Double Null Configuration Research Articles

Graphite Tile Thermal Performance on the New DIII-D Lower Divertor

The lower divertor of the DIII-D tokamak has been modified to provide improved density control of the tokamak plasma during operation in a high triangularity double-null configuration. Union Carbide ATJ grade graphite tiles covering the new lower divertor and vessel floor were designed to have better tile-to-tile alignment and to withstand higher heat flux than existing tiles.Gaps between tiles were successfully reduced from 2.5 to 0.4 mm and tile top surface alignment was greatly improved from 1.0 to 0.1 mm. Small tile gaps along with good vertical edge alignment greatly reduce the number and size of thin edges visible to the plasma, thus minimizing possible carbon introduction into the plasma. Close tile-to-tile alignment was the result of the very flat divertor plate surface, carefully controlled tile positioning, well-machined graphite tiles, and hand filing.Tiles were specified to survive 27 MJ of energy deposited per toroidal row of tiles during a 10 s shot period. When this energy is applied over the narrow triangular heat flux profiles originally specified, modeling shows that the tiles exceed maximum allowable tensile stress. Modeling does show that the tiles are able to absorb the 27 MJ per row without exceeding stress limits in cases where the heat flux profile is less focused than the original design specification.This paper will compare tile design analysis with operational experience obtained during the first 12-week operations campaign with the new divertor.

Design, fabrication and installation of the lower divertor for DIII-D

The geometry of the DIII-D tokamak lower divertor was recently modified to improve tokamak plasma density control during operation in a high triangularity double-null configuration. The primary component of the lower divertor is a toroidally continuous flat cooling plate that was fabricated by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Three rows of graphite tiles are mechanically attached to the plate to shield it from plasma impingement. The plate is water-cooled for heat removal between shots and is heated to 350 °C with hot air and inductive current during vessel baking. The divertor plate is supported 100 mm from the vacuum vessel floor to allow for cryo-pumping. The vacuum tight 90° plate sectors were positioned and welded together inside the vessel forming a toroidally continuous ring. Plasma facing tile designs have evolved from previous installations. To limit erosion caused by plasma impingement on sharp edges, the through tile-face bolt holes were eliminated from graphite in areas of high heat flux. Upgraded floor tiles were installed to improve the target for the plasma strike point for outer leg pumping. Thermal analysis was done for the Union Carbide ATJ grade graphite divertor shelf and vessel floor tiles and results are presented.

B2/Eirene modeling of EAST divertor target power loading with enhanced wall carbon source and additional neon injection

A major concern for EAST and future high-powered steady-state machines such as ITER is the power handling capability of the divertor target plates. Detailed modeling using the B2/Eirene code is carried out to assess the target power loading of the EAST divertor with both single-null and double-null configurations. A remarkable output is that an enhanced wall carbon source from chemical sputtering may lead to strong radiation in the plasma boundary, significantly reducing heat fluxes to the target plates, with little contamination to the core plasma at sufficiently high densities due to strong divertor screening. As a means of further reducing the power fluxes to the target plates and actively controlling the inner/outer divertor asymmetry in power loading, localized neon gas puffing from the outer divertor is investigated using B2/Eirene. The effect of this highly radiating impurity on target power loading and the consequence for core contamination is also presented.

Modelling of radial electric field profile for different divertor configurations**This was originally submitted to the special issue on H-mode Physics and Transport Barriers. This special issue may be accessed online at stacks.iop.org/PPCF/48/i=5A.

The impact of divertor configuration on the structure of the radial electric field has been simulated by the B2SOLPS5.0 transport fluid code. It is shown that the change in the parallel flows in the scrape-off layer, which are transported through the separatrix due to turbulent viscosity and diffusivity, should result in variation of the radial electric field and toroidal rotation in the separatrix vicinity. The modelling predictions are compared with the measurements of the radial electric field for the low field side equatorial mid-plane of ASDEX Upgrade in lower, upper and double-null (DN) divertor configurations. The parallel (toroidal) flows in the scrape-off layer and mechanisms for their formation are analysed for different geometries. It is demonstrated that a spike in the electric field exists at the high field side equatorial mid-plane in the connected DN divertor configuration. Its origin is connected with different potential drops between the separatrix vicinity and divertor plates in the two disconnected scrape-off layers, while the separatrix should be at almost the same potential. The spike might be important for additional turbulent suppression.

Poloidal Field Coil Configuration and Plasma Shaping Capability in NCT

This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter <tex xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">$ S(= q_95ast I_ p/ aast B_ t)$</tex> and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of <tex xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">$sim$</tex> 7 was obtained by double null divertor configuration with a lower aspect ratio of A <tex xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">$sim$</tex> 2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating.

Progress toward fully noninductive, high beta conditions in DIII-D

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario.

