20th International Spherical Torus Workshop (ISTW2019)

28 Oct 2019, 07:40 → 31 Oct 2019, 16:30 Europe/Rome

23/3-02 - FSN Meeting room (Frascati)

Franco Alladio (ENEA-FSN-Fusphy)

 

 

Sponsored by

ISTW 2019 BOOK OF ABSTRACTS.pdf

ISTW-2019_Program_v5.pdf

posterISTW2019.png

28 october

• 2019 Diem ISTW Presentation updated.pdf
• AJThornton_ISTW_final.pdf
• AJThornton_ISTW_v3.pptx
• Buratti_talk_28Oct19.pdf
• ISTW2019-Alladio.pdf
• ISTW2019-Alladio.pptx
• ISTW-2019-majeski.pdf
• ISTW2019_ProtoSphera_Lampasi_29Oct2019.pptx
• Lloyd_ISTW2019_v2.pptx
• Raman-Pegasus-ISTW-V2.pdf
• Report_Globus-M2 experiments at 0.7 T.pdf
• ST_workshop_kurskiev_4.1_off.pdf

29 October

• 20191029-tanaka-v2.pdf
• McNamara_TE_Overview_ISTW2019.pdf
• ST40 Gryaznevich STW2019.pdf
• VEST_ISTW2019_YSHwang_final.pdf

30 October

• Bertelli_STworkshop_Frascati_2019.pdf
• Bertelli_STworkshop_Frascati_2019.pptx
• Menard_ISTW2019_v4.pdf
• Neutral particle - electron collision effect for branching plasma current ramp-up on 28 GHz 2nd harmonics ECCD.pptx
• NSTXU_OV.pdf
• Overview of Recent Progress on the QUEST experiments_Presentation.pptx
• PICCIONE_ROME_301019_v3.pptx
• Raman-CHI-ISTW2019-V6.pptx
• ruizruiz_istw2019_frascati_draft0.pdf

Monday, 28 October

08:00 → 09:00 Registration form

09:00 → 09:15 Logistics and meeting introduction

09:15 → 12:45 Session O1

09:15 Final results of the Phase-1 of the PROTO-SPHERA experiment and first evidences of plasma tori production and sustainment

The PROTO-SPHERA experiment is an unconventional magnetic confinement scheme,
which aims at producing a Spherical Torus (with Ie≤300 kA) around a Plasma Centerpost (with Ie=70 kA) fed by electrodes of annular shape, in contrast with the metal centerpost of conventional Tokamaks. The Phase-1 of PROTO-SPHERA, aimed of obtaining a stable Hydrogen Plasma Centerpost, lasting 1 sec at Ie=8.5 kA longitudinal current level, was successfully ultimate at the begin of 2018. The line averaged plasma electron density on the equatorial plane of the Plasma Centerpost increases linearly with Ie and reaches, at Ie=10 kA in Argon, the considerable value of <ne>=4•1020 m-3, similar to what is obtainable today only in high-field Tokamaks, while in Hydrogen the still respectable value is <ne>=1.5•1020 m-3 at Ie=10 kA; spectroscopic measurements show very pure plasma when we operate in Hydrogen (Balmer lines only), and Langmuir probe measurement show an edge temperature of 4-8 eV and an edge density of few 1019 m-3 in proximity of the anodic plasma when we operate in Argon. The Plasma Centerpost show a strong toroidal rotation, due to E∧B, that prevents any anode anchoring. During the spring/summer 2018 the kink destabilization of the PROTOSPHERA Plasma Centerpost and the plasma tori production has already been preliminarily obtained in the actual Phase-1 of the experiment, by adding 4 impromptu vertical field external PF coils, that are wound outside the vacuum vessel and fed in series with a new power supply based on SuperCapacitors. The result in Hydrogen has been a disruption-free coupled configuration, obtained with stationary poloidal magnetic field, in which a very high aspect ratio (A=R/a≈7) and very high elongation (κ≈3) torus of approximately IST≈7 kA is sustained by Helicity Injection, with MHD bursts with a periodicity of ≈3 ms, around the Ie=10 kA Plasma Centerpost until the DC voltage VPC between electrodes is turned on, i.e. for more than 250 ms. In 2019 the PROTO-SPHERA experiment has been largely modified: a new insulating/transparent vacuum chamber in PMMA substitute the START Aluminum vessel and 6 internal PF coils has been inserted inside the machine. A "Phase-1.5" campaign is now in progress with the aim to produce/sustain, still at the Ie=10 kA level of Plasma Centerpost current, an elongated (κ≈2) low aspect ratio torus (A≤2) with a current of the order of IST≈20 kA and a volume of the closed magnetic surfaces greater than 50%.

Speaker: Dr Paolo Micozzi (ENEA C.R. Frascati)

