Divertor Physics Research Articles

Expanding the role of impurity spectroscopy for investigating the physics of high-Z dissipative divertors

New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, q∥ > 1 GW/m2, while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs) are expected to use extrinsic impurity seeding. Due to transport and atomic physics processes which impact impurity ionization balance, so-called ‘non-coronal’ effects, we do not accurately know and have yet to demonstrate the maximum q∥ which can be mitigated in a tokamak. Radiation enhancement for nitrogen is shown to arise primarily from changes in Li- and Be-like charge states on open field lines, but also through transport-driven enhancement of H- and He-like charge states in the pedestal region. Measurements are presented from nitrogen seeded H-mode and L-mode plasmas where emission from N1+ through N6+ are observed. Active charge exchange spectroscopy of partially ionized low-Z impurities in the plasma edge is explored to measure N5+ and N6+ within the confined plasma, while passive UV and visible spectroscopy is used to measure N1+-N4+ in the boundary. Examples from recent JET and Alcator C-Mod experiments which employ nitrogen seeding highlight how improving spectroscopic coverage can be used to gain empirical insight and provide more data to validate boundary simulations.

Investigations on the heat flux and impurity for the HL-2M divertor

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with = 0.5 MA at the first stage of HL-2M and the high parameters with = 2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, , to achieve a better performance in terms of the parameters of discharges.

A theoretical interpretation of the main scrape-off layer heat-flux width scaling for tokamak inner-wall limited plasmas

The International Tokamak Physics Activity Topical Group on scrape-off layer and divertor physics has amassed a database comprising hundreds of reciprocating Langmuir probe measurements of the main scrape-off layer heat-flux width in inner-wall limited discharges. We have carried out an analysis, based on turbulent transport theory, of the variation of with respect to the dimensionless plasma parameters. Restricting our analysis to circular plasmas, we find that a model based on non-linearly saturated turbulence can well reproduce the values found in the database.

Advanced divertor concept design and analysis for HL-2M

HL-2M is a tokamak device that is under construction and will be put into operation in the near future. Based on the magnetic coil design of HL-2M, standard divertor, snowflake divertor and tripod divertor configurations have been designed. The potential properties of snowflake divertor configurations have been analyzed, such as the low poloidal field (Bp) area around the X-point, the connection length, target plate and magnetic field shear. The linear peeling-ballooning (P-B) mode is studied by BOUT++ code for snowflake divertor configurations. According to the divertor configuration properties of HL-2M, asymmetric target plates have been concept designed to be compatible with the intended single null (SN) divertor configurations as well as double null (DN) divertor configurations. The SOLPS5.0 code is used to predict the details of the divertor plasma under the conditions of the divertor configurations noted above without impurities. This result shows that the peak heat load on an outer target plate of the advanced divertor is about 40% of that of the standard divertor. But more power will be transported to the inner target plate of advanced divertor, and this will cause a higher peak heat load on the inner target plate. The advanced divertor will also have to work under low plasma recycling conditions with high particle temperature and low density in an open divertor target geometry. When the SN configuration changes to a DN tripod divertor configuration, most of the power exhaust is handled by the outer divertor target plates and the peak heat load on these is about 4.1MW/m2 (with a power exhaust of 20MW). This range of optimized divertor configurations and target geometry will enable the study of advanced divertor physics and high performance plasmas in HL-2M tokamak.

Development of microwave interferometer system for divertor simulation experiments in GAMMA 10/PDX

Microwave interferometer has newly been installed on GAMMA 10/PDX for divertor simulation study. A divertor simulation experimental module (D-module) is used to investigate the physics of divertor in the end-cell of GAMMA 10/PDX where an open magnetic field configuration is formed. D-module has a rectangular chamber with an inlet aperture. Two tungsten target plates are mounted in V-shape inside the chamber. In order to develop understandings of divertor simulation experiments the microwave interferometer using heterodyne scheme and a 1D horn-antenna mixer array (HMA) is applied to obtain electron density and density distribution inside the V-shaped target plates. Line-averaged electron density distributions inside D-module are first observed in H2 gas injection experiments.

Alcator C-Mod: research in support of ITER and steps beyond

This paper presents an overview of recent highlights from research on Alcator C-Mod. Significant progress has been made across all research areas over the last two years, with particular emphasis on divertor physics and power handling, plasma–material interaction studies, edge localized mode-suppressed pedestal dynamics, core transport and turbulence, and RF heating and current drive utilizing ion cyclotron and lower hybrid tools. Specific results of particular relevance to ITER include: inner wall SOL transport studies that have led, together with results from other experiments, to the change of the detailed shape of the inner wall in ITER; runaway electron studies showing that the critical electric field required for runaway generation is much higher than predicted from collisional theory; core tungsten impurity transport studies reveal that tungsten accumulation is naturally avoided in typical C-Mod conditions.

