ISTW 2022

Book of Abstracts

 
Contents 
Analysing flat-top operational space of STEP reactor design ......................................................................... 4 
Analysis of Globus-M2 tokamak plasma heating experiments using diamagnetic measurements ................. 5 
Conceptual design of the Globus-3 spherical tokamak with a copper electromagnetic system ...................... 6 
Design of a new charge exchange spectroscopy system in VEST .................................................................. 7 
Development of a Projectile Probe .................................................................................................................. 8 
Development of a washer gun for doublet startup on the SUNIST-2 spherical tokamak ................................ 9 
Development of full electromagnetic plasma start-up model for analysis and prediction of ohmic discharges 
in Versatile Experiment Spherical Torus ....................................................................................................... 10 
Development of holistic whole-device predict-first modeling of NSTX-U .................................................. 11 
Development of hybrid multi-level inverters for the PF coils of the SUNIST-2 spherical tokamak ............. 12 
Disruption processes and their characteristics on the SUNIST spherical tokamak ....................................... 13 
Disruption runaway electrons generation and mitigation attempts in STEP preferred concept .................... 14 
Electron cyclotron plasma ramp-up through effective acceleration of energetic electrons and bulk electron 
heating in QUEST ......................................................................................................................................... 15 
Electron heating during neutral beam injection at Globus-M2 ..................................................................... 16 
Electron Bernstein waves heating experiments in SUNIST .......................................................................... 17 
ENN Roadmap for Proton-Boron Fusion Energy Based on Spherical Torus ................................................ 18 
Equilibrium design for the SUNIST-2 spherical tokamak ............................................................................. 19 
Equilibrium operational space and stability limits in early operation of MAST-U ....................................... 20 
Framework for predictive modeling of neoclassical tearing mode activity in NSTX-U ............................... 21 
Free Boundary Equilibrium Modelling of Disruption Scenarios for STEP .................................................. 22 
Global gyrokinetic study of shaping effects on electromagnetic modes at NSTX aspect ratio ..................... 23 
High temperature plasmas in ST40 ............................................................................................................... 24 
High-fidelity 3D HHFW full wave simulations in NSTX-U......................................................................... 25 
High-Power Reconnection Heating for Formation of High-Beta Spherical Tokamak with Absolute 
Minimum-B in TS-3,4,6 ST merging experiments ........................................................................................ 26 
Improvement of plasma diamagnetic flux measurement system in VEST .................................................... 27 
Improvement of vacuum and wall conditioning systems in Versatile Experiment Spherical Torus .............. 28 
Improvements to the High-Field-Side Transient CHI System on QUEST .................................................... 29 
Initial Simulations of Negative Triangularity Plasmas on MAST-U ............................................................. 30 
Interaction between External Blobs and Internal MHD Modes During Internal Reconnection Event in 
VEST. ............................................................................................................................................................ 31 
Investigation of EBW current drive in open magnetic flux surface using flux loop with HFS injection in 
QUEST .......................................................................................................................................................... 32 
Investigation on anisotropic temperature in Ohmic plasmas in TST-2.......................................................... 33 
Linear Instabilities in a ST40 Hot Ion Plasma............................................................................................... 34 
Magnetic reconnection: From compact fusion to plasma propulsion ............................................................ 35 
Mechanical and electromagnetic design of the vacuum vessel of the SMART tokamak .............................. 36 
Modelling of the STEP ramp-down phase with JETTO ................................................................................ 37 

 
Modification of energy conversion process in magnetic reconnection during merging formation of spherical 
tokamak ......................................................................................................................................................... 38 
Morphology: PROTO-SPHERA experiment vs. the inner wind region of the Crab Nebula supernova 
remnant .......................................................................................................................................................... 39 
Nonlinear simulation of pressure-driven MHD modes and beta-limit study in MTF ................................... 40 
Observation of whistler wave instability driven by temperature anisotropy of energetic electrons in plasmas 
of EXL-50 spherical tours ............................................................................................................................. 41 
Overview of ECRH and EBW modelling on ST40 ....................................................................................... 42 
Overview of EXL-50 Research Progress and Future Plan ............................................................................ 43 
Overview of First Exhaust Physics Results from MAST-U .......................................................................... 44 
Overview of Low Aspect ratio Torus Experiment (LATE): Recent Results and Plans ................................. 45 
Overview of Physics Results from the First Campaign of Experiments on MAST Upgrade ........................ 46 
Overview of the SMall Aspect Ratio Tokamak - SMART ............................................................................ 47 
Physics drivers of the STEP divertor concept design .................................................................................... 48 
Plasma Scenario Development in the Spherical Tokamak for Energy Production (STEP) ........................... 49 
Plasma shaping effects on the neoclassical tearing mode threshold physics ................................................. 50 
PPPL Low-Aspect-Ratio Next-Step Facility Scoping Studies ...................................................................... 51 
Pre-conceptual Design of Liquid Li Plasma Facing Components in NSTX-U ............................................. 52 
Progress of the SUNIST-2 spherical tokamak ............................................................................................... 53 
Properties of anomalous ion heating in lower hybrid wave sustained plasmas ............................................. 54 
Research in Versatile Experiment Spherical Torus (VEST) for the Development of Compact Fusion Reactor
 ....................................................................................................................................................................... 55 
ST Reactor – new considerations. ................................................................................................................. 56 
ST40: Advancing the Physics Basis of Spherical Tokamak Reactors ........................................................... 57 
Status of the PROTO-SPHERA experiment: evidences of plasma tori production and sustainment and 
progress towards Phase-2 .............................................................................................................................. 58 
Steady-State Alpha-Particle Flux on the Walls of the Spherical Tokamak for Energy Production ............... 59 
Studies of fast electron dynamics in rf start-up plasmas on the TST-2 spherical tokamak ........................... 60 
Systemic error analysis of Thomson scattering system on VEST with rotational Raman calibration .......... 61 
The New PEGASUS-III Experiment ............................................................................................................ 62 
The SMART Central Stack Design ............................................................................................................... 63 
Three-dimensional Magnetic Reconnections in PROTO-SPHERA .............................................................. 64 
Transport Analysis of ST40 High Performance Discharges .......................................................................... 65 
Twenty years of research on the SUNIST spherical tokamak ....................................................................... 66 
Verification of toroidal field coil stress analysis based on the measurements in experiments on tokamak 
GLOBUS-M2 ................................................................................................................................................ 67 
Will the MHD model coupled with the plasma diagnostics data be able to reconstruct out of reach core 
plasma parameters in Tokamak devices? ...................................................................................................... 68 
X-I mode for efficient solenoid-free ECCD plasma current start-up for fusion reactor ................................ 69 
Evolution of the D-T Spherical Torus (ST) toward the p-11B ST .................................................................. 70 
 

emmi.tholerus@ukaea.uk; francis.casson@ukaea.uk; thomas.wilson@ukaea.uk;

stephen.marsden@ukaea.uk; hendrik.meyer@ukaea.uk; simon.freethy@ukaea.uk

Analysing flat-top operational space of STEP reactor design

E. Tholerus, F.J. Casson, T. Wilson, S. Marsden, H. Meyer, S. Freethy, and STEP team

CCFE, UKAEA, Culham Science Centre, Abingdon OX14 3DB, UK

The STEP (Spherical Tokamak for Energy Production) is a planned experimental reactor

designed to demonstrate net electricity production. It will operate fully non-inductively, with

auxiliary heating and current drive provided by electron cyclotron (EC) and electron Bernstein

wave (EBW) systems. The target fusion power range Pfus is 1.5 – 1.8 GW, with scientific fusion

energy gain factor Qfus ≥ 11. The currently suggested design point has a 3.6 m major radius,

2.0 m minor radius, elongation κ = 2.8, plasma current Ip ~ 20 MA and toroidal magnetic field

Bt ~ 3.2 T.

Integrated core plasma modelling with JETTO [1] is utilised to explore the flat-top

operational space and the viability of the considered design. Considered key parameters are the

Greenwald density fraction, the normalised beta βN, and the auxiliary power. To confirm MHD

stability of the scenario, βN is compared against the ideal MHD no-wall β-limit, and the

minimum safety factor qmin is required to be above 2 to avoid destabilisation of n = 2 MHD

modes. The line averaged electron density is assumed below the Greenwald density limit to

prevent disruption events. Furthermore, the coil-set is designed to cater for a plasma current Ip

less than about 23 MA, and the exhaust system is expected to handle losses Psep/Rgeo <

40 MW/m. Simplified assumptions are made about the anomalous transport using rescaling of

the Bohm-gyroBohm transport coefficients with relative electron and ion average transport

supported by gyrokinetic modelling and other spherical tokamak experiments. The rescaling of

the transport is done to reach a set target βN. The viability of such assumptions is checked by

applying an upper limit on the confinement factor H of 1.2. Since the suggested operational

space has not been well explored experimentally, several different H-factor scaling laws are

analysed.

Since STEP will operate non-inductively, the main sources of plasma current are the

bootstrap current and auxiliary current drive. The bootstrap current is mostly efficient in high

density scenarios, whereas auxiliary current drive of the available systems is the highest when

operating at lower density. The EBW system has a generally higher expected current drive

efficiency than electron cyclotron current drive (ECCD), but its efficiency vanishes for ρtor less

than about 0.6. ECCD is consequently required for on-axis current drive. Extensive parameter

scans have been performed to identify the key sensitivities of this strongly nonlinear coupled

system. The scans suggest at least three different candidates for flat-top operation. Two of these

scenarios are in the high-density end close to Greenwald density limit, one with EC + EBW

heating and one with EC heating only. The third suggested scenario is in the lower density

regime at around 60 % Greenwald density fraction, using EC heating only.

References

[1] M. Romanelli et al., Plasma and Fusion Research 9, 3204023 (2014)

erina.tkachenko@yandex.ru

Analysis of Globus-M2 tokamak plasma heating experiments using

diamagnetic measurements

E.E. Tkachenko1,2, N.V. Sakharov1, G.S. Kurskiev1, N.S. Zhiltsov1, I.V. Miroshnikov1, Yu.V.

Petrov1, V.B. Minaev1, N.N. Bakharev1, A.A. Kavin3 , E.O. Kiselev1, K.M. Lobanov3, A.N.

Novokhatskii1, A.B. Mineev3, V.V. Solokha1, E.A. Tukhmeneva1, N.A. Khromov1.

1Ioffe Institute, St. Petersburg 194021, Russia

2Peter the Great St. Petersburg Polytechnic University, St. Petersburg, Russia

3JSC «NIIEFA», St. Petersburg, Russia

Globus-M2 is a spherical tokamak with the plasma major radius 0.36 m, the minor radius

0.22-0.23 m and the plasma vertical elongation κ ≤ 2. Discharges with the toroidal magnetic

field up to 0.9 T and the plasma current up to 0.4 MA were studied in regimes with pure ohmic

heating (OH) and regimes with auxiliary heating using neutral beam injection (NBI). Two

injectors were used, with the particle energies E1NBI = 30 keV and E2NBI = 45 keV and the total

output power PNBI =1.25 MW in atomic beams. Taking into account the OH power the total

heating power density achieved the value of Ptotal / Vp ≈ 3 MW/m3.

The evolution of plasma thermal energy WP was estimated using the free boundary

equilibrium code PET [1]. The code solves the Grad–Shafranov equation and also calculates

the magnetic flux measured with the diamagnetic loop. Irrespectively the last closed magnetic

surface was reconstructed by the method of movable filaments [2]. In this method the plasma

current is modeled by a set of 9 current-carrying rings. The input data are the magnetic fluxes

measured by 21 loops placed on the vacuum vessel wall near the plasma boundary, the plasma

current and the currents in PF coils and central solenoid. The signals from the magnetic loops

were also used to determine the distribution of the currents over the vessel wall. In PET

simulation the shapes of plasma pressure and current density profiles, as well as plasma beta (β)

were varied in such a way that the calculated with PET the last closed magnetic surface and the

diamagnetic flux coincided with the experimental values. Such approach made it possible to

determine the plasma thermal energy as well as the normalized internal inductance.

The calculated by PET values of plasma energy was compared with WP values obtained

from kinetic measurements of the plasma temperature and density profiles by TS diagnostic,

CXRS and NPA. The values of effective plasma charge Zeff were obtained from the intensity of

bremsstrahlung.

References

[1] S.A. Galkin, A.A. Ivanov, S.Yu. Medvedev and Yu.Yu. Poshekhonov, Nucl. Fusion, 37

(1997).

[2] V.I. Vasiliev, Yu.A. Kostsov, K.M. Lobanov, L.P. Makarova, A.B. Mineev, V.K. Gusev,

R.G. Levin, Yu.V. Petrov and N.V. Sakharov, Nucl. Fusion 46, 625 (2006).

Vladimir.Minaev@mail.ioffe.ru

Conceptual design of the Globus-3 spherical tokamak with a copper

electromagnetic system

V.B. Minaev1, A.B. Mineev2, N.N. Bakharev1, A.V. Bondar2, E.N. Bondarchuk2,

O.G. Filatov2, V.K. Gusev1, A.A. Kavin2, E.O. Kiselev1, A.Yu. Konin2, A.M. Kudriavtseva2,

G.S. Kurskiev1, Yu.V. Petrov1, I.Yu. Rodin2, N.V. Sakharov1, V.N. Tanchuk2, A.Yu. Теlnova1,

P.B. Shchegolev1, A.A. Voronova2, E.G. Zhilin3, N.S. Zhiltsov1

1Ioffe Institute, St. Petersburg, Russia

2JSC «NIIEFA», St. Petersburg, Russia

3Ioffe Fusion Technology Ltd., St. Petersburg, Russia

The main scientific task of Globus-M launched in 1999 was to study plasma confinement

in spherical tokamak. The poor confinement of high-energy ions during auxiliary plasma

heating was the main reason for the upgrade of the facility aimed at increasing the magnetic

field, which finished with the launch of the Globus-M2 tokamak in 2018. The next step in the

development of the compact spherical tokamak concept (project Globus-3) is a device in which

the duration of the plasma discharge exceeds the characteristic time of stationary plasma

profiles formation.

