During the EAST radiative divertor experiments,one of the key challenges was how to avoid the occurrence of disruptive events caused by excessive impurity seeding.To estimate the required impurity fraction for diverto...During the EAST radiative divertor experiments,one of the key challenges was how to avoid the occurrence of disruptive events caused by excessive impurity seeding.To estimate the required impurity fraction for divertor detachment,we introduce a reduced edge plasma radiation model.In the model,based on the momentum conservation along the magnetic field line,the upstream pressure is determined by the plasma density and temperature at the divertor target,and then the impurity radiation loss is obtained by the balance of the heat and particle fluxes.It is found that the required impurity fraction shows a non-monotonic variation with divertor electron temperature(T_(d))when 0.1 eV<T_(d)<10 eV.In the range of 0.1 eV<T_(d)<1 e V,the position near the valley of required impurity fraction corresponds to strong plasma recombination.Due to the dependence of the volumetric momentum loss effect on the T_(d)in the range of 1 eV<T_(d)<10 eV,the required impurity fraction peaks and then decreases as T_(d)is increased.Compared to neon,the usage of argon reduces the impurity fraction by about twice.In addition,for the various fitting parameters in the pressure-momentum loss model,it is shown that the tendency of required impurity fraction with T_(d)always increases first and then decreases in the range of 1 eV<T_(d)<10 eV,but the required impurity fraction decreases when the model that characterizes the strong loss in pressure momentum is used.展开更多
The divertor design is critical to heat load handling and thus to achievements of highperformance plasma operations in the EHL-2(ENN He-Long 2)tokamak.This paper presents the design of an X-point target(XPT)divertor,f...The divertor design is critical to heat load handling and thus to achievements of highperformance plasma operations in the EHL-2(ENN He-Long 2)tokamak.This paper presents the design of an X-point target(XPT)divertor,featuring a conventional inner divertor and an XPT outer divertor,aimed at the effective control of heat loads,which may be extremely high during high ion temperature scenarios.The divertor target plates are made from carbon-based materials,which can handle heat loads of up to 5 MW/m².Divertor performances,including the heat load controllability,the onset of detachment and the in-out/up-down asymmetry,etc.,are evaluated using both the simple particle-tracking strategy and the complicated SOLPS-ITER code.Special attention is paid to the drift effects on particle/heat transport in the divertor/scrape-off layer region and on the divertor heat loads,focusing on the semi-detached/detached operation regimes.Results from SOLPS-ITER simulations demonstrated that the currently designed magnetic equilibrium and divertor configuration can effectively handle the power heat load in EHL-2.展开更多
The EHL-2(ENN He-Long 2)spherical torus(ST)project focuses on advancing spherical torus technology to address the unique challenges of p-^(11)B fusion,which demands significantly higher ion temperature and heat flux t...The EHL-2(ENN He-Long 2)spherical torus(ST)project focuses on advancing spherical torus technology to address the unique challenges of p-^(11)B fusion,which demands significantly higher ion temperature and heat flux to the divertor plate compared to traditional deuterium-tritium fusion.With a major radius of 1.05 m and a plasma current of 3 MA,the project aims to evaluate and optimize advanced divertor configurations,specifically the Super-X and X-point target(XPT)divertors.The design incorporates an up-down double-null configuration featuring a conventional inner divertor and an XPT outer divertor to effectively reduce the heat flux.The poloidal field(PF)coil system is meticulously optimized to balance engineering constraints with the flexibility in equilibrium configurations.This design is expected to provide a reference equilibrium configuration for other physics design issues and offer critical insight into heat load management.展开更多
Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic pr...Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.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
This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3 D geometry model.We assessed the nuclear r...This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3 D geometry model.We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability,which can support the development strategy of the physical and engineering design of the CFETR.Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R=7.2 m/a=2.2 m in the 22.5°model,and a fusion-power range of around 1-1.5 GW.The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design.The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out.We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years.An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit,which can support the design of the future update of the divertor region systems of the CFETR.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
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 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
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
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
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