The cascading pebble divertor for the spherical tokamak power plant

The design of a power plant based on the spherical tokamak (ST) is being developed in order to explore its potential advantages. The plasma is operated in a double null configuration, forming both an upper and lower divertor. In order to accommodate the high erosion rates and heat fluxes developed in the divertors, a system based on a cascading flow of silicon carbide pebbles is being developed. The pebbles flow into the upper divertor where they fall as a toroidal curtain, which intercepts the divertor particle flux. The pebbles then flow under gravity through ducts to the lower divertor where they form a similar curtain. The bulk temperature of the pebbles rises to about 1150 °C although the outer surface is transiently heated to about 1800 °C. The pebbles pass out of the vacuum chamber into holding tanks and then into a fluidised bed heat exchanger. Here the pebbles are cooled down to about 340 °C and dust and damaged pebbles are removed. The pebbles are transferred to an upper tank by a pneumatic conveyor where the remaining gas is removed and the pebbles flow into the upper divertor again.

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.

H-mode physics of near double null plasmas in MAST and its applications to other tokamaks

High triangularity regimes, approaching double null, have become more important recently. H-mode access and edge stability are affected by the close proximity of the second X-point to the last closed flux surface, but more experimental evidence is needed. Inter-machine comparisons of MAST with ASDEX Upgrade (AUG) and NSTX confirm the reduction of the H-mode power threshold in the true double null configuration (C-DN), first reported on MAST. A more negative radial edge electric field in L-mode observed on MAST and ASDEX Upgrade may be connected to the improved H-mode access. About twice the power density is needed on MAST to achieve edge pressures similar to AUG in L- and H-modes. However, on MAST the pressure is dominated by the edge density in H-mode. So far, in single null (SN) ELMs have not been observed on MAST with neutral beam heating powers PNBI ⩽ 2 MW, due to the different edge profiles in SN H-mode. This is supported by the modelling of the stability of the measured edge profiles against ideal peeling–ballooning.

Small ELM regimes with good confinement on JET and comparison to those on ASDEX Upgrade, Alcator C-mod and JT-60U

Since it is uncertain if ITER operation is compatible with type-I edge localized modes (ELMs), the study of alternative ELM regimes is an urgent issue. This paper reports on experiments on JET aiming to find scenarios with small ELMs and good confinement, such as the type-II ELMs in ASDEX Upgrade, the enhanced D-alpha H-mode in Alcator C-mod or the grassy ELMs in JT-60U. The study includes shape variations, especially the closeness to a double-null configuration, variations of q95, density and beta poloidal. H-mode pedestals without type-I ELMs have been observed only at the lowest currents (≤1.2 MA), showing similarities to the observations in the devices mentioned above. These are discussed in detail on the basis of edge fluctuation analysis. For higher currents, only the mixed type-I/II scenario is observed. Although the increased inter-ELM transport reduces the type-I ELM frequency, a single type-I ELM is not significantly reduced in size. Obviously, these results do question the accessibility of such small ELM scenarios on ITER, except perhaps the high beta-poloidal scenario at higher q95, which could not be tested at higher currents at JET due to limitations in heating power.

Divertor Heat Flux Control Research on DIII-D

Divertor heat flux characterization and control results from DIII-D are summarized. The peak divertor heat flux is found to scale with a simple conduction model having perpendicular transport scaling with plasma current and heating power. In a double-null configuration, the heat flux sharing between divertors is very sensitive to the magnetic balance. Heat flux control in H-mode with edge-localized modes (ELMs) is obtained with deuterium gas puffing resulting in a partially detached divertor (PDD) regime. Important physical processes in the PDD regime include radiation from the intrinsic carbon impurity and deuterium, loss of electron pressure near the separatrix, parallel energy transport in the divertor dominated by convection, and particle flux reduction from deuterium recombination. Divertor neutral pressure is found to be an important control parameter to maintain the PDD regime. Divertor heat flux reduction is also obtained with impurity injection. In one approach divertor radiation is enhanced using induced scrape-off-layer flow to enrich divertor impurity concentration. Another approach uses seeded impurities to produce radiation inside the separatrix in a radiating mantle configuration. Observations of heat flux transients from ELMs and disruptions are summarized. Finally, the implications of these results for next-generation tokamaks are discussed.

