Double-null Research Articles

Plasma burn-mind the gap.

The programme to design plasma scenarios for the Spherical Tokamak for Energy Production (STEP), a reactor concept aiming at net electricity production, seeks to exploit the inherent advantages of the spherical tokamak (ST) while making conservative assumptions about plasma performance. This approach is motivated by the large gap between present-day STs and future burning plasmas based on this concept. It is concluded that plasma exhaust in such a device is most likely to be manageable in a double null (DN) configuration, and that high core performance is favoured by positive triangularity (PT) plasmas with an elevated central safety factor. Based on a full technical and physics assessment of external heating and current drive (CD) systems, it was decided that the external CD is provided most effectively by microwaves. Operation with active resistive wall mode (RWM) stabilization as well as high elongation is needed for the most compact solution. The gap between existing devices and STEP is most pronounced in the area of core transport, owing to high normalized plasma pressure in the latter which changes qualitatively the nature of the turbulence controlling transport. Plugging this gap will require dedicated experiments, particularly on high-performance STs, and the development of reduced models that faithfully represent turbulent transport at high normalized pressure. Plasma scenarios in STEP will also need to be such that edge localized modes (ELMs) either do not occur or are small enough to be compatible with material lifetime limits. The high current needed for a power plant-relevant plasma leads to the unavoidable generation of high runaway electron beam current during a disruption, where novel mitigation techniques may be needed. This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

Investigation of performance enhancement by balanced double-null shaping in KSTAR

We report experimental observations on the effect of plasma boundary shaping towards balanced double-null (DN) configuration on the plasma performance in KSTAR. The transition from a single-null to a DN configuration resulted in improved plasma performance, manifested through changes in the pedestal region, decreased density, and core MHD activity variation. Specifically, the DN transition led to a wider and higher pedestal structure, accompanied by grassy edge-localized modes (ELMs) characteristics. The density decrease was a prerequisite for performance enhancement during DN shaping, increasing fast ion confinement. Optimizing the plasma near the core region was associated with the suppression of sawtooth instabilities and the occurrence of fishbone modes during the DN transition. Integrated modeling demonstrated that secondary effects of the DN shaping could increase core thermal energy confinement.

SOLPS-ITER analysis of a proposed STEP double null geometry: impact of the degree of disconnection on power-sharing

To mitigate the issue of plasma exhaust in reactor scale tight aspect ratio tokamaks such as Spherical Tokamak for Energy Production (STEP), a double-null (DN) configuration is thought to be advantageous over a single-null (SN) configuration. However, practical control of the plasma vertical stability will likely lead to an oscillation around the symmetry point, which may lead to transient loading of the divertors. In this work we investigated the impact of disconnection of the two separatrices on the power-sharing between the divertors in disconnected-double-null configurations for the initial iteration of STEP design using the SOLPS-ITER code without drifts. The power fraction to the primary divertor increased with , reaching ∼95% at the highest which is representative of SN. The total power fraction to the inner divertors (upper + lower), however, did not show an increase with for , where λ q is the parallel heat flux decay length, and even at the highest it showed only ∼30% increase from connected-double-null (CDN), unlike the experimental results for current conventional aspect ratio machines. We found two underlying mechanisms that could explain this result—the total flux compression from the outer midplane to the primary inner divertor target and the parallel current in the primary SOL (between the two separatrices). This work implies that the benefit of DN over SN in power load onto the inner divertor in STEP may be less than found experimentally in conventional tokamaks due to its tight aspect ratio. Further investigations through experiments, especially on STs and simulations with additional physics such as drifts, are the subject of a future investigation.

