A liquid Li divertor is a promising alternative for future fusion devices.In this work a new divertor model is proposed,which is processed by 3D-printing technology to accurately control the size of the internal capil...A liquid Li divertor is a promising alternative for future fusion devices.In this work a new divertor model is proposed,which is processed by 3D-printing technology to accurately control the size of the internal capillary structure.At a steady-state heat load of 10 MW m^(-2),the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation.In order to evaluate the wicking ability of the capillary structure,the wicking process at 600℃ was simulated by FLUENT.The result was identical to that of the corresponding experiments.Within 1 s,liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface.During the wicking process,the average wicking mass rate of lithium should reach 0.062 g s^(-1),which could even supplement the evaporation requirement of liquid lithium under an environment>950℃.Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device(SCU-PSI),and the evolution of the vapor cloud during plasma irradiation was analyzed.It was found that the target temperature tends to plateau despite the gradually increased input current,indicating that the vapor shielding effect is gradually enhanced.The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure.These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid-solid combined target designs.展开更多
Achieving the detachment of divertor can help to alleviate excessive heat load and sputtering problems on the target plates,thereby extending the lifetime of divertor components for fusion devices.In order to provide ...Achieving the detachment of divertor can help to alleviate excessive heat load and sputtering problems on the target plates,thereby extending the lifetime of divertor components for fusion devices.In order to provide a fast but relatively reliable prediction of plasma parameters along the flux tube for future device design,a one-dimensional(1D)modeling code for the operating point of impurity seeded detached divertor is developed based on Python language,which is a fluid model based on previous work(Plasma Phys.Control.Fusion 58045013(2016)).The experimental observation of the onset of divertor detachment by neon(Ne)and argon(Ar)seeding in EAST is well reproduced by using the 1D modeling code.The comparison between the 1D modeling and two-dimensional(2D)simulation by the SOLPS-ITER code for CFETR detachment operation with Ne and Ar seeding also shows that they are in good agreement.We also predict the radiative power loss and corresponding impurity concentration requirement for achieving divertor detachment via different impurity seeding under high heating power conditions in EAST and CFETR phase II by using the 1D model.Based on the predictions,the optimized parameter space for divertor detachment operation on EAST and CFETR is also determined.Such a simple but reliable 1D model can provide a reasonable parameter input for a detailed and accurate analysis by 2D or three-dimensional(3D)modeling tools through rapid parameter scanning.展开更多
On J-TEXT,the temporal evolution of heat flux distribution on the high-field side(HFS)divertor plate has been measured by an infrared(IR)camera during the plasma operation with an island divertor configuration.In expe...On J-TEXT,the temporal evolution of heat flux distribution on the high-field side(HFS)divertor plate has been measured by an infrared(IR)camera during the plasma operation with an island divertor configuration.In experiments,the island divertor configuration is an edge magnetic island chain structure surrounded by stochastic layers,which can be induced by resonant magnetic perturbations(RMPs).The experimental results show that the heat flux distribution on the HFS target plate depends significantly on the edge magnetic topology.Furthermore,the impact of hydrogen fueling using supersonic molecular beam injection(SMBI)on the divertor heat flux distributions is studied on J-TEXT with an island divertor configuration.It has been observed that power detachment can be achieved when the radiation front approaches the last closed flux surface(LCFS)after each SMBI pulse.This result may provide a method of access for divertor detachment on a fusion device with a three-dimensional(3D)boundary magnetic structure.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
Temperature response functions have been developed to investigate sensor design and divertor heat flux estimation in magnetically confined plasmas. The time-dependent heat flux can be derived by fitting the response f...Temperature response functions have been developed to investigate sensor design and divertor heat flux estimation in magnetically confined plasmas. The time-dependent heat flux can be derived by fitting the response function to experimental thermocouple(TC) data. Because the TC signals have a time delay to transit events such as discharge start or confinement transition, the time delay is taken into account in a temperature response function. Such a function accurately describes the signal from each TC channel with time delay in a sensor test using a neutral beam injection. Measurement for commercial TCs shows that the time delay is caused by the finite heat capacity of TC wire and contact heat resistance between TC and target surface.展开更多