Overview of MAST results

Significant progress has been made on the Mega Ampere Spherical Tokamak (MAST) towards a fundamental understanding of transport, stability and edge physics and addressing technological issues for future large devices. Collaborative studies of the L–H transition with NSTX and ASDEX Upgrade confirm that operation in a connected double-null configuration significantly reduces the threshold power, Pthr. The MAST data provide support for a theory for the transition based on finite β drift wave turbulence suppression by self-generated zonal flows. Analysis of low and high field side density gradients in the H-mode pedestal provides support for an analytical model of the density pedestal width dependent on the neutral penetration depth. Adding MAST data to international confinement databases has enhanced confidence in scalings for ITER by significantly expanding the range of β and ε explored and indicates a slightly stronger ε dependence than in current scalings. Studies of core transport have been conducted for well-diagnosed L-mode, H-mode and internal transport barrier (ITB) discharges using TRANSP, and microstability and turbulence studies have been carried out using GS2. Linear micro-stability analysis indicates that ITG modes are typically unstable on all flux surfaces with growth rates that are comparable to the equilibrium E × B flow shearing rate. Mixing length estimates of transport coefficients from ITG (neglecting flow shear) give diffusion coefficients that are broadly comparable with observed thermal diffusivities. Non-linear, collisionless ETG calculations have been performed and suggest radially extended electrostatic streamers up to 100ρe across in radius. Transport from ITG could easily be suppressed in regions where the E × B shear flow rate, ωSE, exceeds the ITG growth rate, possibly contributing to ITBs. Toroidal rotation, driven by neutral beam torque, is the dominant contribution to ωSE via the vϕBθ term in the radial electric field. Early edge localized mode activity on MAST is associated with the formation of narrow filamentary structures following field lines in the edge. These filaments rotate toroidally with the edge plasma and, away from the X-points, accelerate radially outwards from the edge up to 20 cm. Studies of disruptions on MAST demonstrate a complex evolution of core energy loss and resultant divertor power loads, including phases where the target heat flux width is broadened by a factor of 8. Observations of energetic particle modes driven by super-Alfvénic beam ions provide support for a model for the non-linear evolution of toroidal Alfvén eigenmodes (AEs) forming Bernstein–Green–Krushal waves. The AE activity reduces to low levels with increasing β. Plasma start-up without a central solenoid and in a manner compatible with future large spherical tokamak (ST) devices has been demonstrated using breakdown at a quadrupole magnetic null. Closed flux surface plasmas with peak plasma currents up to 370 kA have been generated and sustained for 0.3 s. New error field correction coils have extended the operational space for low density plasmas and enabled scaling studies of error field induced locked mode formation in the ST.

Simulation analysis of divertor performance in EAST

A detailed study of the divertor performance in the EAST has been conducted for both its double null and single null configurations. The results of the application of the SOLPS (B2/Eirene) code package to the analysis of the EAST divertor are summarized. Here we concentrate on the effects of the increased geometrical closure and variation in the magnetic topology on the behavior of divertor plasmas. The results of numerical predictions for the EAST divertor's operational window are also described in this paper.

Accounting of the power balance for neutral-beam heated H-mode plasmas in NSTX

Power balance accountability was investigated for neutral beam heated plasmas in NSTX. Measurements of heat to the divertor strike plates and radiation in the divertor and the core were made as a function of heating power in lower single-null (LSN) and double-null (DN) configurations. Up to 70% of the net input power is accounted in the LSN discharges with ≈15–20% of power lost as fast ions, ≈25–35% incident on the divertor plates, ≈5% radiated in the core, and ≈5–10% radiated in the divertor. In DN discharges, the fast ion loss is comparable to LSN discharges, the core radiated power fraction is ≈10%, and the power deposited in the lower divertor as radiation and heat flux is ≈20–25% of the total. Assuming up/down symmetry, the power accountability in DN is ≈85–90%.

Formation of transport barriers in the MAST spherical tokamak

In the Mega Ampère Spherical Tokamak (MAST) plasmas have been generated with internal (ITB) or edge (ETB) transport barriers. ITBs were achieved in both the electron and the ion energy channel. In the presence of an ITB in the ion energy channel, transport analysis shows that the ion thermal diffusivity, χi, is reduced to almost neoclassical values while the ITB persists. The widely tested criteria for ITB formation (ρs: Larmor radius at sound speed) obtained from dimensional analysis of JET discharges is easily exceeded on MAST. Even without the evidence of an ITB often applies, showing that this criterion in its current form is not generally applicable. ETBs are most easily formed in MAST if in a double null divertor configuration the discharge is vertically balanced, so that both X-points are almost on the same flux surface (CDND), and if the plasma is refuelled from the high field side mid-plane. The H-mode threshold power, Pthr = 0.5 MW, in connected double null diverted (CDND) is only about half of that in a similar disconnected discharge with the ion ∇ B drift towards the X-point on the last closed flux surface (LDND). Pthr scales between lower double null diverted (LDND) and the single null diverted configuration with the plasma surface area on MAST.

