Detached Divertor Operation Research Articles

Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

UEDGE modelling of detached divertor operation for long‐leg divertor geometries in ARC

Simulations of the ARC scrape‐off layer and divertor configuration are carried out using UEDGE to perform the first power‐handling assessment for the divertor design. A convective–diffusive anomalous transport model is implemented in the code to reproduce the radial plasma profiles anticipated for the device at the outer mid‐plane. Levels of exhaust power from the core are varied, and a 0.5% fixed fraction neon impurity seeding is included in investigations to find steady‐state detached solutions. Initial studies employing a super‐X divertor configuration have shown that a stable detached divertor‐operational window exists for reactor exhaust powers in the range of 72–92 MW, which is close to projected levels of exhaust power of 105 MW across the separatrix anticipated for ARC. Under these conditions, the divertor heat flux is fully dissipated by radiation and radial losses to the sidewalls of the divertor channel, leading to acceptable levels of power flux density on plasma‐facing components. Simulations are being extended to study the performance of the X‐point target geometry and to explore the sensitivity to the physics model and input parameters.

Plasma detachment in divertor tokamaks

Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

Assessment of X-point target divertor configuration for power handling and detachment front control

A study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE (Rognlien et al., J. Nucl. Mater. 196, 347, 1992). The model parameters are based on the ADX tokamak concept design (LaBombard et al., Nucl. Fusion 55, 053020, 2015). Several long-legged divertor configurations are considered, in particular the X-point target configuration proposed for ADX, and compared with a standard divertor. For otherwise identical conditions, a scan of the input power from the core plasma is performed. It is found that as the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remaining stable. At low power the detachment front eventually moves to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, the operation window for a detached divertor in the standard divertor configuration is very small, or even non-existent; under the same conditions for long-legged divertors the detached operation window is quite large, in particular for the X-point target configuration, allowing a factor of 5–10 variation in the input power. These modeling results point to possibility of stable fully detached divertor operation for a tokamak with extended divertor legs.

Attainment of a stable, fully detached plasma state in innovative divertor configurations

A computational study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE. Several divertor configurations are considered, with radially or vertically extended, tightly baffled, outer divertor legs and with or without a secondary X-point in the divertor leg volume. For otherwise identical conditions, a scan of the input power from the core plasma is performed. As the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state, which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remains stable. At low power, the detachment front eventually moves all the way to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, for long-legged divertors, the detached operation window is quite large, in particular, for the X-point target configuration using a secondary X-point in the divertor leg volume, allowing a factor of 5–10 variations in the input power. For the same parameters, for the standard divertor configuration, the detached operation window is very small or even non-existent. The present modeling results suggest the possibility of stable fully detached divertor operation for a tokamak with tightly baffled extended divertor legs.

Power exhaust by SOL and pedestal radiation at ASDEX Upgrade and JET

Future fusion reactors require a safe, steady state divertor operation. A possible solution for the power exhaust challenge is the detached divertor operation in scenarios with high radiated power fractions. The radiation can be increased by seeding impurities, such as N for dominant scrape-off-layer radiation, Ne or Ar for SOL and pedestal radiation and Kr for dominant core radiation.Recent experiments on two of the all-metal tokamaks, ASDEX Upgrade (AUG) and JET, demonstrate operation with high radiated power fractions and a fully-detached divertor by N, Ne or Kr seeding with a conventional divertor in a vertical target geometry. For both devices similar observations can be made. In the scenarios with the highest radiated power fraction, the dominant radiation originates from the confined region, in the case of N and Ne seeding concentrated in a region close to the X-point.Applying these seed impurities for highly radiative scenarios impacts local plasma parameters and alters the impurity transport in the pedestal region. Thus, plasma confinement and stability can be affected. A proper understanding of the effects by these impurities is required in order to predict the applicability of such scenarios for future devices.

Conceptual design of divertor and first wall for DEMO-FNS

Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m−2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280–300 °С at 5 MW m−2 and 600–650 °С at 10.5 MW m−2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

Compatibility of detached divertor operation with robust edge pedestal performance

The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, Te⩽2eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

First operation with the JET International Thermonuclear Experimental Reactor-like wall

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 1021 es−1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN≈3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.

Scaling radiative divertor solutions to high power in DIII-D

Detached radiative divertor plasmas are examined in DIII-D as a function of power. The 2D profile of plasma electron density and temperature is reconstructed from divertor Thomson data while energy transport is deduced from power balance measurements. The combination of plasma and energy diagnostics reveals that parallel energy transport transitions to convective flow at a constant 5–10 eV as a function of power. Simple considerations would imply divertor density proportional to power density for Mach 1 flow. However, the divertor, upstream separatrix and core plasma densities increase significantly less than linear with power density. During detached divertor operation H-mode core confinement is maintained though the edge pedestal can experience moderate degradation.