09:50 Present and future plans of PROTO-SPHERA

PROTO-SPHERA is an axi-symmetric magnetic confinement experiment, just like a spherical Tokamak: toroidal and poloidal magnetic fields (i.e. a plasma toroidal current IST) are required inside a spherical torus to contain the plasma. PROTO-SPHERA produces both fields while removing the two central metal conductors: the central toroidal rod and the central ohmic transformer; they are replaced by a plasma centerpost discharge, with helical magnetic force lines, carrying an electrode current Ie, fed by DC electrodes inside the vacuum vessel. PROTO-SPHERA however is not a Spheromak, as no metal flux conserver is present around the spherical plasma. While the presence of the toroidal field in PROTO-SPHERA is obvious, provided that the electrodes produce a stable current carrying centerpost plasma, the sustainment of the current carrying torus by DC electrodes is not obvious and it has been the first achievement of the experiment. In its initial Phase-1 (2014-2018) PROTO-SPHERA was built with 8 magnetic PF coils inside the vacuum vessel, to shape the plasma centerpost only; the electrode current reached in Hydrogen its Phase-1 limit: Ie=10 kA. By adding 4 PF coils outside the vacuum vessel, the first “thin” confined tori (that fill just 7% of the total plasma volume), endowed with divertors emerging from magnetic X-points, were formed and sustained, with Ie=10 kA and IST=7kA. The IST=7 kA toroidal current could be maintained in quasi-steady state as long as the anode-to-cathode DC voltage was applied (1 sec): it is highly plausible that the magnetic reconnections, observed around the X-points, were able to provide the steady-state current drive to the high density (>1020 m-3) confined tori. In the 2019 intermediate Phase-1.5, with Ie still limited to 10 kA, the confined tori will be compressed to low aspect-ratio, and their toroidal plasma current will be increased to IST=20-40 kA: 6 provisional internal PF compression coils have been added, bringing the total number of PF coils (internal and external) to 18. In Phase-2 (2021) the centerpost plasma current Ie will grow from 10 kA to 70 kA; the spherical shaping should allow for a confined toroidal current of IST=300 kA. Hopefully Phase-2 of PROTO-SPHERA will add to the magnetic reconnection current drive, already obtained on “thin” tori, a vigorous and self-organized plasma reconnection heating: a total power of ~20 MW will be input into the overall plasma, while the spherical tori will fill up to 95% of the total plasma volume.

Speaker: Dr Franco Alladio (Centro Ricerche ENEA, Frascati C.P. 65, Rome, Italy)

10:10 Technical Benefits of the PROTO-SPHERA Confinement Configuration

The PROTO-SPHERA experiment can produce plasma spherical tori with a simply connected configuration (the centerpost is practically implemented by the plasma itself) and a high confinement efficiency (β→1). Such configurations are obtained from the self-organization of a plasma arc (screw pinch) in a magnetostatic field. This approach leads to several technical and economic advantages, that even exceeded the initial theoretical expectations.
The experimental system can be operated by only 3 groups of power supplies (for cathode heating, poloidal field and pinch, respectively) with a virtually constant current and with a limited demand of electrical energy. In fact, the recent experiments showed that the plasma tends to a toroidal (tokamak-like) shape even without the very fast current derivatives predicted in the initial design. The plasma current is induced and confined without a toroidal magnet and without variations in the magnetic flux and in the poloidal coil currents. The maximum voltage applied in the structure, even at the arc breakdown, does not exceed 350 V. A high temperature can be reached without additional heating and current drive systems.
Axisymmetric double null configurations with electron density in the order of 1020 m-3 were already observed. The experiments also showed that the PROTO-SPHERA configurations are stable as long as the pinch voltage is applied. The plasma is not terminated nor disrupted even in presence of sudden deformations of the pinch. Therefore, the set-up is consistent without any further coils or radiofrequency sources to cope with plasma instabilities.
The vessel is cylindrical and it can be totally opened and closed in few days for maintenance and modifications. For example, the adjustment of the poloidal field by the introduction of new internal and external coils is relatively simple, as already done with several external coils during the experimental campaigns. Since 2019 the middle part of the cylindrical vessel is non-conducting and transparent in order to avoid undesired couplings for the plasma arcs or for the magnetic field produced by the external coils and in order to make the plasma configuration clearly visible for the researchers and for the diagnostic tools.
The insertion and expulsion of elements, as pinch current or gas, through the configuration follow a simple and well defined path. The fusion products could be ejected through a magnetic nozzle on the cylinder axis (with possible applications in spacecraft propulsion).

Speaker: Alessandro Lampasi (ENEA)

10:30 Analytical studies of PROTO-SPHERA equilibria

Plasma tori formed around a 10 kA centerpost screw-pinch plasma discharge have been observed in phase-1 of PROTO-SPHERA experiments.
A simple model of plasma equilibrium has been developed, which reproduces the morphological features observed so far, in particular the bumpy shape of the centerpost discharge and the slim torus surrounding it.
The model is based on a generalization of the bumpy pinch equilibrium solution of the Grad-Shafranov equation by Taylor [Jensen T.H. and Chu M.S. 1980 J. Plasma Phys. 25 459], which allows reproducing both bumpy pinch and pinch-torus force-free configurations. The Taylor model in cylindrical coordinates is expressed as: $\psi=rJ_1(Kr)+crJ_1(\sqrt{K^2-k_z^2}\,r) \cos(k_zz)$, where $\psi$ is the poloidal flux function and $K$ is the coefficient of proportionality between poloidal current and poloidal flux. The constant $c$ determines the bumpiness of the pinch: bulgy pinches result for $c<0$, while $c>0$ gives pinch-torus combinations with squeezed pinch.
The above expression has been generalized along three lines. First, in order to describe low-current plasmas (including the vacuum case) it has been extended to $K< k_z $. Second, more $z$-dependent terms have been added, which allow obtaining bulgy pinches surrounded by slim tori. Third, a plasma pressure term $p=p_0-\alpha \psi$ has been included. Plasma pressure in the pinch is not negligible, in fact it is sufficiently large to reverse the sign of the azimuthal current with respect to the force-free case.
Analytical solutions have also been found for the spherical G-S equation force-free conditions and with poloidal current proportional to poloidal flux (i.e. relaxed), in the form of combinations of Bessel functions of half-integer order in the spherical radial coordinate and Gegenbauer functions (for example solutions of hydrodynamical equations for symmetric Stokes flows) in the angular coordinate. These functions can be used to construct solutions of boundary-value problems for PROTO-SPHERA.