Accurate measurement of the average plasma parameters in the edge region,including the temperature and density of electrons and ions,is critical for understanding the characteristics of the scrape-off layer(SOL) and d...Accurate measurement of the average plasma parameters in the edge region,including the temperature and density of electrons and ions,is critical for understanding the characteristics of the scrape-off layer(SOL) and divertor plasma transport in magnetically confined fusion research.On the J-TEXT tokamak,a multi-channel retarding field analyzer(RFA) probe has been developed to study average plasma parameters in the edge region under various poloidal divertor and island divertor configurations.The edge radial profile of the ion-to-electron temperature ratio,τ_(i/e),has been determined,which gradually decreases as the SOL ion self-collisionality,v_(SOL)*,increases.This is broadly consistent with what has been observed previously from various tokamak experiments.However,the comparison of experimental results under different configurations shows that in the poloidal divertor configuration,even under the same v_(SOL)*,τ_(i/e) in the SOL region becomes smaller as the distance from the X-point to the target plate increases.In the island divertor configuration,τ_(i/e) near the O-point is higher than that near the X-point at the same v_(SOL)*,and both are higher than those in the limiter configuration.These results suggest that the magnetic configuration plays a critical role in the energy distributions between electrons and ions at the plasma boundary.展开更多
In the Large Helical Device （LHD）, two different divertor configurations, i.e. helical divertor （HD） and local island divertor （LID）, are utilized to control the edge plasma. The HD with two X-points is an intri...In the Large Helical Device （LHD）, two different divertor configurations, i.e. helical divertor （HD） and local island divertor （LID）, are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.展开更多
During the discharging of Tokamak devices, interactions between the core plasma and plasma-facing components (PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFC...During the discharging of Tokamak devices, interactions between the core plasma and plasma-facing components (PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFCs materials. An infrared (IR) diagnostic system consisting of an IR camera and an endoscope was installed on an Experimental Advanced Superconducting Tokamak (EAST) to monitor the surface temperature of the lower divertor target plate (LDTP) and to calculate the corresponding heat flux based on its surface temperature and physical structure, via the finite element method. First, the temperature obtained by the IR camera was calibrated against the temperature measured by the built-in thermocouple of EAST under baking conditions to determine the true temperature of the LDTP. Next, based on the finite element method, a target plate model was built and a discretization of the modeling domain was carried out. Then, a heat conduction equation and boundary conditions were determined. Finally, the heat flux was calculated. The new numerical tool provided results similar to those for DFLUX;this is important for future work on related physical processes and heat flux control.展开更多
In HL-2A tokamaks, the behavior of heat flux deposited on the divertor targets has been studied during deuterium gas fuelling. The heat flux is reduced significantly after supersonic molecular beam injection （SMBI） ...In HL-2A tokamaks, the behavior of heat flux deposited on the divertor targets has been studied during deuterium gas fuelling. The heat flux is reduced significantly after supersonic molecular beam injection （SMBI） fuelling during Ohmic and electron cyclotron resonance heating （ECRH） divertor discharges. The SMBI fuelling causes an increase in the plasma density and this change results in the experienced change of the edge properties. Most of this reduction in divertor target heat flux occurs together with a high plasma radiation region located at near the X-point. The largest reduction in heat flux profiles is observed at the outboard divertor separatrix strike point, while the heat flux far from the strike point remains almost unchanged. In particular, with SMBI multi-pulses gas fuelling, a partially detached divertor regime is observed with a highly radiating region at the X-point. With the onset of the partially detached divertor regime, a sudden drop in both heat flux and power flow on the divertor target is observed. The reduction in power load on the divertor targets is roughly equal to the increase in plasma radiation loss.展开更多
HL-2A tokamak is the first tokamak with divertors in China. The plasma boundary and the position of the striking point on the target plates of the HL-2A closed diwrtor were simulated by the current filament code and t...HL-2A tokamak is the first tokamak with divertors in China. The plasma boundary and the position of the striking point on the target plates of the HL-2A closed diwrtor were simulated by the current filament code and they were in agreement with the diagnostic results in the divertor. Supersonic molecular beam injection （SMBI） system was first installed and tested on the HL-2A tokamak in 2004. In the present experiment low pressure SMBI fuelling on the HL-2A closed divertor was carried out. The experimental results indicate that the divertor was operated in the ＇linear regime＇ and during the period of SMB pulse injection into the HL-2A plasma the power density eonvected at the target plate surfaces was 0.4 times of that before or after the beam injection. It is a useful fuelling method for decreasing the heat load on the neutralizer plates of the divertor.展开更多