The main features of the Globus-3 facility are a long pulse, a high toroidal magnetic field, and

powerful auxiliary plasma heating, which make it possible to consider it as a hydrogen prototype

of a neutron source. The methodology for choosing the parameters of the Globus-3 tokamak in the

presence of physical, engineering and infrastructural restrictions is given.

For the parametric analysis, the GLOBSYS engineering-physical code was developed.

Within the framework of the zero-dimensional approximation, the definitions of the main

scalings of plasma parameters (



P, fBS, fCD, E), as well as the specifics of calculating the

inductance and resistance of plasma in spherical tokamaks, were refined. The code was verified

based on the experimental and predictive data of the Globus-M2, NSTX, NSTX-U, MAST,

MAST-U, and ST40 tokamaks. A preliminary selection of the Globus-3 parameters was carried

out for cases with a “warm” copper magnets, as well as a pre-cooled to liquid nitrogen

temperature. As a basis for further development of the project, it is proposed to accept following

geometric parameters: R0 ~ 0.6 - 0.7 m, a ~ 0.35 - 0.4 m, A  1.7 - 1.8, k95 ~ 1.7 - 1.8. According

to the calculations, the condition for reaching a long pulse of the current plateau can be met

with a toroidal field of 1.5 - 2 T. In the version with "warm" coils, the duration of the plasma

current plateau is 3 s and can be prolonged up to 12 - 13 s in case of pre-cooling. An increase

in the toroidal magnetic field and the transverse size of the plasma improves the conditions for

applying the auxiliary heating and non-inductive current drive. At a plasma current of 0.8 MA,

direct losses during neutral beam injection do not exceed 15% for the energies of 50-150 keV.

Preliminary modeling of the plasma parameters is performed using ASTRA and NUBEAM

codes.

Design of a new charge exchange spectroscopy system in VEST

TaeHee EOM1, Jae Young Jang1,2, C. Sung3, Soo-Ghee Oh4 and Y.S. Hwang1

1SNU, Korea, 2KFE, Korea, 3KAIST, Korea, 4Ajou University, Korea

An ion doppler spectroscopy diagnostic system is installed in Versatile Experiment

Spherical Torus (VEST) to measure ion rotation velocity and temperature. Because a signal of

C III impurity line (464.74nm) from plasma produced within the VEST is a line-integrated

signal, an inversion process is required for this passive IDS system so local measurements of

ion rotation velocity and temperature are difficult. Charge exchange spectroscopy (CES) does

not require an inversion process for profile measurements, because a CES system is composed

of a neutral beam and a spectroscopy system which are crossing each other so local

measurements of ion properties are possible. A combined system of He CES and H beam

emission spectroscopy (BES) was designed in VEST by adopting the concept of the ITER CES

system sharing the most of optical components with the BES to minimize uncertainties of the

neutral beam density along beam path. [2] The new combined CES & BES system will observe

C VI line (529.05nm) and H line (656.3nm) but it measures C III line (464.7nm) and He II line

(468.6nm) before the measurable conditions are met. This work presents the design of a the

new CES system.

Acknowledgements

This work was supported by National R&D Program through the National Research

Foundation of Korea(NRF) funded by Ministry of Science and ICT(NRF-

2021M3F7A1084418)

References

[1] K. J. Chung et al., "Design features and commissioning of the versatile experiment

spherical torus (VEST) at Seoul National University." Plasma Science and Technology

15.3 (2013): 244.

[2] YooSung Kim et al., "Design of combined system of helium charge exchange

spectroscopy and hydrogen beam emission spectroscopy in VEST." Fusion Engineering

and Design 123 (2017): 975-978.

Development of a Projectile Probe

Tingzhi Chang1, Yu Zhou1,2, Binbin Wang1, Shouzhi Wang1, Wenbin Liu1, Yi Tan1 and Zhe

Gao1

1Department of Engineering Physics, Tsinghua University, Beijing, 100084, China

2Southwestern Institute of Physics, Chengdu, 610225, China

Probe, including magnetic probe and Langmiur probe, is extensively used in tokamak

plasma diagnostics. However, probe insertion, especially probe holder, will bring a severe

disturbance to plasma nearby, which limit the probe usage in the boundary of plasma. We now

developed a probe free of holder, named projectile probe, aiming to reduce the influence of

probe insertion and extend its diagnostic range. An electromagnetic launcher based on the

mechanism of coil gun is developed to launched the probe from bottom before discharge.

During discharge, the probe is quasi-static in plasma while acquiring signals, and after

discharge, it is recycled by the launcher preparing for next launch. The launcher is able to send

the probe to the core of plasma (~40cm from bottom limiter) with a location accuracy of ±1cm.

The probe is compactly shaped in Φ30mm x 50mm, integrating a coil head, a MCU capable of

wireless communication, a data acquisition system, an accelerometer and a battery inside. The

entire device has been installed and tested on SUNIST (Sino-UNIted Spherical Tokamak).

Signal from core plasma is obtained and radial motion of MHD activities is observed comparing

signals from poloidal magnetic probe array.

Acknowledgements

This work is supported by NSFC under Grant Nos.11827810 and 11875177 and IAEA

CRP under Research Contract No. 22733 and the Science Challenge Project, No. TZ2016005

Development of a washer gun for doublet startup on the SUNIST-2

spherical tokamak

XIN Tao,1 CHENG Zhengbo,1 and TAN Yi1

1Department of Engineering Physics, Tsinghua University, Beijing, 100084, China

A set of washer plasma guns is developed for the SUNIST-2 spherical tokamak. The

doublet startup progress on the SUNIST-2 ST can be optimized by generating initial plasma

using a set of washer plasma guns. Thus, the first current rings can be pushed up to reach the

potential of the internal coils. The washer plasma gun has been tested off the tokamak device

and on the SUNIST device. In a 3-ms discharge, the plasma current generated by the washer

gun can reach 5-kA off the tokamak device. The electron temperature is at 5-eV level, and the

electron density can reach 1019 m-3. And the impurity level is under control. Also, the washer

gun has been tested on the SUNIST tokamak, and it drives 4-kA plasma current with 1-kA

injected current. The washer gun developed meets the requirements of plasma initialization on

the SUNIST-2 device. Washer gun configuration on the SUNIST-2 device is under design and

further experiments stay to be carried out.

Acknowledgements

This work is supported by NSFC under Grant Nos.11827810 and 11875177 and IAEA

CRP under Research Contract No. 22733 and the Science Challenge Project, No. TZ2016005

yshwang@snu.ac.kr

Development of full electromagnetic plasma start-up model for analysis and

prediction of ohmic discharges in Versatile Experiment Spherical Torus

Seongcheol Kim1, Hongsik Yun1, Hyun-Tae Kim2, Jae Young Jang1,3, Jung-Hwa Kim1,

TaeHee Eom1 and Y. S. Hwang1

1SNU, Korea

2UKAEA, United Kingdom

3KFE, Korea

Plasma start-up model has been developed in VEST [1] by utilizing a burn-through

simulation model, DYON, with a dedicated atomic model [2]. For the model development,

start-up is divided into two phases: burn-through and current ramp-up. Townsend breakdown

model combined with electromagnetic analyses by utilizing finite element method has been

introduced to estimate the plasma volume at the end of the breakdown and burn-through phase.

For the validation of the synthetic plasma volume, fast camera images and triple probe data

have been utilized. At the current ramp-up phase, plasma current evolution is mainly determined

by the electric circuit equation. Plasma circuit parameters such as external inductance have been

improved for the applicability of the model from low aspect ratio plasma (ST) to high aspect

ratio plasma based on the analytic formula derived by Hirshman & Neilson [3]. In order to

confirm the start-up model, several ohmic discharges with different operating conditions in

VEST have been reproduced in comparison with experimental data such as plasma current,

averaged electron temperature and density, averaged ion temperature and hydrogen alpha line

emission signal. Synthetic results with the start-up model show reasonable level of agreements

with experimental data. Following the validation of the modeling against the VEST discharges,

an ohmic discharge with VEST-U will be predicted.

Acknowledgements

This work was supported by National R&D Program through the National Research

Foundation of Korea(NRF) funded by Ministry of Science and ICT(NRF-

2021M3F7A1084418)

References

[1] K. J. CHUNG, Y. H. AN, B. K. JUNG, H. Y. LEE, C. SUNG, Y. S. NA, T. S. HAHM

and Y. S. HWANG, Plasma Science and Technology, 15, 3 (2013).

[2] Hyun-Tae Kim, W. Fundamenski, A.C.C. Sips and EFDA-JET Contributors, Nuclear

Fusion, 52, 103016 (2012).

[3] S. P. Hirshman and G. H. Neilson, Physics of Fluids, 29, 790 (1986).

Development of holistic whole-device predict-first modeling of NSTX-U

K. E. Thome1, W. DeShazer1, D. Eldon1, X. Jian1, M. Knolker1, J. McClenaghan1, O.

Meneghini1,T. Odstrcil1, T. Osborne1, A. Pajares1, R. I. Pinsker1, S. P. Smith1, B. Van

Compernolle1, D. Weisberg1, W. P. Wehner1, A. Welander1, G. Avdeeva2, W. Guttenfelder3, S.

Kaye3

1General Atomics, 2ORAU, 3PPPL

Offline integrated predict-first modeling utilizing previous NSTX and NSTX-U

experimental data is underway to provide predictions and recommendations for NSTX-U’s next

campaign. The goal of this effort is to inform development of a stationary, non-inductive

scenario on NSTX-U with high toroidal beta and high bootstrap fraction utilizing its higher

toroidal field and improved heating and current drive tools. This modeling has three

components: 1) integrated transport modeling, 2) heating and current drive modeling focused

on the use of the high harmonic fast wave (HHFW) antenna, and 3) whole-device simserver

simulations focused on optimizing the plasma rampup. All three tasks are described in detail

below.

The OMFIT integrated modeling workflows kineticEFITtime and STEP, which are

routinely utilized on DIII-D, have been adapted and deployed for public use for both NSTX and

NSTX-U. The kineticEFITtime workflow supports both single-time and multi-time modeling

and is incorporated with both the ONETWO and TRANSP codes. NSTX and NSTX-U

adaptations to this workflow (mostly driven by ST effects compared to conventional tokamaks)

allow the use of isothermal and rotation constraints for EFIT, the correction of the density

asymmetry due to the centrifugal effect, and treatment of the Thomson scattering density and

temperature separately. Many NSTX and NSTX-U shots have been tested in this workflow, and

some have been analyzed in detail with both TGLF and TGYRO. Future work will continue to

adapt and improve this workflow, perform integrated transport analysis over a wider array of

NSTX and NSTX-U discharges, broaden efforts to the edge, and perform predictions for the

next NSTX- U campaign. The NSTX HHFW antenna is now included in the STEP workflow

and has been verified by reproducing previously published NSTX results. The operational space

of the HHFW antenna on NSTX-U is currently being explored using GENRAY and will be

further refined in the future with CQL3D and TORIC.

The NSTX-U control flight simulator GSevolve coupled with the NSTX-U Plasma Control

System (PCS) has been resurrected. The coupling simulates the entire NSTX-U discharge from

breakdown to termination with high resolution current profile evolution in good agreement with

TRANSP simulations. NSTX-U discharges with vertical displacement events have also been

successfully simulated. A simulation mode for poorly controlled NSTX-U experimental

discharges has been created that replaces the NSTX-U PCS and can provide insight into control

instabilities. A numerical optimal control model coupled with GSevolve is currently under

development and has been used to design feedforward shape control. This work could address

previous NSTX-U control issues and assist in the design of a robust NSTX-U rampup with low

ℓI, high kappa, and high non-inductive fraction. Future near-term work will continue to improve

the GSevolve model, incorporate TRANSP into GSevolve, and add an HHFW model.

Acknowledgements

Given that significant progress has been made individually on these three interrelated

NSTX and NSTX-U predict-first modeling tasks, future work will be able to focus on connecting

them more tightly along with incorporating further improvements to improve their physics

fidelity. Work supported by U.S. DOE under DE-SC0021113 and DE-AC02-09CH11466.

Development of hybrid multi-level inverters for the PF coils of the SUNIST-

2 spherical tokamak

Zichong SONG1, Yi TAN1, Shouzhi WANG1, Wenbin LIU2,3

1 Department of Engineering physics, Tsinghua University, Beijing 100084, China

2 College of Physics and Optoelectronic Engineering, Shenzhen University, Shenzhen 518060,

China

3 Advanced Energy Research Center, Shenzhen University, Shenzhen 518060, China

The SUNIST-2 spherical tokamak will test multiple plasma configurations including

droplet, doublets, D-shape and negative triangularity. These configurations require fast and

accurate magnetic control of plasma positions, shaping, vertical displacement events, etc. Three

pairs of PF coils are used to control the plasmas, of which typical electrical parameters are:

resistantce ~10 mΩ, inductance ~500 μH. According to estimations, in order to meet the

requirements above, the power supplies (PSs) of the poloidal field (PF) coils should reach the

following goals: Irated ~10 kA; max |dI/dt| ~1 kA/ms; Tresponse <0.5 ms; Tdischarge ~1 s. The

expected current waveform in PF coils is DC output with steep ramp-up and fast modulation.