Use of MAXFEA-ANSYS tool to study the Electro-Magnetic behaviour of the new ST40 inner vacuum chamber proposal during a plasma VDE

Tokamak Energy Ltd. is developing compact fusion power plants based on two promising technologies, Spherical Tokamaks (STs) and High Temperature Superconductors (HTS), with the aim to build a first demonstrative plant by the 2030′s. On its path to fusion power, Tokamak Energy is presently operating ST40, a new generation ST that is currently the highest field device of its kind. In the near future, an important ST40 upgrade is planned and foresees the manufacturing and the installation of a new vacuum chamber, called IVC2. This upgrade will allow the exploitation of ST40 to its maximum performances enabling Double Null (DN) diverted operations with up to 2 MA of plasma current and 3 T of toroidal field. IVC2 is currently in the final design stage. The necessity to analyse in detail the Electro-Magnetic (EM) behaviour of this component under the loads coming from plasma disruptions has recently emerged in order to verify the design or drive further modifications. In this context, a study to analyse the EM response of IVC2 during a plasma Vertical Displacement Event (VDE) has been carried out using the MAXFEA code in combination with ANSYS APDL, following a recent workflow development. The main output of this study is the evaluation of the transient EM force density in IVC2 and other components for subsequent mechanical assessments. The aim of this paper is to present the results of this analysis to show the successful application of this procedure in the context of ST40 project.

DSMC simulations of neutral gas flow in the DTT particle exhaust system

Divertor Tokamak Test Facility (DTT) is a new European superconducting tokamak, currently under final design, addressed to investigate alternative power exhaust solutions for DEMO. Although the divertor system is not finalized yet, the machine and port geometry set limitations on the divertor pumping system operational space. A numerical study of neutral gas dynamics in the divertor region is performed based on the DSMC method by applying the DIVGAS code. The study includes both single-null (SN) and double-null (DN) divertor configurations. For both configurations, the SolEdge2D–EIRENE plasma simulations have been performed for a deuterium plasma with neon seeding and the extracted information about the neutral particles on the predefined interfaces is imposed as incoming boundary conditions for DIVGAS simulations. In the SN case, two plasma puffing scenarios and three candidate pumping port arrangements have been considered. The divertor dome influence on the pumped fluxes can reach 50%. An increase of the capture coefficient six times leads to a decrease in the pressure at the pumping openings by a factor of about 4.5–7. The influence of the size of the lower vertical opening has been studied showing that the enlarged vertical port may establish as the main pumping opening. In the DN case, when the pumping is performed from both lower and upper divertor the overall pumped fluxes at the upper divertor are always higher than the corresponding ones for the lower divertor by a factor of 2–2.5, mainly due to the difference in the pumping areas. In both SN and DN cases, the neutrals outflux toward the X-point dominates the particle transport in the private flux region. The operational space provided by this first assessment is relatively stable against modified classical divertor geometries and allows a more thorough assessment of the pumping technology of the DTT fusion device in the future.

Influence of elongation and triangularity on plasma response to resonant magnetic perturbations

Systematic toroidal modeling of the plasma response to the n = 1–4 (n is the toroidal mode number) resonant magnetic perturbation (RMP) field is carried out in order to understand the plasma-shaping effect on controlling the type-I edge-localized modes (ELMs) in tokamak experiments. Considered are large variations of the plasma elongation and triangularity at a fixed edge safety factor q a, for limiter plasmas with both single-null (SN) and double-null (DN) divertor-like boundary shapes. Numerical results assuming conformal 3D RMP coils show that (i) the optimum coil phasing between the upper and lower rows for ELM control becomes increasingly sensitive to the plasma elongation with higher-n toroidal spectra, (ii) the optimum coil phasing is however essentially independent of the plasma triangularity for all n = 1–4 RMP fields, (iii) with the same coil current and the optimum coil phasing, high elongation generally favors ELM control but it may be more challenging for plasmas with intermediate elongation and with lower-n (n = 1–2) RMPs, and (iv) higher triangularity is generally always better for ELM control with all n = 1–4 fields for both DN and SN divertor-like plasma boundary shapes.