Single- and double-null divertor configurations in HL-2A are simulated by SWEQU equilibrium code. Lower divertor discharges in the first physics campaign have been achieved by two kinds of power supply method of multi...Single- and double-null divertor configurations in HL-2A are simulated by SWEQU equilibrium code. Lower divertor discharges in the first physics campaign have been achieved by two kinds of power supply method of multipole-field coils. Single-null divertor configuration has been identified by visible photography, target probe arrays and the reconstructed magnetic surface. Magnetic separatrix and minor radius of plasma column are obtained by a reconstructed code of multiple current filaments using 18 Mirnov signals.展开更多
An upgraded infrared （IR） imaging system which provides a wide field of view （FOV） has been installed on the Experimental Advanced Superconducting Tokamak （EAST） to monitor the surface temperatures on plasma fac...An upgraded infrared （IR） imaging system which provides a wide field of view （FOV） has been installed on the Experimental Advanced Superconducting Tokamak （EAST） to monitor the surface temperatures on plasma facing components. Modified magnetic topology induced by lower hybrid wave （LHW） can lead to the formation of striated heat flux （SHF} on divertor plates which can be clearly observed by IR camera. In this paper, LHW power modulation is applied to analyze the appearance of SHF. It is also demonstrated that deuterium （D） pellet injection and supersonic molecular beam injection （SMBI） can to some extent reduce the heat flux on the outer strike point （OSP）, but enhance the SHF on lower outer plates （LOP） of divertor. This may provide an optional approach to actively control the distribution of heat flux on diveror plates, which can protect materials from long duration high-heat flux.展开更多
基金funded by the China Postdoctoral Science Foundation(No.2019M663487)the National Key Research and Development Program of China(No.2022YFE03130000)。
文摘A liquid Li divertor is a promising alternative for future fusion devices.In this work a new divertor model is proposed,which is processed by 3D-printing technology to accurately control the size of the internal capillary structure.At a steady-state heat load of 10 MW m^(-2),the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation.In order to evaluate the wicking ability of the capillary structure,the wicking process at 600℃ was simulated by FLUENT.The result was identical to that of the corresponding experiments.Within 1 s,liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface.During the wicking process,the average wicking mass rate of lithium should reach 0.062 g s^(-1),which could even supplement the evaporation requirement of liquid lithium under an environment>950℃.Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device(SCU-PSI),and the evolution of the vapor cloud during plasma irradiation was analyzed.It was found that the target temperature tends to plateau despite the gradually increased input current,indicating that the vapor shielding effect is gradually enhanced.The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure.These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid-solid combined target designs.
基金Project supported by the National Key Research and Development Program of China (Grant No.2022YFE03030001)the National Natural Science Foundation of China (Grant No.12075283)。
文摘Achieving the detachment of divertor can help to alleviate excessive heat load and sputtering problems on the target plates,thereby extending the lifetime of divertor components for fusion devices.In order to provide a fast but relatively reliable prediction of plasma parameters along the flux tube for future device design,a one-dimensional(1D)modeling code for the operating point of impurity seeded detached divertor is developed based on Python language,which is a fluid model based on previous work(Plasma Phys.Control.Fusion 58045013(2016)).The experimental observation of the onset of divertor detachment by neon(Ne)and argon(Ar)seeding in EAST is well reproduced by using the 1D modeling code.The comparison between the 1D modeling and two-dimensional(2D)simulation by the SOLPS-ITER code for CFETR detachment operation with Ne and Ar seeding also shows that they are in good agreement.We also predict the radiative power loss and corresponding impurity concentration requirement for achieving divertor detachment via different impurity seeding under high heating power conditions in EAST and CFETR phase II by using the 1D model.Based on the predictions,the optimized parameter space for divertor detachment operation on EAST and CFETR is also determined.Such a simple but reliable 1D model can provide a reasonable parameter input for a detailed and accurate analysis by 2D or three-dimensional(3D)modeling tools through rapid parameter scanning.