Also, current and voltage ripple must be minimized.

The PSs of PFs use capacitors as energy storage. Two types of capacitors are used in the

PSs: aluminum electrolyte capacitors (AEC, higher Vmax and Imax, but lower energy density)

and super capacitors (SC, lower Vmax and Imax, but higher energy density). A hybrid multi-level

inverter PS is designed and constructed to make benefit of both AECs and SCs with compact

size and financial acceptance. In this PS, two cascaded H-bridges are connected serially and

driven by 2 different capacitor banks. The SC bank is charged to ~100 V and providing the

majority energy for 10kA discharge during flat-top. While the AEC banks are charged to ~400

V and provide required dI/dt. The output voltage levels of one H-bridge are ±Vin, 0. Then the

total output of this hybrid PS can reach 7 output voltage levels (Vhigh-Vlow and –(Vhigh-Vlow) are

forbidden to avoid mass energy flow from AECs to SCs). Hence, less ripple than conventional

multi-level PS with similar structure. Besides, the energy in coils can be recycled when dI/dt<0.

A MCU based controller and suitable PWM driving algorithm is developed too. The computer

simulation and test circuit experiment result show that this power supply is able to control the

current as predefined waveforms while reducing voltage ripple noise of coil at the same time.

Acknowledgements

This work is supported by NSFC under Grant Nos.11827810 and 11875177 and IAEA

CRP under Research Contract No. 22733 and the Science Challenge Project, No. TZ2016005

*spj18@mails.tsinghua.edu.cn

Disruption processes and their characteristics on the SUNIST spherical

tokamak

Pengjuan Su1*, Binbin Wang1, Zhengbo Cheng1, Menghua Yang1, Shouzhi Wang1, Wenbin

Liu2, Yi Tan1, Long Zeng1, Zhe Gao1

1Department of Engineering Physics, Tsinghua University, Beijing 100084, China

2College of Physics and Optoelectronic Engineering, Shenzhen University, Shenzhen 518060,

China

Disruption event is a major concern in tokamak including spherical tokamak researches.

Methods of avoiding or mitigating the hazards of disruption, such as runaway electrons, are not

yet well developed. The disruption process and its features on the SUNIST tokamak has been

investigated and the potential approaches to dissipative runaway electrons are being validated.

The disruption events on SUNIST usually happen when low toroidal field and low electron

density (   ) discharge condition. The disruptions here include both minor and

major disruption, and most of which are minor. Several significant features of the disruption

process are as follows: 1) the loop voltage increases, 2) the HXR initially increases considerably

and disappears abruptly after the appearance of the magnetic perturbation, 3) the density on

some channels rapidly decrease, while another channels rapidly increase initially and then

decrease again, in what appears to be density ejectment process. The discharge length of

SUNIST is about 20ms. The minor disruption can occur at the pre- (< 4ms), mid- (4~14ms) and

post-term (>14ms) of a discharge pulse, which occur at different time show differences on

magnetic perturbation, HXR, runaway electron evolution. A high frequency magnetic

perturbation, which is identified as Alfvén instability, can be found on mirnov probes,

microwave interferometer and AXUV diagnostics during disruptions. The Alfvén instability is

in the range of 150 kHz~1MHz and lasts about 0.1ms. The Alfvén instability tends to have two

modes, one of which is lower frequency and the other is higher frequency. The toroidal number

is mainly 1 ~ 2 and they propagate in the electron diamagnetic drift direction. Benefiting from

the high density resolution and high frequency resolution of the five-channel microwave

interferometer, density fluctuations caused by the Alfvén instability can be captured by the

interferometer system clearly. The different density response on the multi-channel

interferometer chords have been analysed to help to investigate the spatial location and mode

structure of the mode. In addition to, the radiation fluctuations caused by the Alfvén instability

also can be measured by the AXUV array, and the mode structure is under analysis.

A comprehensive investigation of disruption events, Alfvén instabilities and runaway

electrons has been carried out. The cause of disruptions on SUNIST is thought to be plasma

displacement or impurities. The motivation of the Alfvén instability is regarded to runaway

electrons accelerated by the high loop voltage when disruption occurs. More details will be

studied in the future works.

Disruption runaway electrons generation and mitigation attempts in STEP

preferred concept

A.Fil1, T. Hender1, S. Newton1, M. Hoppe2, E. Berger3, T. Fülöp3, I. Pusztai3, O. Valhagen3

1United Kingdom Atomic Energy Authority, Culham Centre for Fusion Energy, Culham

Science Centre, Abingdon, Oxon, OX14 3DB, UK

2Swiss Plasma Center, EPFL, Lausanne, CH-1015, Switzerland

3Department of Physics, Chalmers University of Technology, Göteborg, SE-41296, Sweden

Generation of Runaway Electrons during plasma disruptions and their potential impact on

the plasma facing components is of great concern for ITER and future reactors based on the

tokamak concept. The STEP (Spherical Tokamak for Energy Production) programme aims at

producing net energy from a prototype fusion energy plant. At its current operating point, the

plasma current is higher than 20 MA and it is expected to be in the seed-insensitive regime of

avalanche multiplication, i.e., any runaway seed would quickly generate a large runaway beam

during unmitigated disruption current quenches (as in ITER and SPARC). This is confirmed by

modelling runaway electron generation in an unmitigated disruption current quench for the

STEP preferred concept flat top operating point, using the state-of-the-art code DREAM [1].

Hot-tail generation of runaways is found to be the dominant primary generation mechanism,

and the avalanche multiplication factor is confirmed to be extremely high. Even by varying

assumptions for the prescribed thermal quench phase (duration, final electron temperature) in a

reasonable physical range, but also the wall time, the plasma-wall distance, and shaping effects,

all STEP unmitigated disruptions generate more than 10 MA of runaway current (up to full con-

version). Mitigation using idealized impurity injection of Argon and D2 has then been modelled,

and so far no scenario allows to avoid runaways while respecting the other constraints of

disruption mitigation (current quench time less than 150ms, and thermal energy radiated

fraction above 90%). This motivates further investigations, using DREAM SPI model, including

a better model for MHD losses during disruptions (i.e., DREAM+JOREK) and studying and

optimizing the mitigation of runaway electrons beams with the D2+MHD scheme, as was proven

successful in JET [2], ASDEX Upgrade and DIII-D. Such runaway mitigation D2 SPI(s) are

planned for ITER, and included in the STEP design. Other additional methods, such as a passive

runaway electrons mitigation coil (REMC, as planned for SPARC) will also be investigated.

References

[1] M. Hoppe, et al., 2021, Computer Physics Communications 268, 108098.

[2] C. Reux, et al., 2021, Physical Review Letters 126, 175001

idei@triam.kyushu-u.ac.jp

Electron cyclotron plasma ramp-up through effective acceleration of

energetic electrons and bulk electron heating in QUEST

H. Idei1, T. Onchi1, A. Ejiri2, T. Kariya3, A. Fukuyama4, M. Ono5, K. Kono1, T. Ido1,

T. I. Tsujimura6, S. Kubo6, R. Ikezoe1, K. Hanada1, M. Hasegawa1, K. Kuroda1, F. Zennifa1,

A. Higashijima1, T. Nagata1, I. Sekiya1, S. Shimabukuro1, S. Murakami4

1Research Institute for Applied Mechanics, Kyushu University, Kasuga, 816-8580, Japan

2Department of Complexity Science and Engineering, The University of Tokyo, 277-8561,

Kashiwa, Japan

3Plasma Research Center, University of Tsukuba, 305-8577, Tsukuba, Japan

4Department of Nuclear Engineering, Kyoto Univ., Kyoto, 615-8540, Japan

5Princeton Plasma Physics Laboratory, Princeton, NJ, 08543, USA

6National Institute for Fusion Science, Toki, 509-5292, Japan

The Q-shu University experiment with steady-state spherical tokamak (QUEST) project

was proposed at Kyushu University to achieve and to investigate steady-state operation in the

spherical tokamak (ST) configuration under controlled plasma-wall interactions. In STs, since

the centre solenoid (CS) fluxes for Ohmic tokamak discharges are inherently limited, electron

cyclotron plasma ramp-up experiments have been conducted to supplement either none or

limited CS fluxes. A high plasma current Ip of 54 kA was attained and was stably sustained

for 0.9 s with possible simplest 28-GHz transmission and launcher systems [1]. The higher Ip

of 66 kA was also achieved with a slow ramp-up of the vertical field. In addition, the dual-

frequency experiments with 8.2-GHz and 28-GHz waves were explored for electron Bernstein

wave heating in the over-dense plasmas [2]. A focusing mirror and polarizer systems were

developed to conduct local 28-GHz electron cyclotron heating and current drive (ECHCD)

experiments. A higher Ip of 80 kA was attained through effective acceleration of energetic

electrons in higher harmonic relativistic Doppler-shifted resonances in the oblique injection [3].

In the quasi-perpendicular injection, relatively high electron temperature of 500 eV was

achieved at 1.5 x 1018 m-3 near the 2nd harmonic resonance layer. In the oblique, a hard X-ray

radiation counts was remarkably large in the energy range of < ~200 keV, whereas the X-ray

count significantly decreased in the quasi-perpendicular injection. The heating effect for highly

energetic and bulk electrons are discussed involving resonant velocity regions in both

injections, based on resonant ellipses or quasi-diffusion terms in the momentum space.

References

[1] H. Idei et al., Nucl. Fusion 57, 126045 (2017).

[2] H. Idei et al., Nucl. Fusion 60, 016030 (2020).

[3] T. Onchi et al., Physics of Plasmas 28, 022505 (2021).

Electron heating during neutral beam injection at Globus-M2

Zhiltsov N.S.1, Kurskiev G.S.1, Bakharev N.N.1, Balachenkov I.M.1, Filippov S.V.1,

Gusev V.K.1, Khromov N.A.1, Kiselev E.O.1, Koval A.N.1, Krikunov S.V.1, Minaev V.B.1,

Miroshnikov I.V.1, Mukhin E.E.1, Novokhatsky A.N.1, Patrov M.I.1, Petrov Yu.V.1,

Sakharov N.V.1, Shchegolev P.B.1, Shulyatiev K.D.1, Skrekel O.M.1, Solokha V.V. 1,

Solovey V.A.2, Telnova A.Yu.1, Tkachenko E.E.1, Tokarev V.A.1, Tolstyakov S.Yu.1,

Tukhmeneva E.A.1, Varfolomeev V.I.1, Voronin A.V.1, Goryainov V.Yu.1, Yashin A.Yu. 1, 3

1 Ioffe Institute, Saint-Petersburg, Russia

2 Petersburg Nuclear Physics Institute named by B.P. Konstantinov of NRC, “Kurchatov

Institute”, Saint-Petersburg, Russia

3 Peter the Great St. Petersburg Polytechnic University, St. Petersburg, Russia

The presentation shows the study of electron heating in the Globus-M2. Globus-M2 is the

compact spherical tokamak (R/a = 0.36 m/0.24 m = 1.5) designed for operation with the toroidal

magnetic field BT up to 1 T. The experiments were performed in a lower single-null divertor

plasma configuration with moderate plasma elongation k ≈ 1.6-1.9. The plasma heating

efficiency was investigated in the range of plasma current and magnetic field: Ip = 0.25-0.4 MA,

BT = 0.7-0.9 T. The auxiliary plasma heating was applied by two neutral beams (NBI) with

particle energy ENBI-1 = 28 keV and ENBI-2 = 45 keV with the total power PNBI ≤1.25 MW in

atomic beam.

The main research tool was the Thomson scattering diagnostics, recently upgraded at the

Globus-M2. The diagnostics provided measurement of the spatial distribution of the electron

temperature Te and density ne in the equatorial plane. The local Te and ne measurements

combined with equilibrium reconstruction data made it possible to estimate electron stored

energy We.

The average electron density scan <ne> with constant Ip, BT, PNBI showed that the shape

of the Te profile did not depend on <ne> in the range of 2.5-7.5×1019 m-3. This resulted in linear

dependence of We on <ne> in this range. The use of auxiliary heating made it possible to double

the stored electron energy at <ne> = 7×1019 m-3 compared to pure ohmic heating. Furthermore,

the presentation also shows dependency of We on plasma parameters, measured in BT and Ip

scans with constant PNBI and <ne>. Finally, the transport analysis carried out with ASTRA

transport code allowed to estimate electron heat diffusivity for regimes with different plasma

current and density.

The experiments were carried out at the Unique Scientific Facility “Spherical Tokamak

Globus-M”, which is incorporated in the Federal Joint Research Center “Materials Science and

Diagnostics in Advanced Technologies”.