Speaker: Dr Paolo Buratti (ENEA)

10:50 Coffee Break

11:15 MAST-U experimental programme: From first plasma to first campaign

The first stage in the experimental programme of MAST-U is planned for late 2019. A modelling framework, using time-dependent vacuum field calculations, has identified viable direct induction scenarios on MAST-U that achieve suitable null quality for rapid breakdown combined with the requirements for plasma equilibrium and passive vertical stability during the early plasma current ramp up [1]. The resulting discharge parameters provide breakdown with a loop voltage of 4.5V, which around half of the total available. The loop voltage gives an electric field of 1.5-2 V/m and a null is formed with a connection length of between 400 and 800 m [1]. The null is formed with the poloidal D coils carrying currents of 5 kA or less, which is well within the capabilities of the supplies. Key to the success of MAST-U is the ability to form advanced divertor configurations, specifically the Super-X divertor [2]. Using a free boundary equilibrium solver (FIESTA) and representative current profiles, a 700 kA plasma current, double null MAST-U discharge has been simulated (κ = 2, li=0.94). The simulated discharge allows understanding of the coils required to form and control a conventional divertor and then shape this into a Super-X configuration. The elongation and internal inductance operating space has been investigated by generating a database of scenarios that will guide scenario development. MAST-U has an experimental campaign covering pedestal physics, fast particles, scenario development and exhaust physics. The focus is on the exhaust physics area, specifically understanding the advantages and disadvantages of the Super-X divertor. The increased divertor radius is thought to encourage detachment as the target temperature is proportional to 1/Rtgt2 [3]. SOLPS modelling [4] shows a reduction of the detachment threshold by a factor 2.4 by changing the target radius by 0.7m. Initial experiments will aim to test model predictions through development of conventional and super-X scenarios; utilising the extensive range of diagnostics available on MAST-U to measure the radiated power, target heat loads and temperatures.
[1] D Battaglia et al, Nucl. Fusion, accepted for publication, 2019
[2] Valanju et al, Phys. Plasmas 16 (2009) 056110
[3] TW Petrie et al, Nucl. Fusion 53 (2013) 113024
[4] D Moulton et al, Plasma Phys. Control. Fusion 59 (2017) 065011
This work has been funded by the RCUK Energy Programme [grant number EP/P012450/1]

Speaker: Dr Andrew Thornton (UKAEA)

11:50 MAST Upgrade - features and status

Presentation on the features and status of MAST Upgrade

Speaker: Dr Brian Lloyd (UKAEA)

12:10 First operation of the Lithium Tokamak Experiment - β

LTX-β, the upgrade to the Lithium Tokamak Experiment, has now been operated with full lithium coatings on all plasma-facing surfaces, and at increased toroidal fields >0.3 T. Plasma current has so far been limited to <100 kA. The upgrade includes a neutral beam injector provided by Tri-Alpha Energy Technologies. The beam is designed to operate at 20 kV, with 35A of ion current. So far >600 kW of beam power has been injected at 18.5 kV. Up to 60% of the injected power is deposited in the plasma at higher target densities, in agreement with NUBEAM modeling. However, initial modeling of the ion orbits indicates that fast ion losses are important. Significant beam fueling is observed under some conditions. New insertable lithium evaporators have been installed on LTX-β, which provide full coverage of the plasma-facing surfaces, with a rapid (10-15 minute) evaporation cycle. Additional improvements to the evaporators, including a larger lithium inventory, and between-shots operation, are planned. LTX-β retains the same plasma geometry, and the heated high-Z liner featured in LTX. Upgrades to the diagnostic set include active CHERs. Further upgrades to the Thomson scattering system and new Lyman- α arrays will permit a determination of energy confinement time as a function of recycling, which is the emphasis for the 2019 – 2020 experimental campaign.

Speaker: R. Majeski (Princeton Plasma Physics Lab)

12:45 → 14:00 Lunch

14:00 → 17:00 Session O2

14:00 Integrated Studies of Solenoid Free Tokamak Startup

The Pegasus Toroidal Experiment has focused on exploring the unique features of extremely low aspect ratio operation to study plasma stability at A ~ 1 and near-unity beta, including access to ohmic H-mode with associated ELM behavior and plasmas with an absolute minimum-B topology. For such studies, the Local Helicity Injection (LHI) technique was developed to initiate and drive toroidal current without use of a central solenoid. LHI utilizes compact, edge-localized current sources (Ainj ≤ 8 cm2, Iinj ≤ 8 kA, Vinj ≤ 1.5 kV), and has initiated > 0.2 MA of plasma current at low-field (BT ~ 0.15 T) and near-unity aspect ratio (A). Recent studies have addressed comparisons of LHI injector locations, MHD characteristics, confinement properties, and the transition to ohmic sustainment. The Pegasus program is now terminating and being replaced with an upgraded facility called Pegasus-III (for Pegasus-Phase III). This new program is focused on developing an understanding and comparative studies of several leading candidates for non-solenoidal startup techniques. These include: DC helicity injection, including LHI and sustained and transient coaxial helicity injection (S-CHI and T-CHI); electron Bernstein wave (EBW) radiofrequency electron heating and current drive; and poloidal field induction. Future capabilities may include electron cyclotron heating and current drive and/or neutral beam current drive. Upgrades include: a new solenoid-free centerstack assembly to provide a 4x increase in BTF to 0.6 T; a new high-stress TF coil system; increased pulse length to study confinement and current evolution; expanded power systems; enhanced diagnostics; and an expanded lab facility. Plasma startup will be enabled by: advanced LHI injectors with shaped electrodes and active helicity injection control; a refractory metal electrode system supporting S-CHI and T-CHI without vacuum electrical breaks or current flow through the vacuum vessel; and an expanded poloidal field coil set. An 8 GHz, ~400 kW EBW system will be added. A goal of the new experiment is to demonstrate non-solenoidal startup to 𝐼p = 0.3 MA and provide guidance for MA-class startup applicable to NSTX-U and ultimately a nuclear facility. This program includes collaborations from the University of Wisconsin, University of Washington, ORNL, and PPPL.
*This work is supported by U.S. Department of Energy grants DE-FG02-96ER54375 and DE-SC0019008.