基金supported by the National Magnetic Confinement Fusion Energy R&D Program of China(No.2018YFE0309103)National Natural Science Foundation of China(Nos.12305243 and 51821005)。
文摘On J-TEXT,the temporal evolution of heat flux distribution on the high-field side(HFS)divertor plate has been measured by an infrared(IR)camera during the plasma operation with an island divertor configuration.In experiments,the island divertor configuration is an edge magnetic island chain structure surrounded by stochastic layers,which can be induced by resonant magnetic perturbations(RMPs).The experimental results show that the heat flux distribution on the HFS target plate depends significantly on the edge magnetic topology.Furthermore,the impact of hydrogen fueling using supersonic molecular beam injection(SMBI)on the divertor heat flux distributions is studied on J-TEXT with an island divertor configuration.It has been observed that power detachment can be achieved when the radiation front approaches the last closed flux surface(LCFS)after each SMBI pulse.This result may provide a method of access for divertor detachment on a fusion device with a three-dimensional(3D)boundary magnetic structure.
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partially performed with the support and under the auspices of the NIFS Collaborative Research Program(Nos.NIFS20KLPR051,NIFS20KUHL099 and NIFS20KUGM153)。
文摘Temperature response functions have been developed to investigate sensor design and divertor heat flux estimation in magnetically confined plasmas. The time-dependent heat flux can be derived by fitting the response function to experimental thermocouple(TC) data. Because the TC signals have a time delay to transit events such as discharge start or confinement transition, the time delay is taken into account in a temperature response function. Such a function accurately describes the signal from each TC channel with time delay in a sensor test using a neutral beam injection. Measurement for commercial TCs shows that the time delay is caused by the finite heat capacity of TC wire and contact heat resistance between TC and target surface.
基金supported by the National Magnetic Confinement Fusion Energy R&D Program of China (No.2018YFE0309100)National Natural Science Foundation of China (No.51821005)。
文摘Accurate measurement of the average plasma parameters in the edge region,including the temperature and density of electrons and ions,is critical for understanding the characteristics of the scrape-off layer(SOL) and divertor plasma transport in magnetically confined fusion research.On the J-TEXT tokamak,a multi-channel retarding field analyzer(RFA) probe has been developed to study average plasma parameters in the edge region under various poloidal divertor and island divertor configurations.The edge radial profile of the ion-to-electron temperature ratio,τ_(i/e),has been determined,which gradually decreases as the SOL ion self-collisionality,v_(SOL)*,increases.This is broadly consistent with what has been observed previously from various tokamak experiments.However,the comparison of experimental results under different configurations shows that in the poloidal divertor configuration,even under the same v_(SOL)*,τ_(i/e) in the SOL region becomes smaller as the distance from the X-point to the target plate increases.In the island divertor configuration,τ_(i/e) near the O-point is higher than that near the X-point at the same v_(SOL)*,and both are higher than those in the limiter configuration.These results suggest that the magnetic configuration plays a critical role in the energy distributions between electrons and ions at the plasma boundary.
基金supported by NIFS under Grant(No.NIFS05ULPP506)in part by the JSPS-CAS Core-University Program in the field of Plasma and Nuclear Fusion
文摘In the Large Helical Device （LHD）, two different divertor configurations, i.e. helical divertor （HD） and local island divertor （LID）, are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.