Electron Bernstein waves heating experiments in SUNIST

Wang Shouzhi1, Wang Binbin, Liu Wenbin, Cheng Zhengbo, Su Pengjuan, Yang Menghua,

Tan Yi, Gao Zhe

1Tsinghua Universiy, Beijing, China

For dense plasma, electron cyclotron wave heating is not available, and electron Bernstein

waves generation by mode conversion is required to overcome the accessibility problem and

heat the core dense plasma. In this paper, electron Bernstein waves heating experiments on the

SUNIST device are presented. Using a microwave source of 2.45 GHz/30 kW and based on the

high density gradient formed spontaneously at the SUNIST boundary, this paper successfully

converts about 60% of the microwave energy into elecron Bernstein waves by means of the FX-

B mode conversion, which is in agreement with the conversion efficiency given by the

numerical calculation model. By scanning the microwave injection conditions and discharge

parameters, this paper finds that the electron Bernstein waves deposit energy at the plasma

boundary through the boundary parametric instability and boundary energetic electron damping.

The impurity ions gain energy from the low-frequency daughter wave of the parametric process

and transport inward to the plasma core, resulting in the growth of electron density and radiation

intensity in the core, and the electron Bernstein waves do not directly heat the core plasma. The

fraction of energetic electrons increases after energy absorption, leading to a decrease in the

temperature of the electrons measured by the triple probe and an enhancement of hard X-rays

in the late stages of the discharge.

xiehuasheng@enn.cn, liuminsheng@enn.cn

ENN Roadmap for Proton-Boron Fusion Energy Based on Spherical Torus

Huasheng Xie, Yuanming Yang and Minsheng Liu*, on behalf of ENN fusion team

Hebei Key Laboratory of Compact Fusion, Langfang 065001, China

ENN Science and Technology Development Co., Ltd, Langfang 065001, China

ENN is committed to generating fusion energy in an environmental-friendly and cost-

effective manner, which requires abundant aneutronic fuel. Proton-boron (p-11B) fusion is

considered as an ideal choice. Studies have indicated that the p-11B fusion, though still difficult,

is possible, based on the new cross section data [1]——if the electron radiation can be reduced

by achieving hot ion mode and high wall reflection. The high beta and compactness of spherical

torus makes it suitable for p-11B fusion. A reactor with major radius R0=4m, central magnetic

field B0=6T, central temperature Ti0=150keV, plasma current Ip=30MA and hot ion mode

Ti/Te=4 can yield p-11B fusion Q>10, as the new spherical torus energy confinement scaling

law is used [2]. A roadmap of p-11B fusion is created with the next generation device named as

EHL-2 (EHL stands for ENN He-Long, literally means “peaceful Chinese dragon”). The main

target parameters include R01.05m, A1.8, B03T, Ti0=30keV, Ip=3MA, and Ti/Te=2. The

existing fusion device EXL-50 [3] will be upgraded simultaneously to provide experimental

support for the new roadmap, which involves central solenoid, vacuum chamber and magnetic

systems installation and upgrading. Commissioning of the upgraded fusion device EXL-50U is

scheduled for 2024, and the construction of EHL-2 is estimated to conclude by 2026.

References

[1] M. H. Sikora and H. R. Weller, J Fusion Energ (2016) 35:538–543.

[2] G.S. Kurskiev et al 2022, Nucl. Fusion 62 016011.

[3] Y. J. Shi et al 2022, Nucl. Fusion 62 086047.

Equilibrium design for the SUNIST-2 spherical tokamak

Zhengbo Cheng,1 Zhe Gao,1 and Yi Tan,1

1Department of Engineering Physics, Tsinghua University, Beijing, 100084, China

A new Free Boundary equilibrium (FBE) code is developed based on EFIT [1] algorithm

and has been used to evaluate the coil currents needed to sustain a variety of magnetic

configurations for the SUNIST-2 device. FBE can be run in the equilibrium mode and the

reconstruction mode. With the equilibrium mode, the prespecified plasma boundary is semi-

fixed by heavy fitting weights of boundary points and then required active coils’ currents can

be obtained. In contrast, by increasing that of prespecified coils’ currents, the corresponding

plasma configuration can be computed. With the reconstruction mode, the magnetic equilibrium

can be reconstructed from external diagnostic signals. Our code adopts the polynomial model

to represent the plasma current profile with a few physical parameters. This way allows the

constraint of global magnetohydrodynamic parameters including diamagnetic flux and axial

safety factor. Besides, the number of plasma columns can be prespecified to allow the

computation of arbitrarily shaped equilibria, with external or internal separatrices and multiple

magnetic axes. Then limiter, double null with positive and negative triangularity, droplet, and

doublet configurations are computed with FBE in the equilibrium mode, and synthetic magnetic

signals are used to reconstruct the equilibrium with FBE in the reconstruction mode. And the

reconstructed equilibrium agrees well with the designed ones.

Acknowledgements

This work is supported by NSFC under Grant Nos.11827810 and 11875177 and IAEA

CRP under Research Contract No. 22733 and the Science Challenge Project, No. TZ2016005

References

[1] Lang Lao, Reconstruction of current profile parameters and plasma shapes in tokamaks,

nuclear fusion, 25, 1611(1985)

Equilibrium operational space and stability limits in early operation of

MAST-U

J. Berkery

Princeton Plasma Physics Laboratory

The MAST-U spherical tokamak has recently completed its first campaign of physics

operation. MAST-U has operated with Ohmic, or one or two neutral beams for heating, at 400-

800 kA plasma current, in conventional or “SuperX” divertor configurations. Equilibrium

reconstructions with magnetics only or additionally with motional Stark effect diagnostic

measurement of the magnetic pitch angle provide key plasma physics parameters vs. time for

each discharge. The disruption event characterization and forecasting (DECAF) code produces

diagrams which show where the prevalence of operation occurred as well as the limits in various

operational spaces. When compared to stability limits, the operation of MAST-U so far has

generally stayed out of low q, low density instability regions, and below the high density

Greenwald limit, high beta global stability limits, and high elongation vertical stability limit.

MAST-U has reached βN/li =3.3 so far, well below the ideal no-wall limit, which was projected

to be about βN/li = 7 [1]. MAST-U still has the potential to reach higher elongation, which

could benefit the plasma performance.

Acknowledgements

This manuscript is supported by U.S. Department of Energy grants DE-SC0018623 and

DE-AC02-09CH11466.

References

[1] Berkery J W et al 2020 Plasma Phys. Control. Fusion 62 085007.

Framework for predictive modeling of neoclassical tearing mode activity in

NSTX-U

J. Yang1, M. Podestà1, E. Fredrickson1, F. Poli1

1Princeton Plasma Physics Laboratory

A TRANSP [1] based framework for the predictive modeling of neoclassical tearing mode

(NTM) activity in tokamaks is developed and tested for NSTX discharges. Since it is natural to

expect that burning plasma devices will have NTMs, this framework can offer a good tool for

the design and execution of the NTM control in burning plasma devices. In this framework, a

measured NTM island width is input along with the equilibrium and profiles to produce a self-

consistent set of thermal and fast ion profiles to be input to the modified Rutherford equation

(MRE) [2]. As an interpretive analysis to validate the framework, a reasonable match of NTM

island width time evolution calculated from the MRE is achieved using the least squares fit of

free parameters that represent measurement uncertainties of input [3]. A recently developed

theory of fast ion contribution to the NTM island growth [4] is found to be essential for the

analysis, and the Kick model [5] implemented in TRANSP provides a vital link between the

fast ions and the NTM. For design purposes, a user can modify any of the input parameters and

re-run the simulation to predict the response of the NTM island. For example, a beam modulation

period can be modified to design the scenario with either larger NTM island for diagnostic

purposes or smaller NTM island for better performance. A demonstrative analysis of NSTX

discharges will be presented, along with the detailed description of the framework.

Acknowledgements

This manuscript is based upon work supported by the U.S. Department of Energy, Office

of Science, and Office of Fusion Energy Sciences, and has been authored by Princeton

University under Contract Number DEAC02-09CH11466 with the U.S. Department of Energy.

References

[1] Poli et al., TRANSP (Computer Software) https://doi.org/10.11578/dc.20180627.4

[2] Fredrickson et al., Phys. Plasmas 7 4112 (2000).

[3] Yang et al., Plasma Phys. Control. Fusion 64 095005 (2022).

[4] Cai, Nucl. Fusion 56 126016 (2016).

[5] Podestà et al., Plasma Phys. Control. Fusion 56 055003 (2014).

Sam.bakes@ukaea.uk, oliver.bardsley@ukaea.uk, francis.casson@ukaea.uk,tim.hender@ukaea.uk

Free Boundary Equilibrium Modelling of Disruption Scenarios for STEP

S. Bakes1, O. Bardsley1, F. Casson1, T. Hender1

1United Kingdom Atomic Energy Authority, Culham Centre for Fusion Energy, Culham

Science Centre, Abingdon, Oxon, OX14 3DB

Reactor scale devices such as STEP will require robust systems for the mitigation and

control of disruptions. Two such examples of relevance to spherical devices are vertical

displacement events (VDEs) as well as sudden changes in poloidal β and internal inductance

(li) through, for example, a loss of confinement. The high elongation operating regime of STEP

makes VDEs a significant risk, and the large amount of stored energy could lead to significant

damage to plasma facing components. In the case of unintended changes in parameters, where

the radial position of the plasma drifts inwards due to a drop in pressure leading to a

corresponding drop in the hoop force, the natural design consequences of a spherical device

introduce significant challenge. The space for the centre column is extremely limited and the

first wall is likely to be very close to the plasma on the inboard side. Thus, any control system

needs to react quickly and with significant magnitude to prevent the plasma from striking the

wall. As with VDEs the plasma would hold significant thermal energy, the relatively small wall

area on the inboard side would lead to higher heat fluxes, along with less space for sacrificial

limiters and the complication that maintenance on the inboard side is more challenging,

meaning that repairs are more expensive.

In order to predict strike points and inform the design of control solutions, the quasi-static

free boundary equilibrium code FEEQS1 has been employed using representative plasma

profiles produced by the JETTO transport code. In the case of VDEs a mode shape has been

calculated using the RZIP linear equilibrium response model2 and applied as an initial condition,

from here the plasma subsequently continues to drift until making contact with a limiter. The

results have been used to tailor the shape of the first wall and limiter placements. In another set

of simulations, an H-mode baseline scenario was taken and a JETTO simulation was run with

the transport coefficients fixed but enhanced edge confinement instantaneously removed. This

resulted in a rapid change in li and βp (~4% and 3% in 0.1s) followed by a more gradual decline

over the duration of 1 second. The outputs of this simulation can then be used in FEEQS to

model the motion of the plasma and feed-forward control scenarios have been investigated,

along with the influence of radially coupled passive structures.

References

[1] H. Heumann et al, J. Plasma Physics 81, (2015)

[2] A. Coutlis et al, Nucl. Fusion 39, (1999)

*asharma@pppl.gov

Global gyrokinetic study of shaping effects on electromagnetic modes at

NSTX aspect ratio

A. Y. Sharma,1* M. D. J. Cole,1 T.

Go¨ rler,

Y. Chen,3 D. R. Hatch,4W. Guttenfelder,1 R.

Hager,1 B. J. Sturdevant,1 S. Ku,1 A. Mishchenko,5C. S. Chang1

1Princeton Plasma Physics Laboratory, Princeton, USA;

2Max Planck Institutefor Plasma Physics, Garching, Germany;

3University of Colorado, Boulder, USA;

4University of Texas, Austin, USA;

5Max Planck Institute for Plasma Physics, Greifswald, Germany;

Linear electromagnetic simulations are performed at NSTX aspect ratio using the global

gyrokinetic particle code XGC. Various magnetic geometries are simulated that are based on

NSTX discharge 132588, which is an enhanced pedestal (EP) H-mode discharge with high

elongation and triangularity at high plasma β. Profile-consistent magnetic equilibria are used,

and approximate compressional magnetic perturbation effects are included by modifying the

particle drifts. High elongation and triangularity are found to cause significant stabilization of

collisionless microtearing modes and kinetic ballooning modes. This study shows that a highly

shaped spherical tokamak could possess an attractive confinement feature when considering

microturbulence-driven transport.

*otto.asunta@tokamakenergy.co.uk

High temperature plasmas in ST40

O. Asunta*, S. A. M. McNamara, P. R. Thomas, J. Bland, P. Buxton, C. Colgan, A.

Dnestrovskij, M. Gemmell, M. Gryaznevich, D. Hoffman, F. Janky, H. Lowe, R. Mirfayzi, G.

Naylor, V. Nemytov, J. Njau, T. Pyragius, A. Rengle, M. Romanelli, C. Romero, M. Sertoli, V.

Shevchenko, J. Sinha, A. Sladkomedova, S. Sridhar, Y. Takase, J. Varje, B. Vincent, H.