MHD stability analysis of type II ELMs in ASDEX Upgrade

We investigated the differences between type I and type II edge localized modes (ELMs) in ASDEX Upgrade using the MHD stability analysis. When plasma conditions are changed from typical type I ELMy conditions to type II ELMy conditions, the character of the edge instabilities change. With increased triangularity and edge safety factor, the low-n peeling–ballooning mode becomes more stable and more localized to the edge region. We find the same mode localization effect in almost double null configuration. When the plasma density is increased, the access to the second stability of the high-n ballooning modes is closed. The changes in the stability properties give a qualitative explanation to the observed small amplitude of the type II ELMs.

Changes in edge and scrape-off layer plasma behavior due to variation in magnetic balance in DIII-D

Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density n e,PED and plasma stored energy W T were sensitive to changes in magnetic balance near the double-null (DN) configuration, e.g., both decreased 20–30% when the DN shifted to a slightly unbalanced DN, where the B ×∇B drift direction pointed away from the main X-point. Recycling at each of the four divertor targets was sensitive to changes in magnetic balance and the B ×∇B drift direction. The poloidal distribution of the recycling in DN is in qualitative agreement with the predictions of UEDGE modeling with particle drifts included. The particle flux at the inner divertor target is shown to be much more sensitive to magnetic balance than the particle flux at the outer divertor target near the DN shape. These results suggest possible advantages and drawbacks for balanced DN operation.

Development of the spherical tokamak power plant

The design of a power plant based on the spherical tokamak (ST) is being developed in order to explore the potential advantages of the ST for power production. The ST power plant operates at a low aspect ratio of typically 1.4–1.6, which allows operation at high beta. This offers stability at high elongation which permits operation at high bootstrap current fraction (>95%), leaving only a modest external current drive requirement. This high elongation also results in a relatively uniform neutron wall loading giving the first wall components a longer life before replacement is required. The ST can be readily maintained on a timescale commensurate with this material life since its simple and robust centre column can be easily removed by lowering it into a hot cell beneath the load assembly and replaced with a new or reconditioned one. This is possible because it does not require inboard breeder blankets to maintain its tritium economy, or a substantial inboard shield since the centre column consists of a water-cooled copper centre rod surrounded by a relatively thin steel neutron shield. The plasma is operated in a double null configuration, forming both an upper and lower divertor. In order to accommodate the high erosion rates and heat fluxes developed in the divertors a novel system based on the use of a cascading flow of silicon carbide pebbles has been evaluated. The layout of the main elements of the power plant has been developed with revised power flows giving the steam cycle a thermal efficiency of about 43%.

Confinement and exhaust in the Mega Ampère Spherical Tokamak

The Mega Ampère Spherical Tokamak (MAST) is now accessing regimes with high normalized confinement relative to international scalings, HH(IPB98(y, 2))>∼1 at high normalized density, n̄e>60% of the Greenwald density. Data from MAST H-modes suggest that the aspect ratio dependency of international confinement and L–H threshold scalings may need to be modified to improve predictions for ITER. Access to H-mode on MAST is strongly affected by both the divertor magnetic geometry and fuelling location, with the formation of an edge transport barrier being facilitated by operation near the symmetric, connected double-null configuration and with poloidally localized inboard gas puffing. The ELMs on MAST appear to be Type III in nature, even in the highest performance plasmas and with the maximum available auxiliary heating power. ELM energy losses are less than 4% of stored energy in all regimes so far explored. These Type III ELMs are associated with a reduction in the pedestal density but no significant change in the pedestal temperature or temperature profile, indicating that energy is convected from the pedestal region into the scrape-off layer. Analysis of the energy observed to arrive at the divertor targets indicates that ELM losses are predominantly on the low field side. ELM effluxes are observed up to 20 cm from the plasma edge at the outboard mid-plane and are associated with the radial motion of a feature at an average velocity of 1.2 km s−1.

Overview on stationary and transient divertor heat loads

The divertor concept and the divertor materials envisaged for ITER FEAT restrict the maximum values of the stationary and transient heat fluxes. The maximum stationary heat flux is limited by the active cooling structure. The limit for transient events is given by the maximum tolerable surface temperature. This paper will review different options for divertor heat flux reduction. Proper geometric orientation of the divertor targets reduces the heat flux to the target plates at most by a factor of 60 relative to the parallel heat flux determined by upstream transport. The radiation capability of closed divertor configurations is doubled compared to open configurations. A further reduction of maximum heat load can be achieved by increasing the wetted area in double-null plasma configurations. The energy transported by transient events (type I ELMs) is up to 30% of the total energy deposited in the divertor. The heat flux profile during ELMs is not significantly broadened and is comparable to the profile between ELMs. The duration of the energy deposition varies with edge plasma parameters between 0.1 and 1 ms. The expected transient heat flux for ITER FEAT conditions is above the tolerable value. This requires the further development of operation scenarios with tolerable ELMs (e.g. type II) preserving the high energy confinement of type I ELM scenarios.