SOL)after pellet injection,which depends strongly on the injection velocity.With increasing injection velocity,the atomic density in the SOL decreases evidently and the deposition time is relatively shortened.Possible effects for triggering of edge localized modes(ELMs)by D and Li pellets are also discussed.With the same pellet size and injection velocity,the maximum perturbation pressure caused by D pellets is obviously higher.It is found that the resulting maximum perturbed pressure is remarkably enhanced when the injection velocity is reduced from 300 m/s to 100 m/s for a pellet with a cross section of 1.6 mm,which indicates that the injection velocity is important for ELM pacing.This work can provide reasonable guidance for choosing pellet parameters for fueling and ELM triggering.展开更多
Impurity seeding has been found effective for divertor detachment operations and the seeding location plays a key role in this process.In this work,we use the fluid code SOLPS-ITER to study the influence of seeding lo...Impurity seeding has been found effective for divertor detachment operations and the seeding location plays a key role in this process.In this work,we use the fluid code SOLPS-ITER to study the influence of seeding locations on divertor and scrape-off layer(D-SOL)plasmas in Experimental Advanced Superconducting Tokamak(EAST)with neon seeding.Simulation results indicate that the neon is a highly effective impurity in mitigating the heat flux and electron temperature peaks on the target of the divertor and achieving the partial detachment on both inner and outer targets.Further,by comparing results of the seeding at the private-flux region(PFR)plate(called‘TP’location)and the outer target(called‘XP’location),we find that the impurity density and power radiation for TP case are higher in core and upstream regions and lower in the divertor region than that for seeding at the XP,and the difference becomes more and more obvious as the seeding rate increases.It clearly demonstrates that the seeding at the XP location is more appropriate than at the TP location,especially in high seeding rate conditions.展开更多
基金supported by National Natural Science Foundation of China(nos. 11405213,11575236,11275231,11305206)the National Magnetic Confinement Fusion Science Program of China (nos.2013GB107003,2014GB106005,2015GB101000)
文摘The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer （SOL） width λq and heat spreading S,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip.Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of λq,L-mode =4.98 × Ip-0.68 and λq,H-mode =1.86 × Ip-1.08.Similar trends have also been demonstrated in the study of heat spreading with SL-mode =1.95 × Ip-0.542 and SH-mode =0.756 × Ip-0.872.In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor （CFETR）.
基金supported by Chinese National Natural Science Foundation(No.10135020)the JSPS-CAS Core-University Program on Plasma and Nuclear Fusion
文摘A detailed study of the divertor performance in EAST has been performed for both its double null （DN） and single null （SN） configurations. The results of application of the SOLPS （B2-Eirene） code package to the analysis of the EAST divertor are summarized. In this work, we concentrate on the effects of increased geometrical closure and of magnetic topology variation on the scrape-off layer （SOL） and divertor plasma behavior. The results of numerical predictions for the EAST divertor operational window are also described in this paper. A simple Core-SOL- Divertor （C-S-D） model was applied to investigate the possibility of extending plasma operational space of low hybrid current drive （LHCD） experiments for EAST.
基金Project supported by the National Key Research and Development Program of China(Grant No.2018YFE0303105)the Fundamental Research Funds for the Central Universities(Grant No.2022FRFK060021)the National MCF Energy Research and Development Program(Grant No.2019YFE03080300).
文摘The HIT-PSI is a linear plasma device built for physically simulating the high heat flux environment of future reactor divertors to test/develop advanced target plate materials.In this study,the geometry-modified SOLPS-ITER program is employed to examine the effects of the magnetic field strength and neutral pressure in the device on the heat flux experienced by the target plate of the HIT-PSI device.The findings of the numerical simulation indicate a positive correlation between the magnetic field strength and the heat flux density.Conversely,there is a negative correlation observed between the heat flux density and the neutral pressure.When the magnetic field strength at the axis exceeds 1 tesla and the neutral pressure falls below 10 Pa,the HIT-PSI has the capability to attain a heat flux of 10 MW·m-2 at the target plate.The simulation results offer a valuable point of reference for subsequent experiments at HIT-PSI.
基金funded by the National Magnetic Confinement Fusion Program of China(Nos.2019YFE03030000,2019YFE03080500 and 2022YFE03060004)National Natural Science Foundation of China(No.U19A20113)。
文摘A major challenge facing the steady-state operation of tokamak fusion reactors is to develop a viable divertor solution with order-of-magnitude increase in power handling capability as compared with present experience.A recently developed divertor concept for this end has been tested recently on EAST tokamak through combining the effects of a closed divertor corner and E×B drifts.The E×B drifts in the divertor move particles towards the outer divertor corner area in the scrape-off layer for B×▽B directed away from the divertor,which can significantly enhance the particle concentration there,facilitating divertor detachment.In recent EAST experiments,the effects have been demonstrated where the lowest electron temperature at the divertor plate is obtained with strike point located close to the corner in the horizontal target and with B×▽B away from the divertor.These experimental results are in reasonable agreement with SOLPS-ITER simulations including drift effects,suggesting that the new divertor concept potentially provides a promising divertor solution for long-pulse operations of future tokamak fusion reactors with much higher power fluxes.