DEMO reactor design using the new modular system code SYCOMORE

A demonstration power plant (DEMO) will be the next step for fusion energy following ITER. Some of the key design questions can be addressed by simulations using system codes. System codes aim to model the whole plant with all its subsystems and identify the impact of their interactions on the design choices. The SYCOMORE code is a modular system code developed to address key questions relevant to tokamak fusion reactor design. SYCOMORE is being developed within the European Integrated Tokamak Modelling framework and provides a global view (technology and physics) of the plant. It includes modules to address plasma physics, divertor physics, breeding blankets, shield design, magnet design and the power balance of plant. The code is coupled to an optimization framework which allows one to specify figures of merit and constraints to obtain optimized designs. Examples of pulsed and steady-state DEMO designs obtained using SYCOMORE are presented. Sensitivity to design assumptions is also studied, showing that the operational domain around working points can be narrow for some cases.

20 years of research on the Alcator C-Mod tokamak

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included.

Towards a physics-integrated view on divertor pumping

One key requirement to design the inner fuel cycle of a divertor tokamak is defined by the torus vessel gas throughput and composition, and the sub-divertor neutral pressure at which the exhaust gas has to be pumped. This paper illustrates how divertor physics aspects can be translated to requirements on the divertor vacuum pumping system. An example workflow is presented that links the realization of detachment conditions with the sub-divertor neutral gas flow patterns in order to determine the appropriate number of torus vacuum pumps. For the example case of a fusion DEMO size machine, it was found that 7 actively pumping cryopumps (ITER-type) are necessary to handle the gas throughput that is needed to manage the heat flux and densities related to detachment onset.

Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)

In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription.

Response to “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)

Relying on coil positions relative to the plasma, the “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the “proximity condition,” used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

DEMO design activities in the broader approach under Japan/EU collaboration

The DEMO design studies in the BA (broader approach in the field of fusion energy) are being conducted by the DEMO Design Activity unit of International Fusion Energy Research Centre for the broader approach (BA) and the Home Teams in EU and Japan since 2011. The activity covers most of the critical issues on the DEMO design. Emphasis during the last two years was on studies to develop the best embodiment of a tokamak as a power reactor consistent with credible operating scenarios and feasible engineering solutions to critical design issues. The technical activities have focused on, for example, plasma physics for DEMO plants, divertor physics and technology, in-vessel components, maintenance schemes and safety research.

Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

Progress of the JT-60SA project

The JT-60SA project implemented by Japan and Europe is progressing on schedule towards the first plasma in March 2019. After careful R&D, procurements of the major components have entered their manufacturing stages. In parallel, disassembly of JT-60U has been completed on time, and the JT-60SA tokamak assembly is expected to start in January 2013. The JT-60SA device, a highly shaped large superconducting tokamak with a variety of plasma control actuators, has been designed in order to contribute to ITER and to complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. Detailed assessments and prediction studies of the JT-60SA plasma regimes have confirmed these capabilities: using ITER- and DEMO-relevant plasma regimes, heating conditions, and its sufficiently long discharge duration, JT-60SA enables studies on magnetohydrodynamic stability at high beta, heat/particle/momentum transport, high-energy ion physics, pedestal physics including edge localized mode control, and divertor physics. By integrating these studies, the project provides ‘simultaneous and steady-state sustainment of the key performance characteristics required for DEMO’ with integrated control scenario development.

On gas flow effects in 3D edge transport simulations for DIII-D plasmas with resonant magnetic perturbations

Pumping of neutral gas and re-fuelling in other locations (e.g. neutral beam injection or gas puffing) are key ingredients in divertor physics. However, these have not been included in previous 3D edge plasma transport simulations with the EMC3–EIRENE code. Including these effects brings the simulations closer to reality and we demonstrate that this has a significant impact on divertor parameters. In particular, we study the impact of the pumping efficiency on ITER similar shape plasmas at the DIII-D tokamak in the presence of resonant magnetic perturbations. Further, we investigate the so-called ‘particle pump-out effect’. We show that in the present transport model and for given pumping and re-fuelling rates, the density at the transition to the core region decreases with increasing perturbation current and also the temperature. The striation patterns in the target particle and heat fluxes are extended into the regular scrape-off layer, which is caused by cross-field transport into a thin layer around the perturbed separatrix.

Vacuum Measuring System on the EAST

The Experimental Advanced Superconducting Tokamak (EAST) is the first whole superconducting divertor tokamak commissioned in the world [1]. The first limiter plasma was obtained in 2006 and the single null and double null divertor plasma in 2007. After the 2nd campaign in 2007 with full metal first walls, the plasma facing components were modified to full doped graphite walls and vacuum system was upgraded to meet the requirement of particles exhaust [23]. The vacuum system is one of the most fundamental and important sub-systems of the EAST. The Vacuum Measuring System (VMS) is required not only for the basic vacuum operation, but also for physics research, such as for wall conditioning, wall retention and divertor physics and so on. The VMS for device and vacuum system operation, wall conditioning and physics research will be introduced, in this paper.