Design and feasibility of breeding blanket vertical segment-based architecture

A design integration study (KDII4) was conducted in the DEMO Pre-Concept Design Phase with the primary goal to develop a pre-concept feasibility design and concept of operation for the Vertical Blanket Segment Architecture. The primary goal was to develop two workable variants for removal of large in-vessel components. This, prompted by two earlier studies, highlighting the integration challenge finding a self-consistent DEMO design point (KDII) and secondly the complexity of operation required to ensure compatibly with the proposed port-based maintenance schemes. Initially, only a single null (SN) divertor configuration was considered, but due to additional identified issues and technical challenges centred on the extraction of the BB segments, a double null (DN) alternative variant has also been investigated. This was prompted by other KDII's studies, especially KDII1 (Design, performance and feasibility of wall protection limiters during plasma transients) and KDII3 (Advanced Magnetic Configurations). Alternative forms of vertical maintenance architecture have been investigated by breaking down the study into Ports, In-vessel Components, Operations and Safety. SN and DN configurations were studied with split and full blankets numbering seven variants in total. Each variant was studied in some detail, comparatively assessing each on its merits. However, no viable solution has been found for down-selection. Nevertheless, the question of vertical maintenance is now better understood. This paper will describe the findings of KDII4 in conjunction with the Remote Maintenance (RM) technology work package, which suggests that the current ‘reference’ design envelope available for RM is too constrained.

Detachment in conventional and advanced double-null plasmas in TCV

Divertor detachment is investigated on the Tokamak à Configuration Variable (TCV) for conventional and alternative double-null (DN) magnetic geometries in L-mode, Ohmic density ramps and then compared to precisely matched single-null (SN) geometries. The poloidal flux expansion at the outer strikepoint(s) is varied by a factor 2 and the total flux expansion by 30% in lower single null (LSN) and DN configurations. Leg power sharing for DN and near-DN in attached conditions is in line with previous studies in other devices, with balanced power partition between upper and lower divertor achieved for |δR sep| ≲ λ q /2. This results in a reduction in the lower outer target integral and peak heat flux compared to a LSN by 30%. Unlike previous studies, λ q is fairly insensitive to δR sep. The detachment threshold in these geometries is investigated from target measurements with wall-embedded Langmuir probes and two-dimensional CIII line emissivity profiles across the two divertor regions. DN plasmas display clear benefits compared to their LSN counterparts. Transitioning from a LSN to DN shows a substantial reduction, of approx. 20%–25%, in the detachment threshold for the active divertor legs as well as a 50% higher radiated fraction at all . The density limit is simultaneously reduced in DN by 10%–20%. Across the explored range, poloidal flux expansion has only a small effect on the detachment threshold in LSN (as seen in previous experiments), and no effect in DN, with similar observations for the total flux expansion. In general, no strong benefits of increased poloidal or total flux expansion are observed across the explored range.

Influence of up-down asymmetry in plasma shape on RMP response

Shaping effect on the plasma response to the externally applied resonant magnetic perturbation field is investigated for both DIII-D and MAST experiments, utilizing toroidal modeling. The plasma boundary shape is systematically varied ranging from single-null (SN) to double-null (DN) configurations, while other equilibrium quantities are kept largely unchanged. The relative amplitude of the computed plasma surface displacement, between the top/bottom of the torus and the outboard mid-plane, is identified as the most reliable indicator that distinguishes the plasma response between the SN and DN configurations. The underlying physics is the weakening of the edge-peeling component in the plasma response with increasing up-down symmetry of the plasma boundary shape.

Assessment of measurement performance for a low field side IDTT plasma position reflectometry system

The Italian Divertor Test Tokamak (IDTT) facility will study advanced exhaust solutions applicable to DEMO. This new machine also opens the possibility to test and validate relevant non-magnetic control diagnostics in support of a DEMO design implementation. Reflectometry, compatible with a full grade reactor implementation, has been proposed as a source of real-time (RT) plasma position and shape measurements for control purposes, in replacement or complement of standard magnetic measurements, to ensure reliability and safety on the machine. Additionally, such a system can be useful to obtain the characteristics of edge turbulence, useful for Neoclassical Tearing Mode (NTM) control or to improve the efficiency of other diagnostics as Collective Thomson Scattering (CTS). This new control technique based on multiple, poloidally distributed, non-magnetic measurements, must be tested, in all of its aspects, before it can be fully implemented in future fusion reactors. IDTT will be one of the best candidates to implement and build a knowledge database of nonstandard reflectometry (away from the equatorial plane) that will be needed on DEMO and is currently unavailable. The performance of three Ordinary mode (O-mode) Plasma Position Reflectometers (PPR) at the Lower Field Side (LFS) on IDTT is assessed using the two-dimensional (2D) full-wave Finite-Difference Time-Domain (FDTD) code, REFMULF, in two of the foreseen IDTT plasma scenarios: 5MA Single Null (SN) and Double Null (DN) configurations.