Willett, J. Wood, D. Zakhar1,

D. Battaglia, S.M. Kaye, L. Delgado-Aparicio, R. Maingi, D. Mueller, M. Podesta2,

E. Delabie, B. Lomanowski3,

O. Marchuk4

1 Tokamak Energy Ltd, Milton Park, Oxfordshire OX14 4SD, UK

2 Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543, USA

3 Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831, USA

4 Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung –

Plasmaphysik, Partner of the Trilateral Euregio Cluster (TEC), 52425 Jülich, Germany

ST40 is a high field Spherical Tokamak (ST), built and operated by Tokamak Energy Ltd

(TE), with Rgeo≈0.4-0.5m, A≈1.6-1.9, κ<1.7, IP≈0.4-0.8MA and BT≈0.8-2.2T. Two co-current

hydrogen neutral beams deliver 0.8MW at 50kV and 0.7MW at 24kV. For deuterium injection,

the corresponding figures are 1.0MW/55kV and 0.8MW/24kV, respectively. For plasma start-

up, ST40 uses a process called merging-compression. The vessel walls are conditioned with

helium GDC and boronisation. The plasmas were mainly limited on the high field side, although

some diverted plasmas were obtained. The primary goal for the 2021-2022 operations was the

attainment of the business objective: Ti(0) > 100MK (8.6keV). A central deuterium ion

temperature of 9.6keV was obtained together with i(0).(0).E ≈ 6 ± 2 × 1018 m−3.keV.s. Such

high ion temperatures have previously been achieved only in significantly larger devices and

never in a ST. To access the high Ti regime, the electron temperature must be high enough that

the collisional power transfer from ions to electrons is significantly smaller than the input power

density. In the best ST40 shots, Te(0) > 3keV was observed, which is in contrast to the Te(0)

<2.4keV in neutral beam heated STs of twice the major radius but with BT <0.8T. The results

from ST40 are consistent with the NSTX observation of an approximately linear BT dependence

of energy confinement time. A limited TF scan and comparisons between high Ti data at 1.6T

and 1.9T, on axis, demonstrate the beneficial effect of high BT. The dominant loss channel is

via electrons and so the observation of high temperatures lends support to the high field ST

route to economical, compact fusion reactors.

Acknowledgements

The Diagnostic Advisory Board, the TE management team, and TE’s investors.

Contributions from US collaborators were funded by a public-private CRADA between

Tokamak Energy Ltd, ORNL and PPPL.

*nbertell@pppl.gov

High-fidelity 3D HHFW full wave simulations in NSTX-U

N. Bertelli1*, S. Shiraiwa1, and M. Ono1

1 Princeton Plasma Physics Laboratory, Princeton, NJ, USA

NSTX-U is equipped with a 6 MW high harmonic fast wave (HHFW) system and 10 MW

neutral beam injection (NBI) for heating and current drive. The HHFW system on NSTX

demonstrated the capability to heat the plasma to temperatures larger than 6 keV [1]. However,

both HHFW experiments and modeling studies in NSTX showed (i) a strong interaction

between scrape-off layer (SOL) plasma and HHFW leading to RF power losses;

(ii) a possible strong interaction between HHFW and NBI. For these reasons, a high-

fidelity 3D modeling for HHFW including the coupling core-edge plasma is crucial to better

understand the interaction between HHFW and SOL plasma and NBI. In this paper we report a

significant advancement of HHFW modeling in NSTX-U simulating the full torus 3D SOL

plasma together with a realistic HHFW antenna and core plasma by making use of Petra-M,

which is an integrated finite-element-method (FEM) open-source multi-physics platform [2,3].

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. A magnetic field and edge density scans are also performed and discussed. The

HHFW-SOL interaction tends to increase for lower magnetic fields and higher edge density. In

particular, the cavity modes are clearly found in the SOL region in agreement with previous 2D

AORSA simulations [4]. This result bodes well for the higher toroidal field operations expected

in NSTX-U. Examples of S-matrix for both vacuum and plasma cases are also obtained paving

the necessary next step to calculate the coupling resistance of the HHFW system for different

plasma scenarios. Finally, an application of the coupling between the full wave TORIC code to

model the core plasma including the kinetic effects and Petra-M to model the SOL plasma is

discussed considering the RF rectified potential [4] to be able to predict the RF impurity

generation and more accurate HHFW power absorption in the core plasma.

Acknowledgements

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

References

[1] G. Taylor et al., Phys. Plasmas 17 (2010) 056114.

[2] S. Shiraiwa et al., EPJ Web of Conferences 157 (2017) 03048.

[3] S. Shiraiwa, 28th IAEA-FEC Conference 2020.

[4] N. Bertelli et al., Nucl. Fusion 54 (2014) 083004.

[5] S. Shiraiwa et al., submitted to Nucl. Fusion (August 2022).

High-Power Reconnection Heating for Formation of High-Beta Spherical

Tokamak with Absolute Minimum-B in TS-3,4,6 ST merging experiments

Y. Ono1, H. Tanabe1, M. Akimitsu2, R. Hiriuchi3, S. Usami3

1 Department of Advanced Energy, University of Tokyo, Japan

2 National Institutes for Quantum and Radiological Science and Technology, Japan

3National Institute for Fusion Science, Japan

The high-power reconnection heating of merging spherical tokamak (ST) plasmas has been

developed by TS-3, 4, 6, UTST, MAST and ST-40 experiments1-4 and PIC simulations in NIFS5-

7 for the future direct access to the burning plasmas4. The ST merging/reconnection produces

huge downstream heating of plasma ions whose energy is as high as ~40% of the reconnecting

(poloidal) magnetic energy, forming an interesting high-beta ST plasmas with hollow current

and reversed-shear profiles4. This high-beta ST often has an absolute minimum-B profile in the

second stability regime for ballooning instabilities. We studied a series of scaling law of the

reconnection/ merging heating4,7: the ion temperature/ ion heating energy scaling proportional

to the square of reconnecting (poloidal) magnetic field Brec and the scaling of electrostatic

potential wells and hills in the tilted downstream whose peak-to- peak value is proportional

both to Brec and toroidal (guide) magnetic field Bt. We observed these scaling laws only if we

trigger the fast reconnection by compressing the current sheet to the order of ion gyrodradius.

The ion heating energy is not or weekly affected by guide (toroidal) field Bt, in the tokamak

operation regime: q>2, because the multi-blob-formation, merging, and ejection cause

intermittent ion heating especially in the high Bt regime, weakening Bt dependence of ion

heating.

References

[1] Y. Ono et al., Physical Review Letters 107, 185001 (2011).

[2] Y. Ono et al., Nuclear Fusion 59, 076025 (2019) .

[3] H. Tanabe et al., Physical Review Letters 115, 215004 (2015).

[4] Y. Ono et al., Fusion Energy 2020, IAC/P4-3, (2021).

[5] R. Horiuchi et al., Phys. Plasmas 26, 092101 (2019).

[6] S. Usami et al., Phys. Plasmas 26, 102103 (2019).

[7] C. Z. Cheng, et al., Phys. Plasmas 28, 072101 (2021) .

Lead author email: tkkim99@snu.ac.kr

Corresponding author email: yhwang@snu.ac.kr

Improvement of plasma diamagnetic flux measurement system in VEST

Taekyoung Kim1, Seongcheol Kim1, Jung-Hwa Kim1, TaeHee EOM1, and Y.S. Hwang1

1Seoul National University, Korea

Plasma diamagnetic flux is an important parameter to measure the stored energy of

tokamak plasma. Toroidal field (TF) coil current measurement can be a simple and accurate

diamagnetic flux measurement method. The time-varying plasma diamagnetic flux induces the

change of TF coil current. This method has been used in Versatile Experiment Spherical Torus

(VEST) by utilizing Hall sensor that monitors total TF coil current.[1] However, the sensitivity

of the sensor becomes too low to measure the current change induced by diamagnetic flux

with reduced sensor gain after doubling TF field strength. So, a Rogowski coil with large

sensitivity for changing current rather than total current is designed with larger cross-sectional

area and more turns. An analog integrator with bypass and reset switches is included to avoid

saturation of integrated signals. The switches control integration time to integrate the signal

only during the plasma discharge instead of whole TF coil current change. Using virial analysis

and plasma equilibrium reconstruction result, the plasma stored energy is obtained from the

measured diamagnetic flux. [2] The kinetic measurements from ion Doppler and Thomson

scattering spectroscopy are used to validate the device and the stored energy data. This work

presents the improvement of the signal level and its validation with kinetic measurements.

Acknowledgements

This research was supported by National R&D Program through the National Research

Foundation of Korea(NRF) funded by Ministry of Science and ICT(NRF-2021M3F7A1084418)

References

[1] J. Yang, Y. Kim, W. I. Jeong, and Y. S. Hwang, Rev. Sci. Instrum, 89, 103508 (2018).

[2] M. W. Bongard, J. L. Barr, R. J. Fonck, J. A. Reusch, and K. E. Thome, Phys. Plasmas, 23,

072508 (2016)

wjeong1227@snu.ac.kr

Improvement of vacuum and wall conditioning systems in Versatile Experiment

Spherical Torus

Won Ik Jeong1, June Young Kim1, and Y. S. Hwang1

1Department of Nuclear Engineering, Seoul National University, Seoul, Korea

Vacuum and wall conditioning systems have been improved for better plasma performance

in Versatile Experiment Spherical Torus (VEST) at Seoul National University. Old vacuum

components such as O-ring are replaced and water vapor pumping speed is increased with

additional CRYO panel. As an additional wall baking system, halogen lamp is newly installed

in the VEST to effectively reduce the oxygen impurity influx from center-stack. The

boronization system is modified to avoid carborane () recondensation during helium

GDC, making boron-containing thin films deposition more uniform and widely spread. With

these improvements, chamber base pressure reached below    Torr, leading higher

plasma currents of up to 0.18 MA and longer discharge lengths of up to 42ms. However, VEST

still needs better control of wall recycling and gas feeding. To monitor neutral gas pressure with

fast time response in-situ neutral pressure gauge has been developed, installed and used with

other diagnostics such as filterscope. As a result, impurity influx as well as recycled hydrogen

fuel are analyzed for various plasma facing components in VEST. A new bakeable, movable

carbon limiter is installed to investigate the possibility of removing water vapor from outer

limiter and vacuum vessel by replacing entire outboard tungsten limiter. Although present He-

boronization system is found to be effective in removing oxygen influxes than typical GDC,

hydrogen concentration is increased significantly due to dissociation of carborane. New wall

condition techniques such as deuterium boronization and in-situ titanium gettering will be

tested for achieving low recycling regime in VEST.

Improvements to the High-Field-Side Transient CHI System on QUEST

Kengoh KURODA1, Roger RAMAN2, Makoto HASEGAWA1, Takumi ONCHI1,

Kazuaki HANADA1, Masayuki ONO3, John ROGERS2, Ryuya IKEZOE1, Hiroshi IDEI1,

Takeshi IDO1, Osamu MITARAI4, Masayoshi NAGATA5, Nobuhiro NISHINO6,

Shoji KAWASAKI1, Takahiro NAGATA1, Aki HIGASHIJIMA1, Shun SHIMABUKURO1,

Ichiro NIIYA1, Izumi SEKIYA1, Yuya OTSUKA1, Kaori KONO1, Kazuo NAKAMURA1,

Akira EJIRI7, Sadayoshi MURAKAMI8

1Kyushu Univ., 2UW, 3PPPL, 4Institute for Advanced Fusion and Physics Education,

5Univ. of Hyogo, 6Bauhutte, 7Univ. of Tokyo, 8Kyoto Univ.

Results from Transient Coaxial Helicity Injection (transient CHI) current start-up in

QUEST show that the CHI generated toroidal current scales in proportion to the injected

poloidal flux. For a given amount of injector poloidal flux, the required injector current scales

inversely with the toroidal field. These are very encouraging scaling relations for the application

of transient CHI to future high-field STs such as the ST40. These results were obtained using a

simplified version of the single biased electrode configuration applied from the high-field-side

on QUEST. Successful transient CHI discharge evolution up to a toroidal current of 43 kA was

observed after the CHI gas injection system was modified to include a gas manifold that

permitted the main CHI discharge to initiate in the injector region instead of elsewhere in the

vessel. These promising results motivated us to significant modify the lower divertor structure

on QUEST to permit the installation of a more permanent and much improved high-field-side

CHI system, as shown in Fig. 1. The main features are the proximity of the injector to the PF5-

1 injector flux coil that allows the injector flux to be increased by a factor of three. The CHI

gas injection system was further modified to a higher-conductance configuration. These

improvements should enable over 100 kA of closed flux current generation on QUEST to be

used as targets for subsequent heating and current drive using ECH.

Figure 1 Divertor electrode configuration for T-CHI discharge in QUEST.

Initial Simulations of Negative Triangularity Plasmas on MAST-U

A.O. Nelson1, S. Freiberger1, H. Wilson1, L. Rondini1, C. Paz-Soldan1

1 Applied Physics and Applied Mathematics, Columbia University, New York, New York, USA

On moderate aspect ratio machines, negative triangularity (NT) has recently emerged as an

attractive alternative configuration capable of accessing high beta while maintaining an L-

mode-like edge. In this work, we present results from initial simulations extending NT to

spherical tokamaks, with the eventual goal of realization on the MAST-U machine. Several new

equilibrium shapes with varying triangularities compatible with the MAST-U coil set are

generated for study. These equilibria are assessed for H-mode inhibition by infinite-n ballooning

modes [1]. A self-consistent core configuration is reached through iteration with the STEP code,

which uses TGYRO to predict the stationary density, temperature, and rotation profiles by

balancing the particle, heating, and torque sources against predicted turbulent and neoclassical

fluxes. The resulting Neg-D equilibria show an enhanced good-curvature trapped particle

fraction compared to positive triangularity counterparts, which leads to improved confinement.

This effect is further enhanced by reducing the plasma elongation and moving towards “comet-

shaped” equilibria with oblate cross-sections, which could be a particularly interesting regime

for spherical tokamaks. Rough estimates of the vertical growth rates for these NT plasmas on

MAST-U suggest that they should be controllable with existing control schemes. These initial

modeling results lay the groundwork for the development of NT spherical tokamak experiments

on the MAST-U machine, which may establish access to new configurations with strong

confinement, high beta and alleviated power handling constraints.

References

[1] A.O. Nelson et al 2022 Nucl. Fusion 62 096020

*corresponding author

Interaction between External Blobs and Internal MHD Modes During

Internal Reconnection Event in VEST.

