Closed corner divertor with B×▽B away from the divertor:a promising divertor scenario for tokamak power exhaust

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.

Improvement of a temperature response function for divertor heat flux monitoring in fusion devices

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.

A study on nuclear analysis of the divertor region of the CFETR

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.

Calculation of the heat flux in the lower divertor target plate using an infrared camera diagnostic system on the experimental advanced superconducting tokamak

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.

Space-resolved vacuum-ultraviolet spectroscopy for measuring impurity emission from divertor region of EAST tokamak

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.

OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations, i.e. the inner divertor, the outer divertor and the dome, in the EAST superconducting tokamak for typical ohmic plasma conditions. It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations. However, it quickly approaches a similar steady state value for Ar recycling efficiency >0.9. OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

As an important component of tokamaks,the divertor is mainly responsible for extracting heat and helium ash,and the targets of the divertor need to withstand high heat flux of 10 MW m-2 for steady-state operation.In this study,we proposed a new strategy,using microchannel cooling technology to remove high heat load on the targets of the divertor.The results demonstrated that the microchannel-based W/Cu flat-type mock-up successfully withstood the thermal fatigue test of 1000 cycles at 10 MW m^(-2)with cooling water of 26 l min^(-1),30°C(inlet),0.8 MPa(inlet),15 s power on and 15 s dwell time;the maximum temperature on the heat-loaded surface(W surface)of the mock-up was 493°C,which is much lower than the recrystallization temperature of W(1200°C).Moreover,no occurrence of macrocrack and‘hot spot’at the W surface,as well as no detachment of W/Cu tiles were observed during the thermal fatigue testing.These results indicate that microchannel cooling technology is an efflcient method for removing the heat load of the divertor at a low flow rate.The present study offers a promising solution to replace the monoblock design for the EAST divertor.

Realization of divertor configuration discharge in J-TEXT tokamak

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.

Divertor detachment operation in helium plasmas with ITER-like tungsten divertor in EAST

Detachment in helium(He)discharges has been achieved in the EAST superconducting tokamak equipped with an ITER-like tungsten divertor.This paper presents the experimental observations of divertor detachment achieved by increasing the plasma density in He discharges.During density ramp-up,the particle flux shows a clear rollover,while the electron temperature around the outer strike point is decreasing simultaneously.The divertor detachment also exhibits a significant difference from that observed in comparable deuterium(D)discharges.The density threshold of detachment in the He plasma is higher than that in the D plasma for the same heating power,and increases with the heating power.Moreover,detachment assisted with neon(Ne)seeding was also performed in L-and H-mode plasmas,pointing to the direction for reducing the density threshold of detachment in He operation.However,excessive Ne seeding causes confinement degradation during the divertor detachment phase.The precise feedback control of impurity seeding will be performed in EAST to improve the compatibility of core plasma performance with divertor detachment for future high heating power operations.

A divertor plasma configuration design method for tokamaks

The efficient and safe operation of large fusion devices strongly relies on the plasma configuration inside the vacuum chamber.It is important to construct the proper plasma equilibrium with a desired plasma configuration.In order to construct the target configuration,a shape constraint module has been developed in the tokamak simulation code(TSC),which controls the poloidal flux and the magnetic field at several defined control points.It is used to construct the double null,lower single null,and quasi-snowflake configurations for the required target shape and calculate the required PF coils current.The flexibility and practicability of this method have been verified by the simulated results.

Simulation of Electric Fields in Small Size Divertor Tokamak Plasma Edge

The fluid simulation of Small Size Divertor Tokamak (SSDT) plasma edge by the B2-SOLPS5.0 2D [1] transport code gives the following results: First, in the vicinity of separatrix the radial electric field result is not close to the neoclassical electric field. Second, the shear of radial electric field is independent on plasma parameters. Third, switching on poloidal drifts (E×B and diamagnetic drifts) leads to asymmetric parallel and poloidal fluxes from outer to inner plates and upper part of SOL for normal direction of toroidal magnetic field. Fourth, for the normal direction of toroidal magnetic, the radial electric field of SSDT is affected by the variation in temperature heating of plasma. Fifth, the parallel flux is directed from inner to outer plate in case of discharge without neutral beam injection (NBI).