Speaker: Raymond Fonck (University of Wisconsin-Madison)

14:35 Microwave Heating and Current Drive of Overdense Plasmas in the Upgraded Pegasus Experiment

Identifying attractive means of initiating current without using induction from a central solenoid remains a critical challenge facing the spherical tokamak (ST) concept, and is desirable for tokamaks in general. An electron Bernstein wave (EBW) heating and current drive system is being designed for use in overdense ST plasmas on the upgraded Pegasus Toroidal Experiment. The research program on Pegasus is studying the physics of non-solenoidal plasma startup, ramp up and sustainment in the ST geometry. Recent startup research on Pegasus has focused on Local Helicity Injection (LHI). This effort will be expanded upon by an EBW heating and current drive system to provide viable means for non-inductive startup and sustainment of ST plasmas. Additionally, EBW has the potential to reduce resistive losses during LHI startup by increasing the electron temperature, simplifying injector requirements, and may also offer current profile control capabilities for transition to post-startup sustainment methods. The upgraded Pegasus Experiment will increase the toroidal field by a factor of 4, allowing for EBW experiments with a source frequency of 8 GHz, providing absorption near the fundamental electron cyclotron resonance. The flexibility of the university-scale experiment allows for the ability to explore different coupling schemes such as low-field side O-mode to X-mode to EBW coupling as well as high field injection of the slow X-mode to directly couple to the EBW.

This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under contract DE-AC05-00OR22725.

Speaker: Dr S.J. Diem (Fusion Energy Division, Oak Ridge National Laboratory)

14:55 CHI Research on PEGASUS 3

The spherical tokamak (ST) may require and the advanced tokamak would considerably benefit from the elimination of the central solenoid. PEGASUS-3 is a ST non-solenoidal startup development station under design and fabrication dedicated to solving the startup problem. On PEGASUS-3, Transient and Sustained coaxial helicity injection (T- and S-CHI) will be explored, as well as possible synergies of CHI with local helicity injection and EBW heating and current drive. T-CHI has shown promising capability on the HIT-II and NSTX STs. However, in both these machines the vacuum vessel was electrically cut. For reactor applications a simpler biased electrode configuration is required. To develop this capability a single biased electrode is being tested on QUEST, where up to 45 kA of toroidal current has been generated using CHI. PEGASUS-3 will use a more advanced double biased electrode configuration with optimized injector electrodes and injector poloidal field coils that should allow the T-CHI system to generate 0.3 MA of closed flux current, the limit permitted by the equilibrium PF coils. Present design indicates that standard divertor coils will provide sufficient flux for CHI studies but may be enhanced with increased current capabilities if needed. The CHI design and the CHI research plan for PEGASUS-3 will be described.

Speaker: Dr Roger Raman (University of Washington)

15:20 Coffee Break

15:50 Globus-M2 operation at toroidal magnetic field strength 0.7 T

Described are first results obtained on the Globus-M2 tokamak – upgraded version of Globus-M machine in spring-summer 2019 operating campaign. Operation was performed at toroidal magnetic field strength 0.7 T and plasma current 0.3 MA. Results of the campaign may be tentatively joined into a few groups. First group is related to initial stage of the discharge. Plasma breakdown conditions were improved noticeably with regard to Globus-M breakdown conditions. Stray fields at the breakdown phase became lower due to careful manufacturing and assembly of Globus-M2 electromagnetic system. The breakdown loop voltage decreases from 7 to 5 Volts (without usage of special preionization technique). Plasma ramp-up speed with 5-7 MA/s was obtained. Second group is related to OH current plateau stage and characterized by relatively low sawtooth activity. Neutral beam auxiliary plasma heating period is the third group which demonstrated very intriguing results. NBI plasma heating and current drive became more efficient at the same NB injector parameters as in Globus-M (28 keV, 0.8 MW). Plasma electron temperature during NBI pulse rose to 800 eV simultaneously with the density rise from 3x1019m-3 to 7x1019m-3. Diamagnetically measured plasma energy content increased from 3 kJ up to 6 kJ which is twice as high as in Globus-M. Ion temperature reaches 1.2 keV at the end of NB heating pulse. H-mode confinement with H factor of about 1.2 was achieved. Neutron flux rise was recorded during D beam into D plasma injection. More than doubling of neutron flux was achieved comparatively to Globus-M. And final group of results is connected with noninductive current drive. Loop voltage drop was recorded during NB injection indicating noticeable amount of non-inductively (mainly bootstrap) driven current. For the first time in spherical tokamaks noninductively driven current was recorded during LH range electromagnetic waves launch with toroidally oriented grill. During RF pulse launched by 10 waveguide grill (2.45 GHz) with the phase delay of 1200 between waveguides and total radiated power of about 150 kW ~ 30% loop voltage drop was recorded.
Achievement of Globus-M2 operation parameter limit (1 T and 0.5 MA) is planned for the later period after a full output power from thyristor rectifier supply will be accessible.