E.C. Jung1,2, Y.S. Na1, S. Kim1, W.I. Jeong1, J.-K. Park2, T.S. Hahm1, Y.-c. Ghim3, and Y.S.

Hwang1*

1Department of Nuclear Engineering, Seoul National University, Seoul 08826, Korea,

Republic Of

2Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543, USA

3Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and

Technology, Daejeon 34141, Korea, Republic Of

Scape-off-layer (SOL) plasmas, which determine boundary conditions of the core plasma,

significantly affect the physics inside the core plasma. Although they are physically

disconnected from the core plasma, they can considerably alter the edge profiles and transport

and often carries a significant open-field-line current that can impact the core plasma. One of the

phenomena largely affected by SOL plasmas is explosive MHD instabilities. Various studies on

edge localized modes and disruptions show the critical effects of SOL plasmas on their onsets

[1,2]. However, the detailed physics of the strong relationship between SOL plasmas and MHD

instabilities has not been fully understood.

In this research, we present experimental observations that blobs, a field-aligned mesoscale

plasma in the SOL region [3], closely interact with internal MHD modes and trigger internal

reconnection events, one of the global MHD explosive events frequently observed in the

spherical torus. Various techniques to systematically interpret visible camera images to

investigate the blobs and internal MHD instabilities are conducted in the versatile experimental

spherical torus (VEST). The results show a strong correlation between SOL blobs and internal

MHD modes, and these correlated structures become synchronized in their rotations slightly

before the onset. An IRE without such synchronization, however, has a less violent impact in

terms of current loss and the strength of magnetic bursting. These results imply that SOL plasma

can interact with internal MHD instabilities via the form of MHD-correlated blobs, and they can

play an important role in the onset and strength of the internal MHD instabilities. The increasing

phase coupling between internal MHD modes and the edge turbulence is also observed with the

MHD-correlated blobs, suggesting the potential importance of edge turbulence in this

phenomenon.

References

[1] Takahashi, H., et al., Physics review letters 100.20 (2008)

[2] Borrass, K., Nuclear Fusion 31.6 (1991)

[3] Krasheninnikov, S.I., Physics Letters A 283.5-6 (2001)

*j.zhou@triam.kyushu-u.ac.jp

Investigation of EBW current drive in open magnetic flux surface using flux

loop with HFS injection in QUEST

ZHOU Junyao1*, HANADA Kazuaki2, KOJIMA Shinichiro3, KONO Kaori2, IDO

Takeshi2, EJIRI Akira4, ONCHI Takumi2, SEKIYA Izumi2, KURODA Kengoh2, YUE Qilin1,

IDEI Hiroshi2, NAGASHIMA Yoshihiko2, IKEZOE Ryuya2, HASEGAWA Makoto2,

HIGASHIJIMA Aki2, SHIMABUKURO Shun2

1Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Fukuoka,

Japan

2Research Institute for Applied Mechanics, Kyushu University, Fukuoka, Japan

3National Institutes for Quantum Science and Technology, Ibaraki, Japan

4Graduate School of Frontier Sciences, The University of Tokyo, Chiba, Japan

Due to the limited space for induction coils in the central part of the torus in spherical

tokamak (ST), RF injection is considered as an effectively auxiliary method to start up plasma

current. The high ratio of plasma frequency and electron cyclotron frequency in ST results in

the poor availability for plasma start-up by electron cyclotron wave (ECW) injection. Electron

Bernstein wave (EBW) is addressing a good possibility, instead of ECW [1]. QUEST has been

performed for efficient EBW with high field side (HFS) ECW injection and higher density could

be obtained than low field side (LFS) one [2,3]. In this presentation, the experimental results of

X-B mode conversion on HFS injection on QUEST to identify the effect of EBWCD is reported.

An incident antenna was located 0.17m above the equatorial plane of vacuum vessel to

inject the X-mode ECW from HFS perpendicularly to the magnetic field. The ECW was

expected to be converted to EBW at upper-hybrid resonance layer, then absorbed near the

electron cyclotron resonance layer. A closed flux surface (CFS) was successfully formed by

applying a magnetic configuration with n-index > 0.3. The direction of EBWCD is decided by

BT and BH directions, where BH denotes horizontal magnetic field. When the BT direction

was counter-clockwise (CCW) from the top view in the experiment, EBWCD was expected to

make an equilibrium current in the top half, where the horizontal magnetic field, BH, was

directed inside. Consequently, a peak of measured driven current appeared in top half, and the

peak changed to locate in the bottom half when the BT direction was inverted to clockwise

(CW). This indicates the main part of the equilibrium plasma current was driven by EBWCD.

The process to form CFS has a little different. In the condition of CCW BT, plasma current

showed almost constant in open magnetic flux surface (OFS) condition, and then jumped

suddenly to the maximum when CFS formed. When the BT direction was inverted, plasma

current gradually increased in OFS until the formation of CFS. To investigate further, an

additional BH, was applied. The plasma current ramp-up rate was obviously increased by

increasing BH for plasma up-shift. While, the BH for plasma down-shift was not useful for

making a CFS formation, which may due to the antenna located above the equatorial plane.

These results indicate that EBWCD played a significant role in formation of CFS and may be

controlled by applying BH to some extent.

References

[1] Ram A K, 2000, PP, 7(10): 4084-4094.

[2] Hatem E, 2020, PPCF 62(3): 035018.

[3] Kojima S, 2020, PFR 15: 2402063.

*peng-yi@g.ecc.u-tokyo.ac.jp

Investigation on anisotropic temperature in Ohmic plasmas in TST-2

Y. Peng1*, A. Ejiri1, N. Tsujii1, K. Shinohara1, O. Watanabe1, S. Jang1, K. Iwasaki1, Y. Ko1, Y.

Lin1, T. Hidano1, Y. Shirasawa1, Y. Tian1 and F. Adachi1

1The University of Tokyo

Non-inductive heating and current drive are essential elements for a steady state plasma in

a fusion reactor. However, heating methods could introduce temperature anisotropy for a short

time duration in fusion plasmas. Temperature anisotropy measurement is helpful in

understanding the heating and current drive mechanisms. Double-pass Thomson scattering (TS)

diagnostics is capable of measuring the perpendicular and parallel temperature (with respect to

the direction of magnetic field) simultaneously.

Some work on the investigation of temperature anisotropy had been implemented in TST-

2 previously. A double-pass TS configuration was developed to study temperature anisotropy

in OH plasmas, and there was temperature anisotropy with around 20% and 50% at the plasma

center and plasma edge respectively. Also, a coaxial multi-pass TS diagnostic system was set

up for the investigation on the existence of temperature anisotropy in the plasma driven by

lower hybrid wave (LHW). The results showed the two temperature components were isotropic

within about 10% error near the center.

There are a couple of drawbacks in the previous double-pass and coaxial multi-pass TS

measurements. The laser rod experienced damages in both configurations, which was caused

by the residual part of backward beam returning to the YAG laser output, while main part of the

backward beam is blocked. Although the coaxial multi-pass TS configuration could improve

the SNR (~2) somewhat, the timing control of polarization of the Pockels cell driver should be

very accurate and stable. Hence, the double-pass TS configuration is more easier to use.

Moreover, the previous double-pass TS configuration had only investigated the temperature

anisotropy in OH plasmas, now we plan to investigate the temperature anisotropy in LHW

plasmas, which is a more difficult condition with low density for TS measurement.

A new double-pass TS system has been developed in TST-2, considering the safety of the

laser device and the measurement performance of the double-pass TS measurement: the residual

backward power returning to the laser device should be suppressed sufficiently; the scattering

light collection efficiency should be high and similar for both forward and backward passes.

The beam radii (1/e power) along the laser forward and backward propagation have been

measured, which is around 0.35 mm to 1.65 mm in plasma region and agree with the design

values. The residual backward power returning to the YAG laser device is indirectly measured,

and it was estimated to be about 0.5% of the laser output power. The double-pass TS system

has been installed in TST-2 and now under testing with reduced YAG laser power of 1-2 W. The

measurement in OH plasmas has been preliminarily implemented. An example temperature at

R = 305mm (near the center) was 79 ± 16 eV and 127 ± 10 eV for the perpendicular and the

parallel (plasma current carrying) directions, respectively. The measurements for LHW

sustained plasmas will also be conducted in near future.

*yren@pppl.gov

Linear Instabilities in a ST40 Hot Ion Plasma

Y. Ren1*, S. Kaye1, A. Dnestrovskii2, M Romanelli2, M. Sertoli2

1 PPPL, 2 Tokamak Energy Ltd

The presently operating high-field compact spherical tokamak ST40 is an important step

toward ST-based fusion reactor [1]. Temperature and density profiles, and their uncertainties in

recent hot ion ST40 plasmas with central ion temperature exceeding 8.6 keV have been inferred

from experimental and modeling constraints in the absence of full profile measurements. Linear

gyrokinetic stability analysis with the initial value GS2 code [2] has been carried out to identify

the most unstable micro-instability for a range of in these plasmas. In one of these plasmas,

it is found that linear growth rates of both ion- and electron- scale () modes decrease

from the edge toward the core of the plasma (i.e., from to 0.3 where is the square root

of normalized toroidal magnetic flux). In particular, at no unstable mode is found at

, and about an order of magnitude reduction in the maximum linear growth rate in the

ion-scale wavenumber range of is seen at compared with the 

location. It is found that TEM and/or ubiquitous mode [3] may be important in this hot-ion

plasma.

References

[1] M. Gryaznevich and O. Asunta, Fusion Eng. Des. 123 177 (2017)

[2] M. Kotschenreuther et al., Comput. Phys. Commun. 88 128 (1995)

[3] Coppi B. and Rewoldt G. 1974 Phys. Rev. Lett. 33 1329

Magnetic reconnection: From compact fusion to plasma propulsion

Fatima Ebrahimi

Princeton Plasma Physics Laboratory, Princeton University, Princeton

Initial plasma-current formation via unconventional methods is a necessity in compact

fusion concepts such as spherical tokamaks (STs), one of the leading fusion reactor designs with

high magnetic field. A 3D current-sheet plasmoid instability, which is a common process both in

space, fusion [1-2] and astrophysical [3] plasmas have been shown to be instrumental during

unconventional magnetic helicity injection current start-up [4] in STs [5]. Multiple

contributions have been made in understanding of the nonlinear dynamics [6-9] during helicity

injection start- up including a transition from Sweet-Parker [8] reconnection to a plasmoid

instability that was for the first time simulated and predicted by global simulations with guide-

field.[6] Inspired by this research, state-of-the-art large-scale extended MHD simulations have

for the first time been used to demonstrate that a particular configuration of externally applied

static electric and magnetic fields can both create plasma and accelerate it by continuously and

spontaneously producing plasmoids, detached plasma objects enclosed by magnetic fields. [10]

Because the system-size plasmoid is an Alfvenic outflow from the reconnection site, its thrust

is proportional to the square of the magnetic field strength and does not ideally depend on the

mass of the ion species of the plasma (i.e., it does not require particular and rare gases as

propellant). Variable exhaust velocities were directly demonstrated in the simulations. Ongoing

research both for fast reconnection in fusion as well as plasma propulsion will be presented.

References

[1] F. Ebrahimi, PoP 23, 120705 (2016)

[2] F. Ebrahimi, PoP 24, 056119 (2017) & arXiv:2110.09706v1 (2021)

[3] J. Rosenberg and F. Ebrahimi, ApJL 920 L29, (2021)

[4] T. R. Jarboe, Fusion Technol. 15, 7 (1989)

[5] R. Raman, et al., PRL 97, 175002 (2006)

[6] F. Ebrahimi and R Raman PRL 114(20):205003, (2015)

[7] F. Ebrahimi and R Raman NF, 56(4):044002, (2016)

[8] F. Ebrahimi, PoP 26(9):092502, (2019)

[9] F. Ebrahimi, et al. PoP 20:090702, (2013)

[10] F. Ebrahimi, JPP 86, Issue 6, (2020)

 
*amancini@us.es 
Mechanical and electromagnetic design of the vacuum vessel of the SMART 
tokamak 
A. Mancini1*, D. J. Cruz-Zabala1, J. Segado-Fernández2, J. Garcia-Dominguez2, J. Ayllon-
Guerola3,M. Garcia-Munoz1, E. Viezzer1 and the PSFT teams 
 
1 Department of Atomic, Molecular and Nuclear Physics, University of Seville, Seville, Spain 
2 Centro Nacional de Aceleradores (U. Sevilla, CSIC, J. de Andalucia), Sevilla, Spain 
3 Department of Mechanical Engineering and Manufacturing, University of Seville, Seville, 
Spain 
 
The SMall Aspect Ratio Tokamak (SMART) is a new spherical machine that is currently 
being under construction at the University of Seville[1][2][3]. SMART will cover three different 
operational phases reaching a maximum inductive plasma current (IP) of 500 kA, a toroidal 
magnetic field (BT) of 1 T and a maximum pulse length of 500 ms[1][2][3] with the main goal of 
studying high plasma confinement  regimes  in  a  broad  triangularity  range  (-0.6≤δ≤0.6)[3][4]. 
These  operating  conditions  present  notable  challenges  for  the  design  of  the  device,  whose 
structural  integrity  needs  to  be  verified  against  all  the  possible events.  One  of  the  critical 
components is the vacuum vessel (VV) that needs to provide high quality vacuum and plasma 
performances  withstanding  the  electro-magnetic  (EM)  loads  during  normal  and  off-normal 
events. EM loads during the normal operation of the machine are due to the interaction of the 
induced eddy current with the magnetic field (forces). Instead, Vertical Displacement 
Events  (VDE)  and disruptions  constitute the  most  important  off-normal conditions  that  the 
tokamak can suffer as they can cause severe damage to the VV due to the significant EM and 
thermal loads delivered over ms timescales. A detailed evaluation of these loads is therefore 
mandatory for the correct dimensioning of the VV and its internal components. We present here 
the electromagnetic and structural analysis performed for the VV design. The modelling has 
been  carried  out  combining  equilibrium  baseline  scenario  for  phase  3[4]  developed  self-
consistently by using FIESTA[5] and ASTRA[6] codes. VDE and disruption events have been 
modelled with a Finite Element Model (FEM) allowing the estimation of the EM forces taking 
into account the effects of both eddy and halo currents[7]. An overview of the VV will also be 
given. 
 
