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.

Thermal Fatigue Study on the Divertor Plate Materials

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.

Predictions for EAST Divertor Performance

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.

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.

The first results of divertor discharge and supersonic molecular beam injection on the HL-2A tokamak

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.

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.

Study of Striated Heat Flux on EAST Divertor Plates Induced by LHW Using Infrared Camera

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.

Numerical Simulation and Experimental Identification of Divertor Configuration in the HL-2A Tokamak

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.

Overall Features of EAST Operation Space by Using Simple Core-SOL-Divertor Model

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.

Reconstruction of heat flux profile on the HL-2A divertor plate with a three-dimensional analysis model

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.

Investigation of Divertor Detachment in EAST by‘Two-Point’Model

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.

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.

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.

Optimization of the Water-Cooled Structure for the Divertor Plates in EAST Based on an Orthogonal Theory

An orthogoual experimental scheme was designed for optimizing a water-cooled structure of the divertor plate. There were three influencing factors： the radius R of the water- cooled pipe, and the pipe spacing L1 and L3. The influence rule of different factors on the cooling effect and thermal stress of the plate were studied, for which the influence rank was respectively R 〉 L1 〉 L3 and L3 〉 R 〉 L1. The highest temperature value decreased when R and L1 increased~ and the maximum thermal stress value dropped when R, L1 and L3 increased. The final optimized results can be summarized as： R equals 6 mm or 7 mm, L1 equals 19 mm, and L3 equals 20 mm. Compared with the initial design, the highest temperature value had a small decline~ and the maximum thermal stress value dropped by 19~ to 24~. So it was not ideal to improve the cooling effect by optimizing the geometry sizes of the water-cooled structure, even worse than increasing the flow speed, but it was very effective for dropping the maximum thermal stress value. The orthogoaal experimental method reduces the number of experiments by 80%, and thus it is feasible and effective to optimize the water-cooled structure of the divertor plate with the orthogonal theory.

Review of Divertor Studies in LHD

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.

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.

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.

Effects of some parameters on the divertor plasma sheath characteristics and fuel retention in castellated tungsten tile gaps

Castellation of plasma facing components is foreseen as the best solution for ensuring the lifetime of future fusion devices. However, the gaps between the resulting surface elements can increase fuel retention and complicate fuel removal issues. To know how the fuel is retained inside the gaps, the plasma sheath around the gaps needs to be understood first. In this work, a kinetic model is used to study plasma characteristics around the divertor gaps with the focus on the H＋ penetration depth inside the poloidal gaps, and a rate-theory model is coupled to simulate the hydrogen retention inside the tungsten gaps. By varying the magnetic field strength and plasma temperature, we find that the H＋ cyclotron radius has a significant effect on the penetration depth. Besides, the increase of magnetic field inclination angle can also increase the penetration depth. It is found in this work that parameters as well as the penetration depth strongly affect fuel retention in tungsten gaps.