Negative ion production near a divertor plate

In detached divertor operation, typical plasma parameters in front of target plates are densities from 1019m−3 up to 1020m−3 and temperatures below 3eV. In this regime, production of negative hydrogen ions can occur, mainly through vibrational excited molecule or by negative surface ionization of hydrogen ions. To study this effect, we have developed a 1D-3V Particle-in-Cell electrostatic model of the electro-negative sheath. The motion of charged particles (e, H+, H2+ and H−) in their self-consistent electric field and of excited neutral-particle states (H(n=1,2,3) and H2(X1Σg+,v=0,…,14)) is simulated. Surface and volumetric processes have been included by using different Monte Carlo Collision methods. The importance of negative ions for the sheath potential and hydrogen fluxes on to the divertor will be calculated for different parameters.

Implications of the use of Carbon-Based Plasma Facing Components in Next Step Fusion Devices

The implications for the use of carbon based PFCs in next step devices are reviewed, with ITER as the reference example device. Issues with respect to steady-state operation concerning the carbon source at the divertor target in realistic conditions (including impurity seeding and mixed material formation caused by interaction of the plasma with the main chamber PFCs) and their implications for tritium retention are discussed taking into account other restrictions for next step plasma operation, such as the achievement of detached divertor operation with peak power loads <10 MW/m2. The effect of transient power loads on the divertor CFC target caused by ELMs are discussed and the implications for the use of CFCs at the divertor target for regimes with repetitive ELMs, acceptable from the divertor lifetime point of view, are reviewed.

Plasma pressure and flows during divertor detachment

MHD theory applied to tokamak plasma scrape-off layer (SOL) equilibria requires Pfirsch–Schlüter current, which, because the magnetic lines are open, normally closes through electrically conducting divertor or limiter components. During detached divertor operation the Pfirsch–Schlüter current path to the divertor target is sometimes blocked, in which case theory predicts that the plasma develops a poloidal pressure gradient around the upstream SOL and a corresponding parallel flow, in order to satisfy all the conditions of MHD equilibrium. This paper reports the only known examples of detached diverted plasma in the DIII–D tokamak with blocked Pfirsch–Schlüter current, and they show no clear SOL poloidal pressure differences. However, the predicted pressure differences are small, near the limit of detectability with the available diagnostics. In the more usual DIII–D `partially detached divertor' operation mode, the Pfirsch–Schlüter current appears to never be blocked, and no unusual poloidal pressure differences are observed, as expected. Finally, a local overpressure is observed just inside the magnetic separatrix near the X-point in both attached and detached Ohmically heated plasmas.

Evolution of 2D deuterium and impurity radiation profiles during transitions from attached to detached divertor operation in DIII-D

This paper presents the detailed evolution of conditions along both the inner and outer divertor legs during the transition from attached ELMing H-mode to partially detached divertor (PDD) operation in DIII-D. Visible emission profiles in a poloidal plane show that in ELMing H-mode prior to deuterium gas injection, CIII emission peaks in the inner SOL near the X-point and deuterium emission (from ionization and recombination) peaks at the inner target plate near the inner strike point (ISP). The spatial profiles of the recombination and ionization zones, determined by forming images of the ratio of intensities from simultaneous images of D α and D γ emission, show that recombination dominates the inner leg emission near the target; ionization dominates in a poloidally narrow zone upstream in the inner leg. After deuterium injection, when the PDD transition begins, the profiles of carbon visible emission show first an increase in the inner SOL near the X-point, followed by increases in emission in the lower regions of the outer leg. Deuterium emission at the transition onset decreases at the ISP and increases across the private flux region below the X-point. As the transition to PDD conditions proceeds the deuterium emission increases in the private flux region; recombination dominates near the floor and ionization higher near the X-point. Carbon emission appears along both divertor legs and at the X-point. In the final quasi-steady PDD state, the recombination emission in the outer leg is near the separatrix and along the target plate; emission from collisional excitation dominates in the upper part of the outer leg just below the X-point, and carbon emission is localized at the X-point. These results suggest that transport of neutral deuterium between the inner and outer divertor legs through the private flux region plays an important role in the initiation of outer leg detachment in DIII-D.