References 
[1] A. Mancini et al., “Mechanical and electromagnetic design of the vacuum vessel of the 
SMART tokamak,” Fusion Engineering and Design, vol. 171, 10 2021. 
[2] M. Agredano-Torres et al., “Coils and power supplies design for the SMART tokamak,” 
Fusion Engineering and Design, vol. 168, 7 2021. 
[3] S.  J.  Doyle  et  al.,  “Magnetic  equilibrium  design  for  the  SMART  tokamak,”  Fusion 
Engineering and Design, vol. 171, 10 2021. 
[4] S. J. Doyle et al., “Single and double null equilibria in the SMART tokamak,” Plasma 
Research Express, vol. 3, 12 2021. 
[5] 
G.
 
Cunningham,
 
High
 
performance
 
plasma
 
vertical
 
position
 
control
 
system
 
for
 
upgraded
 
MAST,
 Fusion Engineering and Design 88 (2013) 3238–3247 
[6] G. Pereverzev G.V. and Yushmanov P.N. 2002 ASTRA automated system for transport 
analysis in a tokamak IPP-Report 5/98, (http://www2.ipp.mpg.de/~git/astra/astra.pdf) 
[7] A. Mancini et al, Electromagnetic VDE and disruption analysis in the SMART tokamak, 
IEEE Transaction on Plasma Science (Manuscript submitted for publication), 2022 

Modelling of the STEP ramp-down phase with JETTO

Zˇ .

Sˇtancar

F. J. Casson1, F. E. Eriksson1, F.

Ko¨chl

E. Tholerus1, and H. F. Meyer1

1United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon, UK

UKAEA’s prototype fusion power plant STEP (Spherical Tokamak for Energy Production)

envisioned to operate in a high confinement regime at a field and current of 3.2 T and 20 MA,

with a normalized pressure βN of ~ 4.4. The fusion power is expected to reach up to 1.8 GW in

the flat-top phase, meaning a high value of plasma stored energy. These operational conditions

neces- sitate the development of several safe and controlled termination pathways for STEP

discharges. In general there are several major constraints that affect the design of a viable ramp-

down: (i) the total plasma current at the end of ramp-down has to be sufficiently low for a

disruption to be tolerable, (ii) the plasma shape and strike points need to be controlled and stay

diverted to manage the decaying heat fluxes, (iii) the electron density measured relative to the

plasma current should not rise above the Greenwald limit, (iv) the internal inductance needs to

be kept sufficiently low to allow for vertical stability control.

In the paper we present the preliminary predictive modelling of STEP’s ramp-down phase

using the JETTO core transport code. We begin by describing the operational and modelling

limitations, specifically the assumptions on the relations between the internal inductance li,

plasma elongation κ, and βN, which dictate plasma vertical and MHD wall stability. We then

discuss results of the initial modelling exploration of viable ramp-down scenarios. The plasma

current in STEP is designed to be fully non-inductive, a combination of bootstrap and externally

EC- driven current, with the external heating based solely on EC. The ramp-down of both the

current and fusion power can thus be controlled by the injected EC power. The simulations are

launched from a converged reference flat-top scenario, after which the EC power and current

are gradually ramped-down, monitoring the current diffusion and fusion power decay. In order to

navigate within the stable li/κ and βN/li phase space, we vary the EC ramp-down rate and its

waveform, the pellet fuelling source rate, and include feedback controls on the Greenwald

density fraction. We show that in order to stay within the operational limits, the EC current

ramp-down rate should be below ~ – 5 kA/s, resulting in a total phase duration of several tens

of minutes. We show that it is the decay rate of the induced current that dictates the speed of the

overall total current ramp-down. It is found that the feedback current rise rate and peak value

grow with increasing EC ramp-down rates, and can be the dominant current component with a

clear rollover shape. Due to the peakedness of the inductive current density profile it can have

a detrimental effect on li and the q-profile, which becomes non-monotonic. However, for ramp-

down rates below – 5 kA/s the inductive current reaches a saturation point, contributing below

30 % to the total current, and stays stable up until the last 10 % of the ramp-down phase, when

the remaining EC and bootstrap contributions diminish. In order to decouple the fusion power

and current ramp-down a scenario in which the pellet fuelling isotope is changed from DT to

pure D is discussed.

Acknowledgements

This work has been funded by STEP, a UKAEA program to design and build a prototype

fusion energy plant and a path to commercial fusion.

*inomoto@k.u-tokyo.ac.jp

Modification of energy conversion process in magnetic reconnection during

merging formation of spherical tokamak

M. Inomoto1*, H. Jin1, T. Suzuki1, K. Nakatsukasa1, H. Tanabe1, Y. Ono1

1Graduate School of Frontier Science, the University of Tokyo, 277-8561, Kashiwa, Japan

Merging formation technique has been developed as one of the center-solenoid free start-

up schemes of spherical tokamak (ST) plasmas. Two initial STs are inductively formed with

use of outer poloidal field coils and then they axially collide and merge to form a single ST

through magnetic reconnection. Since poloidal magnetic field reconnects during the merging

formation process, the existence of high toroidal field, that is called “guide field”, perpendicular

to the reconnecting field substantially changes the reconnection dynamics. Charged particles

are kept magnetized even in the vicinity of the reconnection X-point where high toroidal guide

field exists. The reconnection (toroidal) electric field is nearly parallel to the magnetic field in

the vicinity of the X-point and charged particles, particularly electrons, gain energy from the

electric field component parallel to the magnetic field as parallel acceleration.

In the downstream region of the guide-field reconnection, it is considered that the self-

generated in-plane static electric field will cancel the parallel component of the inductive

electric field, resulting in forming the steady reconnection outflow with equal electron and ion

velocities. This well-known steady condition could be broken when the in-plane static electric

field had been suppressed. Axially-continuous but azimuthally-discontinuous conducting

structure is capable of connecting a magnetic field line at two locations in the reconnection

downstream region to release the charge separation condition [1] and enhance electron parallel

acceleration [2]. This short-circuit effect could make further modification of reconnection

dynamics such as reduction of ion cross-field outflow velocity, potentially changing the energy

partition between ions and electrons. Adequate energization of electrons will improve the

merging formation scheme suitable for providing a target ST with sufficient electron

temperature to be sustained by external heating sources.

References

[1] M. Inomoto, et al., Nucl. Fusion 61, 116069 (2021).

[2] M. Inomoto, et al., J. Phys. Soc. Japan 91, 094501 (2022).

*franco.alladio@protosphera.it

Morphology: PROTO-SPHERA experiment vs. the inner wind region of the

Crab Nebula supernova remnant

F. Alladio1* and PROTO-SPHERA Group2

1INAF – IAPS, Via Fosso del Cavaliere 100, 0133 Rome, Italy

2Centro Ricerche ENEA, Frascati C.P. 65, Rome, Italy

The inner (e+/e-) wind region of the Crab Nebula supernova remnant, which is

continuously replenished by plasma emitted by the central pulsar, has a morphology quite

similar to the one of the PROTO-SPHERA experiment. A closed magnetic flux surface plasma

torus and two kinked and broad plasma jets, with open magnetic field lines, characterize both

systems. The recent observation of the tilted azimuthal rotation of the plasma and of the non-

axisymmetric nature of the recurring magnetic reconnections, which create and sustain the

PROTO-SPHERA plasma torus, show that the production of the closed toroidal flux surfaces

is associated in PROTO-SPHERA with a plasma that acts as an oblique rotator. Oblique plasma

rotators are ubiquitous in the case of pulsar winds. Another similarity is the presence of a neutral

gas dominated component that surround the plasma tori in both cases, PROTO-SPHERA and

the Crab Nebula remnant. Apart from the presence of the gravitational field in the case of the

supernova remnants, whose virial action is replaced by external poloidal magnetic field coils in

the case of PROTO-SPHERA, important differences between the plasma of the two objects are

evident: the Crab Nebula wind plasma is a relativistic e+/e- couple plasma, whereas the

PROTO-SPHERA plasma is, at the moment, a few tens of eV Hydrogen or Helium plasma.

Magnetic reconnections in the inner wind region seem to be the origin of the powerful X-ray

flares that have been observed since about 10 years ago from the Crab Nebula. A quite relevant

observation is that the supernova remnants are the main sources of high energy cosmic rays in

the Galaxy: the Crab Nebula is a Pevatron accelerator, able to produce cosmic rays at energies

higher than 100 TeV. An increasing consensus among the astro-particle scientific community

suggests that such accelerations is due to the magnetic reconnections occurring in the inner

wind region of the supernova remnants. Interesting geometrical and dynamical similarities

could therefore exist in the magnetic reconnections, which in PROTO-SPHERA occur on a

limited azimuthal portion of the two circles of ordinary X-point, one above and one below the

equator; in the inner (e+/e-) wind part of the Crab Nebula supernova remnant, magnetic

reconnections could occur as well limited azimuthal portion of circles at the edge of the light

cylinder.

*cliu@pppl.gov

Nonlinear simulation of pressure-driven MHD modes and beta-limit study

in MTF

Chang Liu1*, Dylan P. Brennan and Aaron Froese2

1 Princeton Plasma Physics Laboratory

2 General Fusion

Magnetized Target Fusion (MTF) is a new-concept fusion device currently pursued by

General Fusion Inc. In this device, a spherical tokamak can be formed by injecting plasma into

a spherical cavity. The plasma is then compressed by liquid metal liner driven by an array of

pistons, to level up the plasma density and temperature to reach the fusion criterion. Both ideal

and resistive linear MHD stability for n=1 modes have been studied using DCON code for

different values of shaft currents and different compression states. Assuming the compression

process is almost self-similar and the q profile is almost fixed, a stable corridor has been

identified [1], which means that resistive tearing mode can be suppressed during the

compression process when an appropriate q profile is achieved.

In this work we conducted linear and nonlinear ideal-MHD simulation of high-n modes

using the M3D-C1 code, to study the effects of pressure-driven instabilities. This work is

inspired by a recent study on the connection between the pressure-driven MHD modes and the

soft-beta limit in NSTX [2]. We focus on the scenario with high compression ratio. It is found

that with a large core plasma beta (>20%), pressure-driven modes can become unstable with n

ranging from 4 to 8. In the nonlinear simulation, it is found that the excited modes can lead to

flattening of temperature profile at core and near the edge, but steepen the temperature gradient

in the intermediate region. The steepening can finally lead to a second wave of MHD mode

bursts, breaking flux surfaces and cause significant transport of internal energy.

The simulation result indicates that avoiding pressure-driven mode is important for a stable

operation of MTF, and the beta-limit for a highly compressed plasma is lower compared to a

normal-sized ST like NSTX. Further simulation regarding the kinetic effects in the MHD modes

will also be discussed.

References

[1] D. Brennan, A. Froese, M. Reynolds, S. Barsky, A. Wen, Z. Wang, M. Delage, and M.

Laberge, Nucl. Fusion 61, 046047 (2021).

[2] S.C. Jardin, N.M. Ferraro, W. Guttenfelder, S.M. Kaye, and S. Munaretto, Phys. Rev.

Lett. 128, 245001 (2022)

Figure 1. MHD mode energy and plasma pressure evolution in M3D-C1 nonlinear simulation

*wmyuan@mail.ustc.edu.cn

Observation of whistler wave instability driven by temperature anisotropy

of energetic electrons in plasmas of EXL-50 spherical tours

Mingyuan Wang1,2*, Yuejiang Shi1,2, Jiaqi Dong1,2, Shikui Cheng1,2, Mingsheng Tan3,

Songjian Li1,2, Shaodong Song1,2, Tiantian Sun1,2, Hanyue Zhao1,2, Xianli Huang1,2, Yumin

Wang1,2, Yingying Li1,2, Xianming Song1,2, Minsheng Liu1,2, Baoshan Yuan1,2, Y-K Martin

Peng1,2, and ENN team.

1 Hebei Key Laboratory of Compact Fusion, Langfang 065001, China

2 ENN Science and Technology Development Co., Ltd., Langfang 065001, China

3 Institute of Energy, Hefei Comprehensive National Science Center, Hefei 230031, China

Whistler waves driven by temperature anisotropy energetic electrons are observed in

electron cyclotron wave (ECW) steady state plasmas on XuanLong-50 (EXL-50). The whistler waves

have multiple frequency bands which are proportional to the Alfvén velocity. Increasing of the

bulk electron density or decreasing of the energetic electron temperature anisotropy reduces the

intensity of the whistler wave and decreases the loss of energetic electrons. Moreover, small

drops in plasma current correlated with the whistler waves are also observed, probably also

affecting the equilibrium profile. These results suggest that ECW current-drive through the

creation of energetic electrons may excite such waves potentially causing a limitation to the

current drive efficiency. Details will be presented. These provide a novel experimental way to

study wave-particle interactions and whistler wave control of runway electrons on future fusion

devices.