The divertor target components for the Chinese fusion engineering test reactor(CFETR)and the future experimental advanced superconducting tokamak(EAST)need to remove a heat flux of up to20 MW m-2.In view of such a hig...The divertor target components for the Chinese fusion engineering test reactor(CFETR)and the future experimental advanced superconducting tokamak(EAST)need to remove a heat flux of up to20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure(TTS)are designed in the heat sink to improve the flat-tile divertor target’s heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height(H),width(W*),thickness(T),and spacing(L),on the HTP.The research results show that the flat-tile divertor target’s HTP is sensitive to the TTS parameter changes,and the sensitivity is T>L>W*>H.The HTP first increases and then decreases with the increase of T,L,and W*and gradually increases with the increase of H.The optimal design parameters are as follows:H=5.5 mm,W*=25.8 mm,T=2.2 mm,and L=9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux(HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2under the tungsten tile thickness<5 mm and the flow speed7 m s^(-1).The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by13%and30%compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of20 MW m-2of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only930℃.The flat-tile divertor target’s HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the flat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.展开更多
The measurement of impurity distribution in the divertor region of tokamaks is key to studying edge impurity transport.Therefore,a space-resolved vacuum-ultraviolet(VUV)spectrometer is designed to measure impurity emi...The measurement of impurity distribution in the divertor region of tokamaks is key to studying edge impurity transport.Therefore,a space-resolved vacuum-ultraviolet(VUV)spectrometer is designed to measure impurity emission in the divertor region on EAST.For good spectral resolution,an eagle-type VUV spectrometer with 1 m long focal length with spherical holograph grating is used in the system.For light collection,a collimating mirror is installed between the EAST plasma and the VUV spectrometer to extend the observing range to cover the upper divertor region.Two types of detectors,i.e.a back-illuminated charge-coupled device detector and a photomultiplier-tube detector,are adopted for the spectral measurement and high-frequency intensity measurement for feedback control,respectively.The angle between the entrance and exit optical axis is fixed at 15°.The detector can be moved along the exit axis to maintain a good focusing position when the wavelength is scanned by rotating the grating.The profile of impurity emissions is projected through the space-resolved slit,which is set horizontally.The spectrometer is equipped with two gratings with 2400 grooves/mm and2160 grooves/mm,respectively.The overall aberration of the system is reduced by accurate detector positioning.As a result,the total spectral broadening can be reduced to about 0.013 nm.The simulated performance of the system is found to satisfy the requirement of measurement of impurity emissions from the divertor area of the EAST tokamak.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
A simple Core-SOL-Divertor （C-S-D） model has been developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a ...A simple Core-SOL-Divertor （C-S-D） model has been developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operational space are also presented, From this study for the EAST operational space, it is evident that the C-S-D model is a useful tool for understanding qualitatively the overall features of the plasma operational space.展开更多
A three-dimensional analysis model based on the finite element method （FEM） is developed, which can derive the evolution and distribution characteristics of heat flux deposited on the divertor plate from the surface...A three-dimensional analysis model based on the finite element method （FEM） is developed, which can derive the evolution and distribution characteristics of heat flux deposited on the divertor plate from the surface temperature measured by infrared thermography diagnostics. The numerical simulations of surface heating due to localized power bursts and the power deposition calculations demonstrate that this analysis can provide accurate results and useful information about localized hot spots compared with the normal one- and two-dimensional calculations. In this paper, the details of this three- dimensional analysis are presented, and some results in ohmic heating and electron cyclotron resonant heating （ECRH） discharge on HL-2A are given.展开更多
Divertor plasma detachment offers one of the most promising operating modes for fusion devices because of low target power loading. In this article a ＇two-point＇ model is used to investigate the formation of detachm...Divertor plasma detachment offers one of the most promising operating modes for fusion devices because of low target power loading. In this article a ＇two-point＇ model is used to investigate the formation of detachment and explore the route to detachment in EAST, in order to find an ideal operation window. The simulation results show that impurity radiation and ionneutral friction are the main causes of divertor plasma detachment at the target plates. Raising the safety factor and reducing the upstream power density provide effective means to achieve the detachment due to the increased radiation power fraction. Puffing Ar and Ne impurities and raising the safety factor can bring the upstream high plasma temperature region （above 100 eV） and the low target plasma temperature region （below 10 eV） close to each other in terms of the separatrix density. But it is difficult to find a common operating region which satisfies both conditions. High recycling and detached regimes provides an ideal operation window because of the steady upstream condition and low target power load.展开更多