Preliminary engineering assessment of alternative magnetic divertor configurations for EU-DEMO

• The present work deals with a preliminary engineering development of alternative magnetic divertor configurations for EU-DEMO. • Parametric CAD master models for each configuration have been developed, including PF and TF coils, VV, ports and in-vessel components. • The configurations have been analyzed in terms of geometrical envelope and ports layout, considering structural and Remote Handling (RH) requirements. One of the main challenges in the roadmap to the realization of fusion energy is to develop a heat and power exhaust system able to withstand the large loads expected in the divertor of a fusion power plant. The challenge of reducing the heat load on the divertor targets is addressed, within the mission 2 ‘Heat-exhaust systems’, through the investigation of divertor configurations alternative to the standard Single Null (SN), such as the Snowflake (SF), Double Null (DN), X and Super-X (SX) divertors. This paper focuses on a preliminary engineering assessment of the alternative configurations proposed for the EU DEMO reactor. Starting from the description of the optimized plasma shape developed for each configuration, the 3D geometrical description of the Magnet System and of the main Mechanical Structures (Vacuum Vessel and in-vessel components) is presented. Based on the 3D geometry, the compatibility of the location and dimensions of ports with Remote Maintenance needs is discussed and possible design optimizations are proposed both for the Magnets system and the mechanical structures design. Finally, the various configurations are compared with regard to the engineering and feasibility aspects.

Dependence on plasma shape and plasma fueling for small edge-localized mode regimes in TCV and ASDEX Upgrade**In the future we will refer to the author list of the paper as the EUROfusion MST1 Team

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened.

Electromagnetic analyses of single and double null configurations in DEMO device

The DEMO design has to face a number of challenges. According to the DEMO device limitations, the power load impinging on the Plasma Facing Components (PFCs) in the plasma Single Null (SN) magnetic configuration (which is for DEMO as in ITER the baseline configuration) clearly underlines the need to explore alternative divertor concepts. In this paper, a comparison between the Single and Double Null (DN) configurations is proposed. In particular, a comparative electromagnetic analysis of the magnetic configurations is performed in terms of vertical stability (VS) and disruptions in the 2D axisymmetric case.

Effects of carbon impurities on the power radiation and tungsten target erosion in EAST

Tungsten (W) and Carbon (C) are, respectively, used as the plasma-facing materials for upper and lower divertors in EAST. This provides an opportunity to study the plasma performance with both W and C divertors simultaneously. In this work, simulations are carried out by using the edge fluid-plasma/kinetic-neutral code SOLPS5.0-EIRENE99, with emphasis on the C impurity transport and its effect on the power radiation and erosion of the upper W divertor target. Three typical magnetic field equilibrium configurations, i.e., lower single null, upper single null (USN), and double null (DN), are studied. The main deuterium (D) plasma and C impurities at each ionization state and their neutrals are considered in the simulation. The SOLPS output such as averaged incident ion energy and particle flux density of each species are used to calculate the erosion rate of W walls using the empirical formula for physical sputtering yield. The transport and performance of C impurities on different plasma conditions of each configuration are investigated by changing the C source strength and upstream plasma density. The upper W target erosion by C ions under different conditions is also evaluated. The simulation results indicate that (1) carbon is a strong radiator in EAST divertor temperatures of 10–20 eV; (2) the amount of carbon transported from the lower carbon divertor to the upper tungsten divertor in both USN and DN configurations is predicted to be too small to cause significant W sputtering; and (3) C6+ dominates upper W erosion in the DN case.

Tritium breeding performance of a DEMO based on the Double Null divertor configuration

Different design modifications of a Double Null (DN) DEMO were analysed by means of 3D Monte Carlo calculations using the MCNP code to assess the effect of a DN divertor configuration and various in-vessel components (IVC) on the tritium breeding performance. A simplified DN DEMO model was set up to this end. A Tritium Breeding Ratio (TBR) of 1.14 was obtained with this model for the HCPB breeder blanket concept without taking into account any auxiliary equipment such as limiters or port plugs. The inclusion of such design modifications results in a significant reduction of the TBR for different breeder blanket concepts. To compensate the loss of breeder space in the reactor either the integration of breeder materials in the divertor body or the enlargement of the available breeder zone could be considered. Both of these options can provide an increase of the TBR and ensure a sufficient margin for a DN DEMO with a HCPB breeder blanket. Meeting the TBR requirement with a Pb-Li based breeder blanket concept would be more challenging and requires significant changes to the overall IVC configuration.