Helium exhaust in ELMy H-mode plasmas with W-shaped pumped divertor of JT-60U

By injecting a neutral beam of 60 keV helium (He) atoms as central fueling of helium into the ELMy H-mode plasmas, helium exhaust has been studied in the W-shaped pumped divertor on JT-60U. Efficient He exhaust was realized by He pumping using argon frosted cryopumps in the JT-60U new divertor. The He fueling rate by He beam injection was balanced with He pumping. In steady state, good He exhaust capability was successfully demonstrated in attached ELMy H-mode plasmas. A global particle confinement time of τ* He=0.7 s and τ* He/ τ E=4 ( τ E: the energy confinement time) was achieved in attached plasmas, well within the range generally considered necessary for successful operation of a future fusion reactor, such as ITER. The enrichment factor of He was obtained to be about 1.0, which is five times larger than the ITER requirement of 0.2. Good He exhaust capability was also obtained in detached ELMy H-mode plasmas, which was comparable to one in attached plasmas. This result of the helium exhaust is sufficient to support a detached divertor operation on ITER.

Stability of the detachment front in a tokamak divertor

Experimental observations show that when a divertor plasma detaches, radiation regions (fronts) move towards the X-point. This indicates that the localization of the fronts in the divertor is unstable. We study theoretically the stability of detached divertor operation and compare our results with experimental data. We show that the impurity radiation and plasma recombination effects play a crucial role in the evolution of detached plasma parameters. In the tokamak SOL these effects result in the saturation of upstream plasma density and temperature as the total number of particles (ions and neutrals) in the SOL is increased, and cause both impurity radiation and ionization–recombination fronts propagation towards the X-point. These conclusions are supported by experimental data from Alcator C-Mod.

Overview of EU CFCs development for plasma facing materials

To improve the properties of carbon materials, the tritium inventory should be reduced, chemical erosion and RES have to be suppressed to increase the resistance to water/oxygen at elevated temperatures. In addition, in the next generation devices, i.e., the International Thermonuclear Experimental Reactor (ITER), plasma disruption, slow transients, and ELMs, which can occur as off-normal events as the result of a transition from detached divertor operation to attached operation causes extremely high heat loading to carbon protection material. Therefore, Carbon fiber composites (CFCs) with high thermal conductivity (300 W m −1 K −1 at 20°C, 145 W m −1 K −1 at 800°C) are favourable. In framework of European Fusion Technology program, a great effort has been made to develop CFCs to meet all requirements. This paper presents an overview in progress of EU CFCs development. The characteristics of CFCs with respect to thermal-mechanical properties, erosion by plasma, tritium retention, H 2O/O 2 reactions, and neutron irradiation effects were reported.

Neutral particle dynamics in the Alcator C-Mod tokamak

An experimental study of neutral particle dynamics has been carried out over a wide range of operating conditions in the Alcator C Modified (Alcator C-Mod) tokamak using special purpose pressure gauges. Both divertor and midplane pressures are found to depend primarily on the edge plasma regimes defined by the scrape-off layer heat transport. While the maximum divertor neutral pressure (30-60 mtorr) is achieved at high core plasma densities (ne=(2-4)*1020 m-3), corresponding to the detached divertor state, the maximum ratio of divertor to midplane pressure (~70) is achieved in the high recycling transport regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas, the divertor pressure and compression are maximized when the strike point is located at the bottom of the vertical target plate. Modelling shows that the high pressure sustained during detached divertor operation (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat

Modelling of transition from attached to detached state and self-consistent calculations of He levels in ITER

This paper describes models that address both power and particle control with the ITER divertor. For power control, the characteristics of the transition from attached to detached divertor operation are studied with an extension of our previous simple model [Y. Igitkhanov et al., 22nd EPS, Bournemouth (1995) p. IV-317]. This new model has been applied to JET experiments to show that the model can reproduce global features of the transition. The model has been applied to ITER to identify the range for detached operation in terms of the heating power, upstream density, and impurity level, and the relation between divertor control and core plasma conditions. For particle control, we use the PRETOR transport code together with a two point divertor model [K. Borrass and G. Janeschitz, Nucl. Fusion 34 (1994) 1203] for the edge boundary conditions to study the sensitivity of the central helium concentration to the pumping speed and helium enrichment in the divertor.

Comparison of detached and radiative divertor operation in Alcator C-Mod

The divertor of the Alcator C-Mod tokamak [Phys. Plasmas 1, 1511 (1994)] routinely radiates a large fraction of the power entering the scrape-off layer. This dissipative divertor operation occurs whether the divertor is detached or not, and large volumetric radiative emissivities, up to 60 MW m−3 in ion cyclotron range of frequency (ICRF) heated discharges, have been measured using bolometer arrays. An analysis of both Ohmic and ICRF-heated discharges has demonstrated some of the relative merits of detached divertor operation versus high-recycling divertor operation. An advantage of detached divertor operation is that the power flux to the divertor plates is decreased even further than its already low value. Some disadvantages are that volumetric losses outside the separatrix in the divertor region are decreased, the neutral compression ratio is decreased, and the penetration efficiency of impurities increases.