基金supported by the National Natural Science Foundation of China(Nos.51505120 and 11105028)the National Magnetic Confinement Fusion Science Program of China(No.2015GB102004)
文摘During the discharging of Tokamak devices, interactions between the core plasma and plasma-facing components (PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFCs materials. An infrared (IR) diagnostic system consisting of an IR camera and an endoscope was installed on an Experimental Advanced Superconducting Tokamak (EAST) to monitor the surface temperature of the lower divertor target plate (LDTP) and to calculate the corresponding heat flux based on its surface temperature and physical structure, via the finite element method. First, the temperature obtained by the IR camera was calibrated against the temperature measured by the built-in thermocouple of EAST under baking conditions to determine the true temperature of the LDTP. Next, based on the finite element method, a target plate model was built and a discretization of the modeling domain was carried out. Then, a heat conduction equation and boundary conditions were determined. Finally, the heat flux was calculated. The new numerical tool provided results similar to those for DFLUX;this is important for future work on related physical processes and heat flux control.
基金supported by National Natural Science Foundation of China(No.10805016)the National Magnetic Confinement Fusion Science Program of China(No.2009GB104008)
文摘In HL-2A tokamaks, the behavior of heat flux deposited on the divertor targets has been studied during deuterium gas fuelling. The heat flux is reduced significantly after supersonic molecular beam injection （SMBI） fuelling during Ohmic and electron cyclotron resonance heating （ECRH） divertor discharges. The SMBI fuelling causes an increase in the plasma density and this change results in the experienced change of the edge properties. Most of this reduction in divertor target heat flux occurs together with a high plasma radiation region located at near the X-point. The largest reduction in heat flux profiles is observed at the outboard divertor separatrix strike point, while the heat flux far from the strike point remains almost unchanged. In particular, with SMBI multi-pulses gas fuelling, a partially detached divertor regime is observed with a highly radiating region at the X-point. With the onset of the partially detached divertor regime, a sudden drop in both heat flux and power flow on the divertor target is observed. The reduction in power load on the divertor targets is roughly equal to the increase in plasma radiation loss.
基金Project supported by the National Science Foundation of China （Grant Nos 19775011, 10075016 and 10475024）.The authors wish to thank the HL-2A team members for their hard work.
文摘HL-2A tokamak is the first tokamak with divertors in China. The plasma boundary and the position of the striking point on the target plates of the HL-2A closed diwrtor were simulated by the current filament code and they were in agreement with the diagnostic results in the divertor. Supersonic molecular beam injection （SMBI） system was first installed and tested on the HL-2A tokamak in 2004. In the present experiment low pressure SMBI fuelling on the HL-2A closed divertor was carried out. The experimental results indicate that the divertor was operated in the ＇linear regime＇ and during the period of SMB pulse injection into the HL-2A plasma the power density eonvected at the target plate surfaces was 0.4 times of that before or after the beam injection. It is a useful fuelling method for decreasing the heat load on the neutralizer plates of the divertor.
文摘Single- and double-null divertor configurations in HL-2A are simulated by SWEQU equilibrium code. Lower divertor discharges in the first physics campaign have been achieved by two kinds of power supply method of multipole-field coils. Single-null divertor configuration has been identified by visible photography, target probe arrays and the reconstructed magnetic surface. Magnetic separatrix and minor radius of plasma column are obtained by a reconstructed code of multiple current filaments using 18 Mirnov signals.
基金supported by National Natural Science Foundation of China(No.11275234)the National Magnetic Confinement Fusion Programof China(No.2014GB106001)
文摘An upgraded infrared （IR） imaging system which provides a wide field of view （FOV） has been installed on the Experimental Advanced Superconducting Tokamak （EAST） to monitor the surface temperatures on plasma facing components. Modified magnetic topology induced by lower hybrid wave （LHW） can lead to the formation of striated heat flux （SHF} on divertor plates which can be clearly observed by IR camera. In this paper, LHW power modulation is applied to analyze the appearance of SHF. It is also demonstrated that deuterium （D） pellet injection and supersonic molecular beam injection （SMBI） can to some extent reduce the heat flux on the outer strike point （OSP）, but enhance the SHF on lower outer plates （LOP） of divertor. This may provide an optional approach to actively control the distribution of heat flux on diveror plates, which can protect materials from long duration high-heat flux.