Overview of results from the Large Helical Device

The physical understanding of net-current-free helical plasmas has progressed in the Large Helical Device (LHD) since the last Fusion Energy Conference in Geneva, 2008. The experimental results from LHD have promoted detailed physical documentation of features specific to net-current-free 3D helical plasmas as well as complementary to the tokamak approach. The primary heating source is neutral beam injection (NBI) with a heating power of 23 MW, and electron cyclotron heating with 3.7 MW plays an important role in local heating and power modulation in transport studies. The maximum central density has reached 1.2 × 1021 m−3 due to the formation of an internal diffusion barrier (IDB) at a magnetic field of 2.5 T. The IDB is maintained for 3 s by refuelling with repetitive pellet injection. In a different operational regime with moderate density less than 2 × 1019 m−3, a plasma with a central ion temperature reaching 5.6 keV exhibits the formation of an internal transport barrier (ITB). The ion thermal diffusivity decreases to the level predicted by neoclassical transport. In addition to the rotation driven by the momentum input due to tangential NBI, the existence of intrinsic torque to drive toroidal rotation is identified in the plasma with an ITB. This ITB is accompanied by an impurity hole which generates an impurity-free core. The impurity hole is due to a large outward convection of impurities in spite of the negative radial electric field. The magnitude of the impurity hole is enhanced in the magnetic configuration with a large helical ripple and for heavier atoms. Another mechanism for suppressing impurity contamination is identified at the plasma edge with a stochastic magnetic field. A helical system shares common physics issues with tokamaks such as 3D equilibria, transport in a stochastic magnetic field, plasma response to a resonant magnetic perturbation, divertor physics and the role of radial electric field and meso-scale structure.

Particle exhaust and recycling control by active divertor pumping in EAST

The EAST tokamak was built to carry out long pulse plasma discharges up to 1000s. To facilitate this, an active divertor pumping (ADP) system was installed under the lower outer (LO) divertor target plate (DTP) for plasma density control and impurity exhaust. The ADP system was used in the first EAST divertor physics campaign, and achieved very promising results on neutral gas particle exhaust and recycling control. The system can improve divertor screening for neutrals and reduce the ratio H/(H+D), and appears to be effective at controlling the main chamber hydrogenic recycling. The influence of ADP on lower inner (LI) and LO DTP is observed. Furthermore, we attempt to find an optimized strike point to mitigate DTP heat load during ADP operation. This article reports on these interesting results, in particular, for the lower single null (LSN) plasma configuration.

Transport Characteristics in the Stochastic Magnetic Boundary of LHD: Magnetic Field Topology and Its Impact on Divertor Physics and Impurity Transport

Transport characteristics of the stochastic magnetic boundary of the Large Helical Device (LHD) are investigated, based on three-dimensional Monte-Carlo Braginskii-type fluid model code, EMC3, coupled with the kinetic neutral transport code EIRENE, in direct comparison with experimental observations for aspects of the relation between the magnetic topology and the resulting transport in terms of counter acting flux tube flows and impurity screening/transport. Divertor probe measurements show a rather weak divertor parameter dependence on upstream density in contrast to those of tokamaks at high-recycling regime. This is found to be due to the loss of parallel momentum via cross-field interaction between the stochastic flux tubes, where strong flow shear exists. The three-dimensional modeling predicts an impurity screening potential of the stochastic scrape-off layer (SOL) at high densities. The remnant island geometry affects the energy transport, which leads to suppression of the thermal forces by increasing cross-field energy flux across islands at high collisionality. The screening effect is most pronounced at the edge surface layers with a strong friction force exerted by the background plasma flow, where the flow toward divertor is enhanced due to the rich ionization source. Modeling results are compared to the edge carbon emission obtained in experiments, where a reasonable agreement on the density dependence is found, indicating the existence of the impurity screening mechanism in the stochastic SOL of LHD.

Features and initial results of the EAST divertor plasma experiments

The first dedicated divertor physics experiments were carried out on Experimental Advanced Superconducting Tokamak (EAST) in April 2009. Detailed measurements of plasma parameters at the divertor targets were made using an extensive array of divertor diagnostics, and some insights into several divertor physics issues were gained. This work was focused primarily on ohmic discharges. The behavior of divertor plasma, power asymmetry between inner and outer targets, and different fueling methods were investigated in both single null (SN) and double null (DN) divertor discharges. Divertor plasma detachment was achieved, for the first time on EAST, by ramping up the density during the discharge with well controlled, steady divertor configuration. The screening of carbon has been studied by puffing methane into the divertor. In addition, radiative divertor experiments were performed by localized argon injection, leading to a significant reduction in the heat flux on the target plates due to enhanced radiation in the divertor. This provides a means to avoid excess heat load on divertor components, specifically the divertor target tiles, which is essential for high power and long pulse operations, as envisioned for EAST.