Speaker: Vasily Gusev (Ioffe institute, St. Petersburg, Russian Federation)

Report_Globus-M2 experiments at 0.7 T.pdf

16:25 Thermal energy confinement at the Globus-M spherical tokamak and first results from the Globus-M2 experiments

The presentation is devoted to the thermal energy confinement study at the compact spherical tokamaks Globus-M and Globus-M2. Experiments were performed under auxiliary heating using neutral beam injection (NBI) in plasma with lower null magnetic configuration (major radius R = 0.35 m, minor radius a = 0.21-0.22, elongation ~1.9, triangularity δ~0.35) for the ranges of plasma current and toroidal magnetic field: Ip=0.12-0.25 MA, BT=0.25-0.5 T. It have been shown that energy confinement time (τE) dependence on BT is very strong, while the τE dependence on plasma current Ip is significantly weaker than IPB98(y,2) scaling predicts: τE~I_p^(0.48±0.21) B_T^(1.28±0.12). The improvement of τE was mostly by electron heat diffusivity decrease with toroidal field rise, while the ion heat diffusivity was in line with neoclassical theory predictions.
The first NBI experiments were carried out at the Globus-M2 for the increased range of plasma current and toroidal magnetic field: Ip =0.25-0.3 MA and BT = 0.7 T. During NBI heating (D-beam, 28 keV, 0.8 MW) the plasma total stored energy measured by the diamagnetic coil increased more than twice (in comparison with the Globus-M results). Diamagnetic measurements were confirmed by the kinetic (electron and ion temperature profiles) measurements. Thermal energy confinement time was estimated by 1.5D ASTRA transport modeling while the beam absorbed power was derived using two codes: NUBEAM code and 3D fast ion tracking algorithm. The obtained τE values are higher than those predicted by IPB98(y,2) and are in good agreement with the Globus-M scaling.

Speaker: Gleb Kurskiev (Ioffe institute)

ST_workshop_kurskiev_4.1_off.pdf

Tuesday, 29 October

08:00 → 09:00 Registration

09:00 → 13:00 Session 03

Convener: Stephanie Diem (ORNL)

09:00 Tokamak Energy and the high-field spherical tokamak route to fusion power

Tokamak Energy is a privately funded company based in the UK with a mission to deliver a faster route to fusion. Founded in 2009, Tokamak Energy is developing compact fusion power plants based on two promising technologies: Spherical Tokamaks (STs) and magnets made from High Temperature Superconductors (HTS). The inherent compactness and improved efficiency of the spherical tokamak, coupled with the favourable properties of HTS magnets, open a route to efficient power production at significantly lower net power outputs than previously considered possible.

Currently, two main development streams are progressing in parallel: i) advancing high field spherical tokamak physics and engineering on ST40, the highest field (BT=3T) device of its kind; and ii) HTS technology development, which has characterised tape performance and developed key technologies and is now focused on demonstrating a high field tokamak magnet system at substantial scale.

In the next stage of development, these new technologies and understanding will be combined to deliver the world’s first device capable of fusion energy gain and industrial scale power production – ST-F1. ST-F1 is currently in the concept design stage and is aiming to demonstrate the viability of commercial fusion.

Speaker: Steven McNamara (Tokamak Energy Ltd)

09:35 Faster Fusion: ST40, engineering, commissioning, first results

ST40, engineering, commissioning, first results

Speaker: Mikhail Gryaznevich (Tokamak Energy Ltd)

10:10 Overview of recent progress on non-inductive start-up experiment in LATE

LATE (Low Aspect ratio Torus Experiment) is a small device with the toroidal magnetic field up to 0.16 T at R = 0.25 m. It has no center solenoid and the plasma current is initiated and maintained by ECH/ECCD alone. There are three launchers for 2.45 GHz microwave and one launcher for 5 GHz microwave, each of which is installed on the radial port and injects microwave in the left-handed circular polarization to excite electron Bernstein wave (EBW) via O-X-B process. Overdence ST plasmas with 6 ~ 7 times the plasma cutoff density and plasma current of ~ 10 kA are formed when the fundamental ECR layer is located near the plasma core and EBW is excited in the 1st frequency band.

When EBW at the 2nd frequency band of 5 GHz is excited in the ST plasma produced by EBW at the 1st frequency band of 2.45 GHz, plasma current is driven strongly while the bulk electron parameters such as density are nearly the same. It is suggested that the injected EBW at the 2nd frequency band is absorbed mainly by high energy tail electrons in the low field side at Doppler shifted ECR and drives the plasma current while EBW at the 1st frequency band heats the bulk electrons.

Intermittent events of plasma ejection through LCFS occur in the highly overdense plasma produced by 2.45 GHz microwave. The central density decreases about 20 % and strong magnetic activity appears during the events of ~ 100 μsec. Multiple magnetic probe signals show that a current channel escapes to the upper wall and TAE like oscillations are excited. Heavy ion beam probe measurement shows that space potential near the plasma core increases about 50V during the ejection event and recovers at once.

• This work is supported by JSPS KAKENHI Grant Numbers 18H03689, 18H01198 and 17K06992.