Study on divertor plasma behavior through sweeping strike point in new lower divertor on EAST

A series of L-mode discharges have been conducted in the new‘corner slot’divertor on the Experimental Advanced Superconducting Tokamak(EAST)to study the divertor plasma behavior through sweeping strike point.The plasma control system controls the strike point sweeping from the horizontal target to the vertical target through poloidal field coils,with keeping the main plasma stability.The surface temperature of the divertor target cools down as the strike point moves away,indicating that sweeping strike point mitigates the heat load.To avoid the negative effect of probe tip damage,a method based on sweeping strike point is used to get the normalized profile and study the decay length of particle and heat flux on the divertor target λ_(js),_λ(q).In the discharges with high radio-frequency(RF)heating power,electron temperature T_(e) is lower and λ_(js)is larger when the strike point locates on the horizontal target compared to the vertical target,probably due to the corner effect.In the Ohmic discharges,λ_(js),λ_(q) are much larger compared to the discharges with high RF heating power,which may be attributed to lower edge T_(e).

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

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.

Upgrade of an integrated Langmuir probe system on the closed divertor target plates in the HL-2A tokamak

A newly designed divertor Langmuir probe diagnostic system has been installed in a rare closed divertor of the HL-2A tokamak and steadily operated for the study of divertor physics involved edge-localized mode mitigation,detachment and redistribution of heat flux,etc.Two sets of probe arrays including 274 probe tips were placed at two ports(approximately 180°separated toroidally),and the spatial and temporal resolutions of this measurement system could reach6 mm and 1μs,respectively.A novel design of the ceramic isolation ring can ensure reliable electrical insulation property between the graphite tip and the copper substrate plate where plasma impurities and the dust are deposited into the gaps for a long experimental time.Meanwhile,the condition monitoring and mode conversion between single and triple probe of the probe system could be conveniently implemented via a remote-control station.The preliminary experimental result shows that the divertor Langmuir probe system is capable of measuring the high spatiotemporal parameters involved the plasma density,electron temperature,particle flux as well as heat flux during the ELMy H-mode discharges.

Towards advanced divertor configurations on the J-TEXT tokamak

Developing advanced magnetic divertor configurations to address the coupling of heat and particle exhaust with impurity control is one of the major challenges currently constraining the further development of fusion research.It has therefore become the focus of extensive attention in recent years.In J-TEXT,several new divertor configurations,including the high-field-side single-null poloidal divertor and the island divertor,as well as their associated fundamental edge divertor plasma physics,have recently been investigated.The purpose of this paper is to briefly summarize the latest progress and achievements in this relevant research field on J-TEXT from the past few years.

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73×10^(19)m^(-3) to 6.49×10^(19)m^(-3),while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(>50%)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3 D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.

Study on divertor heat flux under n=3 and n=4 resonant magnetic perturbations using infrared thermography diagnostic in EAST

Resonant magnetic perturbations(RMPs)with high toroidal mode number n are considered for controlling edge-localized modes(ELMs)and divertor heat flux in future ITER H-mode operations.In this paper,characteristics of divertor heat flux under high-nRMPs(n=3 and 4)in H-mode plasma are investigated using newly upgraded infrared thermography diagnostic in EAST.Additional splitting strike point(SSP)accompanying with ELM suppression is observed under both RMPs with n=3 and n=4,the SSP in heat flux profile agrees qualitatively with the modeled magnetic footprint.Although RMPs suppress ELMs,they increase the stationary heat flux during ELM suppression.The dependence of heat flux on q_(95)during ELM suppression is preliminarily investigated,and further splitting in the original strike point is observed at q 495=during ELM suppression.In terms of ELM pulses,the presence of RMPs shows little influence on transient heat flux distribution.

Analysis of Divertor Heat Flux with Infrared Thermography During Gas Fuelling in the HL-2A Tokamak

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.

Numerical Simulation on Subcooled Boiling Heat Transfer Characteristics of Water-Cooled W/Cu Divertors

In order to realize safe and stable operation of a water-cooled W/Cu divertor under high heating condition,the exact knowledge of its subcooled boiling heat transfer characteristics under different design parameters is crucial.In this paper,subcooled boiling heat transfer in a water-cooled W/Cu divertor was numerically investigated based on computational fluid dynamic(CFD).The boiling heat transfer was simulated based on the Euler homogeneous phase model,and local differences of liquid physical properties were considered under one-sided high heating conditions.The calculated wall temperature was in good agreement with experimental results,with the maximum error of 5%only.On this basis,the void fraction distribution,flow field and heat transfer coefficient(HTC)distribution were obtained.The effects of heat flux,inlet velocity and inlet temperature on temperature distribution and pressure drop of a water-cooled W/Cu divertor were also investigated.These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.