Atomic and molecular processes relevant to the volumetric recombinationphenomena were investigated in a linear divertor plasma simulator MAP-Ⅱ. Volumetric recombinationis induced in He plasma by puffing of He or H_2....Atomic and molecular processes relevant to the volumetric recombinationphenomena were investigated in a linear divertor plasma simulator MAP-Ⅱ. Volumetric recombinationis induced in He plasma by puffing of He or H_2. In the He puffing case, the reduction of the ionflux is dominated by the electron-ion recombination. In the H_2 puffing case, however, it isdominated by the molecule-assisted recombination (MAR), which is characterized by the disappearanceof the Helium Rydberg spectra and by the existence of the hydrogen negative ions. Currentachievement and the future prospect are described.展开更多
To extend the operation region of the Joint-Texas Experimental tokamak(J-TEXT) to the divertor configuration and even the H-mode,the divertor configuration discharge has been realized for the first time in the J-TEXT ...To extend the operation region of the Joint-Texas Experimental tokamak(J-TEXT) to the divertor configuration and even the H-mode,the divertor configuration discharge has been realized for the first time in the J-TEXT tokamak.Along with the establishment of a power supply for the divertor configuration,the construction of relevant diagnostics,and the installation of the divertor target on the high-field side,divertor discharge has been tested.Through the equilibrium calculation and position stability analysis,the control strategy has evolved to be more stable.High-density experiments and auxiliary heating experiments have been carried out on the divertor configuration.The special midplane single-null(MSN) divertor configuration is shown to be more stable than the limiter configuration in the density limit condition and can reach a higher density in the experiment.In the ECRH experiment,the power injection enhances the electron temperature and density,while more heat outflux is loaded on the divertor target tiles and causes more intensive recycling and impurity release.The future plan for the divertor configuration operation in the J-TEXT tokamak is also included.展开更多
基金supported by National Natural Science Foundation of China(No.12375227)Innovation in Fusion Engineering Technology of Institute(No.E35QT1080C)。
文摘During the EAST radiative divertor experiments,one of the key challenges was how to avoid the occurrence of disruptive events caused by excessive impurity seeding.To estimate the required impurity fraction for divertor detachment,we introduce a reduced edge plasma radiation model.In the model,based on the momentum conservation along the magnetic field line,the upstream pressure is determined by the plasma density and temperature at the divertor target,and then the impurity radiation loss is obtained by the balance of the heat and particle fluxes.It is found that the required impurity fraction shows a non-monotonic variation with divertor electron temperature(T_(d))when 0.1 eV<T_(d)<10 eV.In the range of 0.1 eV<T_(d)<1 e V,the position near the valley of required impurity fraction corresponds to strong plasma recombination.Due to the dependence of the volumetric momentum loss effect on the T_(d)in the range of 1 eV<T_(d)<10 eV,the required impurity fraction peaks and then decreases as T_(d)is increased.Compared to neon,the usage of argon reduces the impurity fraction by about twice.In addition,for the various fitting parameters in the pressure-momentum loss model,it is shown that the tendency of required impurity fraction with T_(d)always increases first and then decreases in the range of 1 eV<T_(d)<10 eV,but the required impurity fraction decreases when the model that characterizes the strong loss in pressure momentum is used.
基金sponsored by the Fundamental Research Funds for the Central Universities(No.232024G-10)and National Natural Sciences Foundation of China(Nos.12075052,12275098 and 12275307)the ENN Group and the ENN Energy Research Institute.
文摘The divertor design is critical to heat load handling and thus to achievements of highperformance plasma operations in the EHL-2(ENN He-Long 2)tokamak.This paper presents the design of an X-point target(XPT)divertor,featuring a conventional inner divertor and an XPT outer divertor,aimed at the effective control of heat loads,which may be extremely high during high ion temperature scenarios.The divertor target plates are made from carbon-based materials,which can handle heat loads of up to 5 MW/m².Divertor performances,including the heat load controllability,the onset of detachment and the in-out/up-down asymmetry,etc.,are evaluated using both the simple particle-tracking strategy and the complicated SOLPS-ITER code.Special attention is paid to the drift effects on particle/heat transport in the divertor/scrape-off layer region and on the divertor heat loads,focusing on the semi-detached/detached operation regimes.Results from SOLPS-ITER simulations demonstrated that the currently designed magnetic equilibrium and divertor configuration can effectively handle the power heat load in EHL-2.