Speaker: Prof. Hitoshi Tanaka (Kyoto University)

10:45 Coffee Break

11:10 Overview of Versatile Experiment Spherical Torus (VEST): Progress and Plans

Y. S. Hwang and VEST team
Department of Nuclear Engineering, Seoul National University, Korea
yhwang@snu.ac.kr

Speaker: Prof. Yong-Seok Hwang (Seoul National University)

11:45 Scaling Physics of Reconnection Heating for ST Merging Startup Experiments

The high-power reconnection heating of merging tokamak plasma has been developed mainly in TS-3, TS-4 and MAST experiments. This unique method is caused by the promising scaling of ion heating energy that increases with squire of reconnecting magnetic field Brec. We studied mechanisms for this scaling of reconnection (ion) heating up to 2.3keV mainly using TS-3 and TS-6 experiments and PIC simulations and found the following features:
(i) the ion heating energy is as high as ~40-50% of poloidal magnetic energy of two merging tokamak plasmas and, (ii) the ion heating energy is not affected by (guide) toroidal field Bt, in the region of Bt/ Brec >1, under the two conditions: (a) compression of the current sheet to the order of ion gyroradius and (b) isolation of the merging tokamak plasmas from coils, electrodes and walls. The sheet compression to the order of ion gyroradius was found to be a key condition to realize the fast reconnection as well as the high power ion heating consistent with the Brec2-scaling prediction. Under this condition, the ion heating energy is determined uniquely by Brec ~ Bp not by Bt in the conventional tokamak operation region: Bt/ Brec >2 (or q0>2). The merging tokamak plasmas need to be fully pinched off from the PF coils without any link with coils, electrodes or walls for the purpose of minimizing loss of the hot ions heated by the reconnection/ merging. Most of the laboratory experiments of magnetic reconnection tends to have ion and electron temperatures as low as 5-30 eV due to large energy loss through magnetic fluxes intersecting coils, fluxcores, electrodes and walls. They are sacrificing the plasma confinement for the better controllability of magnetic reconnection. The ion heating was found to occur not only in the local downstream area of reconnection but also in the global merging tokamak area, increasing ratio of the ion heating energy to the electron heating energy of reconnection around 4. This promising scaling realized ion temperature as high as 2.3keV in 2019 and is expected to realize the burning ion temperature >10keV (under electron density ne~1.5x1019 [m-3]) by increasing Brec (~poloidal magnetic field Bp) over 0.6T, leading us to the high-Brec field merging tokamak experiments: ST-40 in Tokamak Energy Inc. and TS-6 in U. Tokyo.
[1] Y. Ono et al., Nuclear Fusion 59, 076025 (7pp) , (2019)

Speaker: Yasushi Ono (University of Tokyo)

12:20 Investigation of ion heating/transport process of magnetic reconnection during CS-free merging plasma startup of spherical tokamak in TS-6

Ion heating/transport process of CS-free merging plasma startup through magnetic reconnection has been investigated in the TS-3U (TS-6) spherical tokamak using ultra-high resolution 96CH/320CH 2D ion Doppler tomography diagnostics. In addition to the previously reported high-temperature plasma startup up to $\sim$250eV in TS-3 and $\sim$1.2keV in MAST, this research focused on the detailed characteristics of high guide field reconnection ($B_t > 3B_{rec}$) motivated by MAST merging/compression experiment which demonstrates promising performance for the connection to a quasi-steady scenario and pioneers a new frontier for reconnection studies: fine structure formation by reconnection heating. By identifying the double-axis field configuration with the X-point on the midplane using in situ magnetic probe diagnostics, the detailed measurement successfully revealed that the ion temperature profile forms two types of characteristic heating structure, both around the X-point and downstream. The former is affected by the Hall effect to form a tilted heating profile, while the latter is affected by the transport process which a forms a poloidal double-ring-like structure. The achieved ion heating mostly depends on the reconnecting component of the magnetic field, and the contribution of the guide field to decrease the heating efficiency tends to be saturated in the high guide field regime. Under the influence of better toroidal confinement with higher guide field, the downstream ion heating is transported vertically, mostly by parallel heat conduction, and finally forms a poloidal ring-like hollow distribution aligned with the closed flux surface at the end of merging.
[1] H. Tanabe et.al., Phys. Rev. Lett. 115, 215004 (2015)
[2] H. Tanabe et al, Nucl. Fusion 59, 086041 (2019)

Speaker: Dr Hiroshi Tanabe (University of Tokyo)

htanabe_ISTW.pdf

12:40 Group Photo

13:00 → 14:00 Lunch

14:00 → 17:00 Poster session: Poster Session 1 and coffee break

Convener: Franco Alladio (ENEA-FSN-Fusphy)

14:00 Posters set up

18:00 → 19:00 IEA ST Technology Collaboration Programme - Executive Committee Meeting (No. 13)

Wednesday, 30 October

08:00 → 09:00 Registration

09:00 → 12:45 Session 04

Convener: Brian Lloyd (UKAEA)

09:00 NSTX-U: Recent Results and Plans

NSTX-U: Recent Results and Plans
S.M. Kaye
Princeton Plasma Physics Laboratory, Princeton NJ 08543

After an 18-month sequence of key reviews, the NSTX-U Recovery Project was recently approved to initiate procurement and fabrication of components as a step towards completion of the Recovery and commencement of physics operations in the 2021-2022 time frame. The elements of the Recovery, which include new PF coils, new support structure including center stack casing, redesign of graphite tiles with improved heat flux handling capabilities, will insure highly reliable operations. The first few years of operation will focus on two primary goals: assessing ST performance at up to 6 times lower collisionality than could be achieved in NSTX, and development of fully non-inductively sustained plasmas for pulse lengths up to 5 s. The longer-term goal is the development of a full toroidal deployment of liquid lithium divertor modules to assess this transformative solution to handling the high heat fluxes expected in next-step devices. During the Recovery outage, research has targeted topics from NSTX/NSTX-U data or through collaborations on other devices that could facilitate the achievement of the NSTX-U research goals once operation starts. These topics include understanding mechanisms that control pedestal, core and fast ion transport and stability, and developing associated predictive models, developing real-time control capabilities and wall conditioning. This presentation will describe the recent progress in this research.