基金supported by National Natural Science Foundation of China （No. 10975158）, the National Magnetic Confinenmnt Fusion Research Program of China （Nos. 2009GB106002, 2010GB104005） and in part by the JSPS-CAS Core University program in the field of ＇Plasma and Nuclear Fusion＇
文摘The edge transport code SOLPS5.0 is used to model edge plasmas in the experi- mental shots on JT-60U and the profiles of the transverse particle and heat transport coefficients D, Xe and Xi along the outer midplane are obtained by fitting the simulational results to the experimental data in L-mode shot 39090 and H-mode shots 37851, 37856. The modelling and fitting results show that within the pedestal region in H-mode shots 37851 and 37856 the radial particle transport coefficient D exhibits a significant drop, but, for L-mode shot 39090, the obvious drop in both D and Xe was not found.
文摘A simulation on HCSB-DEMO （helium-cooled solid breeder, HCSB） edge plasma, by using 2D edge plasma transport code SOLPS5.0, is presented. There is 400 MW heat power crossing CIB （core interface boundary）. The heat flux profiles and peak flux at the divertor targets with different boundary densities axe investigated. It is indicated that the HCSB-DEMO divertor should operate at a proper upstream density in order to avoid a high heat flux at the divertor targets. When the upstream density is 0.63x 102~ m-3, the peak heat flux at the divertor targets will be above 17 MW/m2. The cross-field transport, the power crossing CIB and the power fraction taken by electrons and ions and SOL （scrape-off-layer） thickness are analyzed as unknown parameters. It is shown that the peak flux at the divertor target is very sensitive to these parameters. The simulation results will be used in the divertor design for HCSB-DEMO.
基金supported by the National Magnetic Confinement Fusion Science Program of China(No.2009GB104008)National Natural Science Foundation of China(Nos.10975048,11175061)
文摘One of the critical issues to be solved for HL-2M is the power exhaust.Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern,which in turn has a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region.The numerical simulation SOLPS5.0 package is used to design and explore the divertor target plates for HL-2M.We start with the choice of a proper target plate geometry,which has a smaller incidence angle in the permissible space,and then discuss the method of gas puffing to reduce the heat flux density on the target and the effects of gas puffing on the divertor plasma performance.
文摘A comparative study of SN （single null）, DDN （disconnected double-null） and DN （double null） diverters for HCSB-DEMO （helium-cooled solid breeder, HCSB） is reported in this paper by using the 2D edge plasma transport code SOLPS5.0. There is a heat power of 400 MW crossing CIB （core interface boundary）. The peak heat flux at targets with different upstream densities is investigated. It is indicated that the peak heat flux at the outer target with a SN diverter is lower than that at the outer-down target with a DDN diverter under the same upstream density, but is higher than that at the outer target with a DN diverter. The diverter should operate at a proper upstremn density to avoid strong high heat flux at the targets. The peak heat flux at the targets and first wail are sensitive to the SOL （scrape-off-layer） grid thickness. The simulated results wilt provide data for the design of diverter in HCSB-DEMO.
基金supported by the National Magnetic Confinement Fusion Science Program of China(No.2009GB104008)National Natural Science Foundation of China(Nos.10975048,11175061)
文摘One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Pack- age is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the di- vertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.
文摘A low density plasma edge of small size divertor tokamak has been modeling by “B2SOLPS0.5.2 D” fluid transport code. The results of modeling are: 1) Formation of the strong “ITB” has detected more reliable with discovery that, low density plasma is necessary and important condition for it to form. 2) Reduction of plasma density play significantly role in the formation of the strong ITB as global parameter, possibly through change in the steep density gradient which stabilize “ITG” mode. 3) The radial electric field of small size divertor tokamak plasma edge is plasma density dependence and maximum radial electric field shear is found at low plasma density. 4) In the “NBI” discharge the toroidal (parallel) velocity at low plasma density is co-current and upward direction. 5) The structure of plasma pressure and radial electric field in quiescent H-mode are obtained.
文摘Asymmetries between the divertor legs of small size divertor (SSD) tokamak plasma edge are noticed to reverse when the direction of toroidal magnetic field is reversed. In the present paper the small size divertor tokamak plasma edge under effect of toroidal magnetic field reversal is simulated by B2SOLPS0.5.2D fluid transport code. The simulation demonstrate the following results: 1) Parallel (toroidal) flow flux and Mach number up to 0.6 at higher plasma density reverse with reverse toroidal magnetic direction in the edge plasma of small size divertor tokamak. 2) The radial electric field is toroidal magnetic direction independence in edge plasma of small size divertor tokamak. 3) For normal and reverse toroidal magnetic field, the strong ITB is located between the positions of the maximum and minimum values of the radial electric field shear. 4) Simulation result shows that, the structure of radial electric field at high field side (HFS) and low field side (LFS) is different. This difference result from the change in the parallel flux flows in the scrape off layer (SOL) to plasma core through separatrix. 5) At a region of strong radial electric field shear, a large reduction of poloidal rotation was observed. 6) The poloidal rotation is toroidal magnetic field direction dependence.