基金supported by the ENN Group and the ENN Energy Research Institute.
文摘The EHL-2(ENN He-Long 2)spherical torus(ST)project focuses on advancing spherical torus technology to address the unique challenges of p-^(11)B fusion,which demands significantly higher ion temperature and heat flux to the divertor plate compared to traditional deuterium-tritium fusion.With a major radius of 1.05 m and a plasma current of 3 MA,the project aims to evaluate and optimize advanced divertor configurations,specifically the Super-X and X-point target(XPT)divertors.The design incorporates an up-down double-null configuration featuring a conventional inner divertor and an XPT outer divertor to effectively reduce the heat flux.The poloidal field(PF)coil system is meticulously optimized to balance engineering constraints with the flexibility in equilibrium configurations.This design is expected to provide a reference equilibrium configuration for other physics design issues and offer critical insight into heat load management.
基金This work was supported by National Natural Science Foundation of China No.19889502.
文摘Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.
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supported by the National Key Research and Development Program of China(Nos.2017YFE0300500,2017YFE0300503)。
文摘This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3 D geometry model.We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability,which can support the development strategy of the physical and engineering design of the CFETR.Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R=7.2 m/a=2.2 m in the 22.5°model,and a fusion-power range of around 1-1.5 GW.The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design.The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out.We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years.An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit,which can support the design of the future update of the divertor region systems of the CFETR.
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 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 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 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.
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.
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supported by the National MCF Energy R&D Program(No.2018YFE0312300)the National Key Research and Development Program of China(No.2017YFA0402500)the Science Foundation of the Institute of Plasma Physics,Chinese Academy of Sciences(No.Y45ETY2302)。
文摘The divertor target components for the Chinese fusion engineering test reactor(CFETR)and the future experimental advanced superconducting tokamak(EAST)need to remove a heat flux of up to20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure(TTS)are designed in the heat sink to improve the flat-tile divertor target’s heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height(H),width(W*),thickness(T),and spacing(L),on the HTP.The research results show that the flat-tile divertor target’s HTP is sensitive to the TTS parameter changes,and the sensitivity is T>L>W*>H.The HTP first increases and then decreases with the increase of T,L,and W*and gradually increases with the increase of H.The optimal design parameters are as follows:H=5.5 mm,W*=25.8 mm,T=2.2 mm,and L=9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux(HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2under the tungsten tile thickness<5 mm and the flow speed7 m s^(-1).The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by13%and30%compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of20 MW m-2of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only930℃.The flat-tile divertor target’s HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the flat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.
基金the National Magnetic Confinement Fusion Science Program of China(Nos.2017YFE0301300,2019YFE03030002 and 2018YFE0303103)National Natural Science Foundation of China(No.12175278)+7 种基金Anhui Province Key Research and Development Program(No.202104a06020021)ASIPP Science and Research Grant(No.DSJJ-2020-02)Anhui Provincial Natural Science Foundation(No.1908085J01)Distinguished Young Scholar of Anhui Provincial Natural Science Foundation(No.2008085QA39)Instrument Developing Project of the Chinese Academy of Sciences(No.YJKYYQ20180013)the Comprehensive Research Facility for Fusion Technology Program of China(No.2018-000052-73-01-001228)the University Synergy Innovation Program of Anhui Province(No.GXXT-2021-029)CAS President’s International Fellowship Initiative(No.2022VMB0007)。
文摘The measurement of impurity distribution in the divertor region of tokamaks is key to studying edge impurity transport.Therefore,a space-resolved vacuum-ultraviolet(VUV)spectrometer is designed to measure impurity emission in the divertor region on EAST.For good spectral resolution,an eagle-type VUV spectrometer with 1 m long focal length with spherical holograph grating is used in the system.For light collection,a collimating mirror is installed between the EAST plasma and the VUV spectrometer to extend the observing range to cover the upper divertor region.Two types of detectors,i.e.a back-illuminated charge-coupled device detector and a photomultiplier-tube detector,are adopted for the spectral measurement and high-frequency intensity measurement for feedback control,respectively.The angle between the entrance and exit optical axis is fixed at 15°.The detector can be moved along the exit axis to maintain a good focusing position when the wavelength is scanned by rotating the grating.The profile of impurity emissions is projected through the space-resolved slit,which is set horizontally.The spectrometer is equipped with two gratings with 2400 grooves/mm and2160 grooves/mm,respectively.The overall aberration of the system is reduced by accurate detector positioning.As a result,the total spectral broadening can be reduced to about 0.013 nm.The simulated performance of the system is found to satisfy the requirement of measurement of impurity emissions from the divertor area of the EAST tokamak.