Speaker: Stanley Kaye (PPPL)

09:35 Validation of gyrokinetic simulations in NSTX including comparisons with a synthetic diagnostic for high-k scattering.

A new extensive validation study performed for a modest beta NSTX NBI-heated H-mode predicts that electron thermal transport can be entirely explained by short-wavelength electron-scale turbulence fluctuations driven by the electron temperature gradient mode (ETG), both in conditions of strong and weak ETG turbulence drive. For the first time, local, nonlinear gyrokinetic simulations carried out with the GYRO code [Candy JPP 2003] reproduce the experimental levels of electron thermal transport while simultaneously matching the frequency spectrum of electron-scale turbulence, the shape of the wavenumber spectrum and the ratio of fluctuation levels between strongly driven and weakly driven ETG turbulence conditions. Ion thermal transport is shown to be very close to neoclassical levels predicted by NEO [Belli PPCF 2008], consistent with stable ion-scale turbulence predicted by GYRO. Comparisons between high-k fluctuation measurements [Smith RSI 2008] and simulations are enabled via a novel synthetic high-k diagnostic developed for GYRO. The frequency spectra characteristics of electron-scale turbulence (spectral peak and width) can be reproduced by the synthetic spectra, but prove not to be critical constraints on the simulations. However, the shape of the high-k wavenumber spectrum and the fluctuation level ratio between the strong and weak ETG conditions can also be simultaneously matched by electron-scale simulations within sensitivity scans about the experimental profile values, and prove to be great discriminators of the simulations analyzed. Electron-scale simulations were also able to isolate the effect of safety factor and magnetic shear to match the shape of the measured fluctuation wavenumber spectrum. This work is the strongest experimental evidence to date that ETG-driven turbulence can dominate in the outer-core of modest beta NSTX H-modes.

This work has been supported by US DOE contracts DE-AC02-09CH11466 and DE-FG02-91ER54109. Computer simulations were carried out at NERSC, supported by the Office of Science of the U.S. DOE under Contract No. DE-AC02-05CH11231, and at the MIT-PSFC partition of the Engaging cluster at the MGHPCC facility (www.mghpcc.org), which was funded by DOE grant number DE-FG02-91-ER54109.

Speaker: Dr Juan Ruiz Ruiz (1. MIT-Plasma Science and Fusion Center 2. Rudolf Peierls Centre for Theoretical Physics, University of Oxford)

10:10 3D HHFW full wave simulations with realistic antenna geometry and SOL plasma in NSTX-U

In this paper we report an application of the Petra-M on NSTX-U. Petra-M, is a recent developed and open source code, which is based on the scalable MFEM C++ finite element library and allows for FEM analysis from geometry/mesh generation, FEM assembly, FEM system equation solution, and visualization in one platform [1, 2]. The first full torus 3D high harmonic fast wave (HHFW) simulations for NSTX-U plasmas including the scrape-off-layer (SOL) region with realistic antenna geometry and core plasma will be presented. A scan of the antenna phasing is performed showing a strong interaction between FWs and the SOL plasma for lower antenna phasing, which is consistent with previous NSTX HHFW observations. The antenna spectrum for different antenna phasing will be also shown. A first attempt to couple the 3D RF solver with the full-orbit following particle code SPIRAL [3] will be discussed with the aim to show the impact of the effect of the 3D wave field on the fast ion population from NBI beams in NSTX-U.

Work supported by U.S. DOE Contracts DE-SC00108090 and DE-AC02-09CH11466

[1] S. Shiraiwa et al., EPJ Web of Conferences 157 (2017) 03048.
[2] S. Shiraiwa et al., Nucl. Fusion, in preparation (2019).
[3] G. J. Kramer et al., Plasma Phys. Control. Fusion 55 (2013) 025013.

Speaker: Nicola Bertelli

10:45 Coffee Break

11:15 Machine learning approaches to predict the ideal stability properties of spherical tokamak plasmas

Machine Learning approaches to predict the ideal stability properties of spherical tokamak plasmas

A.Piccione 1, J.W. Berkery 2, S.A. Sabbagh 2, Y. Andreopoulos 1

1 Department of Electronic and Electrical Engineering, University College London, London, WC1E 7JE, UK

2 Department of Applied Physics and Applied Mathematics, Columbia University, New York, NY 10027, USA

Lead author email: a.piccione@ucl.ac.uk

One of the biggest challenges to achieve the goal of producing fusion energy in tokamak devices is avoidance, or mitigation, of disruptions of the plasma current due to instabilities. In order to analyse these disruptions, the Disruption Event Characterization and Forecasting (DECAF) framework has been developed [1], integrating physics models of several causal events that can lead to a disruption. Two different machine learning approaches are proposed to improve the ideal magnetohydrodynamic (MHD) no-wall limit component of the full kinetic stability model included in DECAF.
First, a powerful and partially interpretable machine learning algorithm, the Random Forest Regressor [2], was adopted to reproduce the DCON [3] computed change in plasma potential energy without wall effects, $\delta W_{no-wall}^{n=1}$. When trained on a large database of equilibria from the National Spherical Torus Experiment (NSTX), the Random Forest can significantly improve the prediction performance as well as the classification of stable/unstable points. Furthermore, this tree-based method provides an analysis of the contribution of each input feature, showing that the plasma parameters that most affect the estimated value of $\delta W_{no-wall}^{n=1}$ are the ones expected by the underlying physics. Secondly, a multilayer perceptron neural network has been trained on sets of calculations with the DCON code, to get an improved closed form equation of the no-wall limit as a function of the relevant plasma parameters provided by the Random Forest. Although being slightly worse than the Random Forest in classification performance, this approach can directly provide an estimated value of $\beta_{N,no-wall}^{n=1}$, which is key to determine the mode growth rate inside the overall kinetic stability model [4]. The model has been incorporated into DECAF and tested against a set of experimentally stable and unstable discharges. The portability of the model is also investigated, showing initial encouraging results by testing the NSTX-trained algorithm on the Mega Ampere Spherical Tokamak (MAST).