文摘The effect of poloidal E × B and diamagnetic drifts in edge plasma of Small Size Divertor (SSD) Tokamak is studied with two-dimensional B2SO- LPS-0.5.2D fluid transport code. The simulation results show the following: 1) For normal toroidal magnetic field, the increasing of core plasma density leads to large divertor asymmetries due to poloidal E × B and diamagnetic drifts. 2) Switching on the E × B and diamagnetic drifts leads to large change in poloidal distribution of radial electric field and induced counter-clockwise circulation (flow) around the x-point. 3) Switching on the E × B and diamagnetic drifts leads to the structure of poloidal distribution of radial electric field is nonmonotonic which responsible for negative spikes. 4) Switching on the E × B and diamagnetic drifts in vicinity of separatrix leads to the structure of poloidal distribution of radial electric field that has viscous layer. 5) Switching on the E × B and diamagnetic drifts results in torque generation. This torque spins up the toroidal rotation. 6) The E × B drift velocity depends on the plasma temperature heating and doesn't depend on plasma density.
基金supported by National Natural Science Foundation of China(Nos.12075283 and 11975271)。
文摘An investigation into tungsten(W)impurity behaviors with the update of the EAST lower W divertor for H-mode has been carried out using SOLPS-ITER.This work aims to study the effect of external neon(Ne)impurity seeding on W impurity sputtering with the bundled charge state model.As the Ne seeding rate increases,plasma parameters,W concentration(C_(W)),and eroded W flux(Γ_(W)^(Ero))at both targets are compared and analyzed between the highly resolved bundled model‘jett’and the full W charge state model.The results indicate that‘jett’can produce divertor behaviors essentially in agreement with the full W charge state model.The bundled scheme with high resolution in low W charge states(<W^(20+))has no obvious effect on the Ne impurity distribution and thus little effect on W sputtering by Ne.Meanwhile,parametric scans of radial particle and thermal transport diffusivities(D_(⊥)andχ_(e,i))in the SOL are simulated using the‘jett’bundled model.The results indicate that the transport diffusivity variations have significant influences on the divertor parameters,especially for W impurity sputtering.
基金the National Key R&D Program of China under Grant Nos.2017YFE0301100 and 2019YFE03030004the National Natural Science Foundation of China under Grant No.11575039Users with Excellence Program of Hefei Science Center CAS(2020HSCUE010)。
文摘To understand the effect of injected deuterium(D)pellets on background plasma,the ablation of D pellets and the transport of D species in both atomic and ionic states in the EAST device are simulated using a modified dynamic neutral gas shield model combined with the edge plasma code SOLPS-ITER.The simulation results show that there is a phenomenon of obvious atomic deposition in the scrape-off layer(SOL)after pellet injection,which depends strongly on the injection velocity.With increasing injection velocity,the atomic density in the SOL decreases evidently and the deposition time is relatively shortened.Possible effects for triggering of edge localized modes(ELMs)by D and Li pellets are also discussed.With the same pellet size and injection velocity,the maximum perturbation pressure caused by D pellets is obviously higher.It is found that the resulting maximum perturbed pressure is remarkably enhanced when the injection velocity is reduced from 300 m/s to 100 m/s for a pellet with a cross section of 1.6 mm,which indicates that the injection velocity is important for ELM pacing.This work can provide reasonable guidance for choosing pellet parameters for fueling and ELM triggering.
基金supported by the National Key Research and Development Program(No.2018YFE0303105)National MCF Energy R&D Program(No.2019YFE03080300)National Natural Science Foundation of China(No.11975087)。
文摘Impurity seeding has been found effective for divertor detachment operations and the seeding location plays a key role in this process.In this work,we use the fluid code SOLPS-ITER to study the influence of seeding locations on divertor and scrape-off layer(D-SOL)plasmas in Experimental Advanced Superconducting Tokamak(EAST)with neon seeding.Simulation results indicate that the neon is a highly effective impurity in mitigating the heat flux and electron temperature peaks on the target of the divertor and achieving the partial detachment on both inner and outer targets.Further,by comparing results of the seeding at the private-flux region(PFR)plate(called‘TP’location)and the outer target(called‘XP’location),we find that the impurity density and power radiation for TP case are higher in core and upstream regions and lower in the divertor region than that for seeding at the XP,and the difference becomes more and more obvious as the seeding rate increases.It clearly demonstrates that the seeding at the XP location is more appropriate than at the TP location,especially in high seeding rate conditions.