Investigation of the effects of impurity seeding under different magnetic configurations in L-mode plasma in EAST tokamak

Experiments with a small amount of Ar puffed from the lower outer target into L-mode plasma in double null (DN), lower single null (LSN), and upper single null (USN) configurations studied the effects of magnetic configurations and the seeded location relative to the X-point on the performance of Ar seeding. With Ar seeded, the peak heat flux and the integrated power load onto the target decreased most significantly in the LSN configuration, although the increase of the line integrated radiation in the main plasma was the lowest in this case and the highest in the USN configuration. The radiation fraction in the divertor rose from about 10% to 23% of the heating power in LSN, and from around 15% to 22% in DN, while it almost did not change in the USN configuration with Ar seeding. The line emission of Ar XVI increased following the puff and then started to degrade near the end of LSN and USN phases. No such degradation was seen in the DN phase, meaning that some Ar dwelt in the core till the LSN phase began, which explained the peaking of radiation in the core prior to the seeding in the LSN phase. Dedicate Modeling indicates that the LSN configuration has the highest divertor radiation fraction. The highest radiation from outside the divertor is in the USN configuration. With increasing Ar seeding, the LSN configuration sees an increase of the radiation inside the divertor while in the USN configuration, it is the radiation from outside the divertor that increases most strongly. In the DN configuration, the radiation from both inside and outside the divertor increases with seeding, with the former more strongly.

Creation and Characterization of a Double Null Puroindoline Genotype in Spring Wheat

Wheat (Triticum aestivum L.) grain hardness is controlled by the Ha locus, which is composed of two closely linked genes, Puroindoline a (Pina) and Puroindoline b (Pinb). Hard grain results from mutations in either of the Pin genes. Previous results have shown that the Pina‐D1b (Pina null) allele has harder grain than other naturally occurring Pin alleles. Our goal was to create, identify, and characterize a double null Pin genotype by identifying a Pinb null mutation in a variety carrying the Pina‐D1l null allele. Seeds of Fortuna, which has a premature stop codon in Pina, were treated with ethyl methanesulfonate. Two premature stop codon mutations were identified in Pinb using direct sequencing. The double null Pin haplotype was characterized after backcrossing to the parent variety Fortuna to create Pina null populations segregating for the presence of Pinb. The double null group was 6 units harder than the single null with no difference in other kernel characteristics. The milling characteristics differed between the two classes; the double null class had less break flour with a greater fraction of large and a smaller fraction of small flour particles compared with the single null class. Neither water absorption nor loaf volume was impacted by the change in grain hardness; however, Na2CO3 tests indicated greater starch damage in the double nulls. The double null Pin genotype may find a niche in hard wheat products for which flours with larger particle size are desired.

Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG) to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF) and the X-divertor (XD), that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN) and double-null (DN) would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

Achieving temporary divertor plasma detachment with MARFE events by pellet injection in the EAST superconducting tokamak

A new pellet injection system has been equipped on the experimental advanced superconducting tokamak (EAST) in the 2012 campaign, with a pellet size of ϕ 2 mm × 2 mm, a frequency of 1 Hz–10 Hz and velocity of 150 m s−1–300 m s−1. The deuterium pellet is well-known for plasma fuelling as well as for triggering the edge localized mode (ELM). In the 2012 campaign, pellet injection experiments were successfully carried out on EAST. Temporary plasma detachment achieved by deuterium pellets has been observed in a double null (DN) divertor configuration, with multi-pellet injections at a repetition frequency of 2 Hz. The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection, which have a duration of 30–60 ms with the maximum degree of detachment (DOD) reaching 3.5 and 37, respectively. Meanwhile, the multifaceted asymmetric radiation from the edge (MARFE) phenomena was also observed at the high field side (HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%–70%, which may be the main cause of divertor plasma detachment. The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.

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.