The authors acknowledge the help of S.P. Gerhardt, M.D. Boyer and R.J. Akers. This research was supported by US DOE Grant DE-SC0018623 and EPSRC Grants EP/R025290/1, EP/P02243X/1.

[1] S.A Sabbagh et al., to be submitted to Physics of Plasmas (2019)
[2] L. Breiman, Machine Learning, 45, 5–32 (2001)
[3] A.H. Glasser, Physics of Plasmas 23, 072505 (2016)
[4] J.W. Berkery et al., Physics of Plasmas 24, 056103 (2017)

Speakers: Mr A. Piccione (University College London) , Mr J.W. Berkery (Columbia University) , Mr S.A. Sabbagh (Columbia University)

11:35 Transient CHI start-up of low inductance plasma for advanced ST scenarios

Demonstration of fully solenoid-free start-up at the ~1 MA level and ramp-up to full current of low inductance plasma would be a major step for the ST program, as it would be a convincing enough step for the consideration of ST designs with a reduced size solenoid. In support of this objective, the TSC code, with the inclusion of plasma transport, has been used to study the Transient CHI closed flux current start-up potential in a BT =1 T NSTX-U device and at increased BT =3 T, to assess the potential capability on ST-40, as the ST-40 vessel dimensions are comparable to those of NSTX-U.

In reactors based on the ST/AT concepts, minimizing the recirculation power fraction is desired for an attractive and economical electrical power generation system. Such devices must operate at high fractions of the bootstrap current drive. The ST configuration with a low aspect ratio is particularly attractive in this regard. In addition, to sustain the high-performance regime, density and pressure profile control should be demonstrated using reactor relevant core fueling systems that during each fueling pulse does not significantly perturb the optimized profiles necessary for maintaining the bootstrap current drive.

The immediate advantage of solenoid-free start-up is that it provides more space to add structural material to the toroidal field coil to support a higher peak field and higher field on axis – which could greatly increase fusion performance and/or reduce the cost of the facility. Reducing the size of the central solenoid would allow the valuable in-board region to be used for other more critically important systems that are necessary for sustained steady-state operation. Some of these new capabilities are highspeed pellet injection from the inboard side (without the use of a guide tube with strong curvature) and high field side RF launchers for heating and current drive.

During transient CHI start-up, as well as during the merging-compression start-up pioneered on the START and MAST devices, all the start-up current is produced during the initial plasma formation pulse, and one does not continue to drive current after the plasma has formed. Consequently, the scaling of these concepts to larger devices is relatively well understood as a reliable predictive model is difficult to develop as the plasma response during the current ramp-up is not well known, especially at high current.

Transient CHI TSC simulations at 1 T show that closed flux start-up currents >500 kA should be achievable with the present divertor coil set on NSTX-U. NSTX-U is also well equipped with high power neutral beam injection capability that is optimized for sustained non-inductive current drive of the CHI target. Simulations also show that if the BT is increased to 3 T, and with increased divertor coil current rating, closed flux start-up currents more than 0.6 MA should be possible. Work is in progress to assess the requirements for 1 MA closed flux current start-up in a 3 T device. TSC simulations also show that at these high levels of magnetic flux injection the CHI plasma self-heats to 100s of eV electron temperature. Additional heating, if required, could be provided by ECH as the CHI plasma electron density is below the density limit for ECH systems considered for STs. The ST-40 device would be a particularly attractive device on which to demonstrate MA level current start-up followed by sustainment with reactor-relevant ECH.

Speaker: Dr Roger Raman

11:55 Configuration Studies for a Low-Aspect-Ratio Liquid-Metal-Wall Sustained High-Power-Density Tokamak Facility

Utilizing high-Z solid walls such as tungsten is challenging for next-step/Fusion Nuclear Science Facility (FNSF)/Pilot Plant applications due to a range of issues. These issues include Plasma-Facing Component (PFC) material damage from erosion and re-deposition and neutrons, high-Z impurity accumulation and associated core plasma radiative collapse, relatively low heat flux limits at the PFC, and thermal plasma pedestal energy confinement reduction. Liquid metal walls and divertors are increasingly being studied as a possible means of addressing these challenges. However, the impact of liquid metal systems on device configuration and core plasma performance at the Proof-of-Performance level (or higher) in a tokamak configuration has not yet been systematically investigated. In this work we explore possible configurations for a sustained high-power-density tokamak facility with lower aspect ratio (A = 1.8-2.5) dedicated to the study and development of a range of liquid metal divertor and first-wall concepts. Such a device would build upon past and expected results from liquid metal test-stands, the Lithium Tokamak Experiment (LTX), the National Spherical Torus Experiment Upgrade (NSTX-U), and the Experimental Advanced Superconducting Tokamak (EAST), but the device configuration is driven primarily by the needs (space, plumbing, thermal insulation, etc.) of liquid metal systems. Configuration studies build upon previous low-A High Temperature Superconductor (HTS) tokamak pilot plant studies that incorporated a liquid metal divertor for high-heat-flux mitigation and as a means of reducing poloidal field coil current and simplifying the magnet layout and maintenance scheme. Initial physics scenario and engineering configuration studies for a next-step liquid-metal wall and divertor toroidal confinement facility are described.

Speaker: Dr Jonathan Menard (Princeton Plasma Physics Laboratory)

12:45 → 14:00 Lunch