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 in part by the JSPS-CAS Core-University Program in the field of Plasma and Nuclear Fusionalso carried out as a joint project under the Facility Utilization Program of JAERI.
文摘A simple Core-SOL-Divertor （C-S-D） model has been developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operational space are also presented, From this study for the EAST operational space, it is evident that the C-S-D model is a useful tool for understanding qualitatively the overall features of the plasma operational space.
基金Project supported by the National Natural Science Foundation of China (Grant No. 10805016)the National Magnetic Confinement Fusion Science Program,China (Grant No. 2009GB104008).
文摘A three-dimensional analysis model based on the finite element method （FEM） is developed, which can derive the evolution and distribution characteristics of heat flux deposited on the divertor plate from the surface temperature measured by infrared thermography diagnostics. The numerical simulations of surface heating due to localized power bursts and the power deposition calculations demonstrate that this analysis can provide accurate results and useful information about localized hot spots compared with the normal one- and two-dimensional calculations. In this paper, the details of this three- dimensional analysis are presented, and some results in ohmic heating and electron cyclotron resonant heating （ECRH） discharge on HL-2A are given.
基金supported by National Natural Science Foundation of China(No.10675129)
文摘Divertor plasma detachment offers one of the most promising operating modes for fusion devices because of low target power loading. In this article a ＇two-point＇ model is used to investigate the formation of detachment and explore the route to detachment in EAST, in order to find an ideal operation window. The simulation results show that impurity radiation and ionneutral friction are the main causes of divertor plasma detachment at the target plates. Raising the safety factor and reducing the upstream power density provide effective means to achieve the detachment due to the increased radiation power fraction. Puffing Ar and Ne impurities and raising the safety factor can bring the upstream high plasma temperature region （above 100 eV） and the low target plasma temperature region （below 10 eV） close to each other in terms of the separatrix density. But it is difficult to find a common operating region which satisfies both conditions. High recycling and detached regimes provides an ideal operation window because of the steady upstream condition and low target power load.
基金The project partly supported by TEPCO Research Foundation
文摘Atomic and molecular processes relevant to the volumetric recombinationphenomena were investigated in a linear divertor plasma simulator MAP-Ⅱ. Volumetric recombinationis induced in He plasma by puffing of He or H_2. In the He puffing case, the reduction of the ionflux is dominated by the electron-ion recombination. In the H_2 puffing case, however, it isdominated by the molecule-assisted recombination (MAR), which is characterized by the disappearanceof the Helium Rydberg spectra and by the existence of the hydrogen negative ions. Currentachievement and the future prospect are described.
基金supported by the National MCF Energy R&D Program of China(Nos.2018YFE0301104 and 2018YFE0310300)National Natural Science Foundation of China(No.51821005)
文摘To extend the operation region of the Joint-Texas Experimental tokamak(J-TEXT) to the divertor configuration and even the H-mode,the divertor configuration discharge has been realized for the first time in the J-TEXT tokamak.Along with the establishment of a power supply for the divertor configuration,the construction of relevant diagnostics,and the installation of the divertor target on the high-field side,divertor discharge has been tested.Through the equilibrium calculation and position stability analysis,the control strategy has evolved to be more stable.High-density experiments and auxiliary heating experiments have been carried out on the divertor configuration.The special midplane single-null(MSN) divertor configuration is shown to be more stable than the limiter configuration in the density limit condition and can reach a higher density in the experiment.In the ECRH experiment,the power injection enhances the electron temperature and density,while more heat outflux is loaded on the divertor target tiles and causes more intensive recycling and impurity release.The future plan for the divertor configuration operation in the J-TEXT tokamak is also included.