Figure1. Time

evolution of (a) the

magnetic fluctuation

power spectra; (b) the

toroidal magnetic field

(near the magnetic axis),

and line integrated

density. Variation of

mode frequency spectrum

with (c) line integrated

density at constant

toroidal magnetic field

and (d) Alfvén velocity.

References

[1] Fülöp T, Pokol G, Helander P, et al. Destabilization of magnetosonic-whistler waves by a

relativistic runaway beam[J]. Physics of Plasmas, 2006, 13(6): 062506.

[2] Spong D A, Heidbrink W W, Paz-Soldan C, et al. First direct observation of runaway-

electron-driven whistler waves in tokamaks[J]. Physical Review Letters, 2018, 120(15):

155002.

Overview of ECRH and EBW modelling on ST40

E. du Toit1, V. Shevchenko1, M. Ono2 and N. Bertelli2

1Tokamak Energy Ltd, 173 Brook Drive, Milton Park, Oxon, OX14 4SD, UK

2Princeton Plasma Physics Laboratory, Princeton, NJ, 08543, USA

Electron cyclotron resonance heating (ECRH) and current drive (ECCD) is a well

established and widely used approach to heating plasmas and driving current, and is of

particular importance for enabling non-inductive start-up and ramp-up in spherical tokamaks,

due to the limited space for a shielded central solenoid.

A dual-frequency ECRH system is currently under development on the ST40 spherical

tokamak, based at Tokamak Energy Ltd., Oxfordshire, UK. The system will employ two

gyrotrons with a maximum output power of 1 MW each. Both gyrotrons can be tuned to operate

at either 105GHz or 140GHz, which allows for a study of non-inductive plasma start- up, current

ramp-up and sustainment.

In this paper we will present modelling results of ECRH capabilities in ST40 of both

available gyrotrons with various launch configurations and operating frequencies, including

methods under consideration for non-inductive plasma start-up. Modelling of plasma heating

and current drive was performed using ray-tracing codes GENRAY and RT-4, as well as the

Fokker-Planck code CQL3D.

A low field side (LFS) steerable launcher will be used for the study of conventional 2nd

harmonic X-mode ECRH. Modelling shows good absorption across a wide range of parameters.

This launcher can also be used for the study of fundamental X-mode heating, due to the high

magnetic field on the ST40 tokamak. Modelling results show that this setup can be used to

generate significant currents at low densities and temperatures applicable to start- up conditions.

The LFS launcher will also be used for the excitation of electron Bernstein waves (EBW)

in an O-X-B scheme first demonstrated on MAST [1]. An O-mode beam is launched from the

LFS and converted to X-mode with a grooved mirror polarizer located on the centre post. The

X-mode propagates back into the plasma and undergoes mode conversion to EBWs at the upper

hybrid resonance (UHR), which are subsequently absorbed. A high field side (HFS) launcher is

also considered for the excitation of EBW via direct X-B mode conversion.

Modelling results for both setup schemes show strong absorption of EBWs across a range

of parameters, with a maximum current drive efficiency of 0.15A/W expected.

References

[1] V.F. Shevchenko et al., Nucl. Fusion 50, 022004 (2010).

Overview of EXL-50 Research Progress and Future Plan

Yuejiang Shi, Bing Liu, Shaodong Song, Xianming Song, Huasheng Xie, Yuanming Yang,

Jiaqi Dong, Kaiming Feng, Baoshan Yuan, Y.-K.Martin. Peng, Minsheng Liu, EXL-50 team+

1Hebei Key Laboratory of Compact Fusion, Langfang 065001, China

2Enn Science and Technology Development Co., Ltd, Langfang 065001, China

XuanLong-50 (EXL-50) is the first medium size spherical torus (ST) in China. The

engineering design of EXL-50 started in October 2018, the main body of the device was

completed in June 2019, and the first plasma discharge was successfully obtained in July 2019.

The toroidal field (TF) of EXL-50 is around 0.5T at major radius = 50cm. An important

characteristic of EXL-50 is that it does not have a central solenoid (CS). CS-free non-inductive

current drive via electron cyclotron resonance heating (ECRH) is one of the main research target

in EXL-50’s first experimental phase. Discharges with plasma currents of 50kA -160 kA are

routinely obtained in EXL-50, with the current flattop sustained for up to or beyond 2 s. The

averaged current drive effectiveness on EXL-50 is around 1 A/W for the discharges with ECRH

heating power less than 200kW. The plasma current reaches Ip > 80 kA for high density (>5 ×

1018 m−2) discharge with 150kW ECRH. The plasma current in EXL-50 is mainly carried by the

energetic electrons. Multi-fluid equilibrium model has been successfully applied for the

reconstruction of magnetic flux surface and equilibrium of EXL-50’s plasma. The physics

mechanism for the solenoid-free ECRH current drive and energetic electrons has also been

investigated. The initial experimental results also show that the energy confinement of EXL-

50’s plasma has some special properties. On the other hand, low hybrid current drive (LHCD)

system and ion cyclotron resonance heating (ICRH) system have also been installed on EXL-

50. The initial experimental results of LHCD and ICRH are also presented. Neutral beam

heating (NBI) system will be installed on EXL-50 in the end of 2022. The TF coil of EXL-50

will upgraded to provide higher magnetic field (>1T @R=0.5m) in 2024. The vacuum vessel

and PF coils will also be modified for flexible plasma shaping and current control. Except for

non-inductive current drive for high density plasmas, investigation of hot ion mode and

transport/confinement performance of ST plasmas in high magnetic field are the main purposes

of EXL-50’s upgrade.

References

[1] Y.-K. M. Peng et al, Plasma conditions under which the p-11B fusion Lawson criterion

could be eased, https://assets.researchsquare.com/files/rs-93644/v1/ea84d194-4141-4f9b-

90c3-e262fdccb237.pdf;P.R.China Patent pending (application no. 202010292584.X);

[2] Yuejiang Shi, et al, Solenoid-free current drive via ECRH in EXL-50 spherical torus

plasmas, Nucl. Fusion 62(2022) 086047;

[ 3 ] Ishida A, Peng Y K M and Liu W, Four-fluid axisymmetric plasma equilibrium model

including relativistic electrons and computational method and results, Phys. Plasmas,

28(2021)032503;

[4] Guo Dong, et al, Experimental study of the characteristics of energetic electrons

outside LCFS in EXL-50 spherical torus, Plasma Phys. Control. Fusion

64(2022)055009;

[5] Mingyuan Wang, et al, Experimental study of non-inductive current start-up using

electron cyclotron wave on EXL-50 spherical torus, Plasma Phys. Control. Fusion

64(2022)075006

† See the author list of J.R. Harrison et al 2019 Nucl. Fusion, 59 112011

Overview of First Exhaust Physics Results from MAST-U

S. Elmore for the MAST-U Team†

United Kingdom Atomic Energy Authority, Culham Centre for Fusion Energy, Culham Science

Centre, Abingdon, OX14 3DB, UK

The particle and power exhaust solution for fusion reactors must produce tolerable power

loads on plasma-facing-components whilst being compatible with the high-performance fusion

core, especially in compact, high power devices such as strongly heated spherical tokamaks.

Alternative divertor configurations are being explored as a possible exhaust solution in MAST-

U, which has the unique capability to compare a conventional divertor with alternative

configurations due the large number of divertor coils. The focus of the first campaign on MAST-

U has been developing an elongated Super-X divertor to test this exhaust solution in the new

tightly baffled divertor chambers. The Super-X configuration moves the outer strike point to a

greater radius which increases the length of field lines connecting the outer mid-plane to the

outer divertors and the total flux expansion, which can be significantly higher in spherical

tokamaks. This reduces the heat flux to the outer target and improves access to the detached

divertor regime due to the combined effect of increased wetted area, enhanced volumetric power

losses and cross-field transport.

This contribution will present the exhaust results from the first MAST-U physics campaign.

The aim of the main exhaust experiment was to study the advantages of the closed divertor with

the Super-X configuration. An ohmic heated L-mode scenario in double-null configuration with

plasma currents of 600 and 750 kA and elongation of ~2 was developed for both conventional

and Super-X configurations. Midplane Thomson scattering measurements of electron

temperature and electron density were similar for the two configurations, which shows that the

core plasma conditions were not adversely affected by the Super-X configuration or detached

outer divertors. Ion flux measurements at the outer target by Langmuir probes indicate that the

detachment threshold is 50% lower for the Super- X configuration. During attached divertor

conditions, the peak heat flux measured by IR thermography was a factor ~10 lower for the

Super-X configuration. Multi-wavelength imaging and spectroscopy diagnostics identified

distinct regions of ionisation, molecular activated recombination and electron-ion

recombination in the divertor chamber during the detachment evolution. Results from the multi-

fluid SOLPS-ITER code for this experiment indicate (i) particle drifts do not seem to play a

major role in the outer scrape-off-layer; (ii) the outer target ion flux rollover in conventional

could be driven by power loss from the core, rather than from the divertor, which motivates

going to a higher heating power in future experiments.

Acknowledgements

This work has been funded by the RCUK Energy Programme [grant number

EP/T012250/1]. To obtain further information on the data and models underlying this paper

please contact PublicationsManager@ukaea.uk.

Lead-author e-mail: h-tanaka@energy.kyoto-u.ac.jp

Overview of Low Aspect ratio Torus Experiment (LATE): Recent Results

and Plans

H. Tanaka, M. Uchida, R. Kajita, W. Ootaka, S. Ozaki, M. Hamaguchi, R. Kobayashi,

M. Komeda Y. Yasui

Graduate School of Energy Science, Kyoto University, Kyoto, Japan

The primary objective of Low Aspect-ratio Torus Experiment (LATE) is to study on non-

in- ductive production of spherical tokamak plasmas by using ECH/ECCD. In order to produce

the over-dense plasmas, O-mode microwave is injected obliquely to the toroidal field from

outboard side to excite electron Bernstein wave (EBW).

Recently, by using a newly developed five-pin probe antenna and two-dimensional (2-D)

me- chanical probe driving system, the 2-D wave pattern of phase and amplitude has been

directly measured for an overdense ECR plasma with 1.5 GHz microwave. In the case of O-

mode injec- tion, an EBW-like wave pattern has been detected for the first time, in a localized

region near the upper hybrid resonance layer. The pattern has a short wavelength of about 2 mm

and is also elec- trostatic and backward. By adjusting the toroidal magnetic field, it is found that

both the position and size of the EBW region have changed, which suggest the localized

condition of efficient O-X-B conversion and high collisional damping rate of EBWs.

In May 2021, the LATE device and all the diagnostic apparatuses and the heating systems

were disassembled and moved to Uji campus (~10 km away from Yoshida campus) because of

repair work of Plasma Wave Experiment Building. In May 2022 after the completion of the re-

pair work, all the materials have moved back and begun re-assembling. New 66 ports in total

have been added to the original vacuum vessel. The first plasma with this new vessel will be

produced in this October.

Acknowledgements

This work is supported by JSPS KAKENHI Grant Numbers 21H01062.

Overview of Physics Results from the First Campaign of Experiments on

MAST Upgrade

J. Harrison on behalf of the MAST-U and EUROfusion Tokamak Exploitation teams

United Kingdom Atomic Energy Authority, Culham Centre for Fusion Energy, Culham Science

Centre, Abingdon, Oxon, OX14 3DB, UK

MAST Upgrade is a new spherical tokamak based on the MAST device [1], with low

aspect ratio (R/a ~ 0.85/0.6 ~ 1.4), increased toroidal field (up to 0.8T at R = 0.8m) and

maximum plasma current (up to 2 MA) and pulse length (up to 5 s). The two positive ion neutral

injectors (NBI) are capable of injecting up to 2.5 MW to provide on and off-axis heating and

current drive. It has a large number of poloidal field coils, providing significant flexibility to

modify the shape of the core plasma (elongation κ ~ 2.2 achieved, maximum ~2.5, triangularity

δ ~ 0.5) and magnetic geometry in the divertors. The divertors are capable of producing and

studying conventional and alternative divertor configurations, including Super-X, X-divertor,

snowflake and X-point target with up-down symmetric, tightly baffled divertor chambers (see

[1] and references therein). An extensive suite of diagnostics enables detailed physics studies

of key issues for fusion, including transport of thermal and super-Alfvenic fast particles in the

core and pedestal, MHD stability and plasma exhaust.

First plasma was achieved in October 2020, followed by commissioning of real-time

control systems and the first physics campaign took place from March to October 2021. A major

theme of the physics campaign was plasma exhaust, specifically characterizing and

understanding the differences between conventional and Super-X divertor configurations. In

spherical tokamaks, the Super-X configuration is predicted to have more significant benefits

due to the strong reduction of the toroidal field with increasing strike point major radius and

commensurate increase in the wetted area. Initial experiments with 3.2 MW of injected NBI

power demonstrated a factor ~10 reduction in outer divertor power loads in the Super-X

configuration compared with a conventional divertor, in good agreement with modelling

predictions [2, 3]. More detailed experiments in L-mode confirmed these findings and found a

factor ~2 reduction in the outer mid-plane separatrix density to detach the outer divertors.

A combination of increased divertor closure and better plasma shaping improves control of

the edge density which in turn results in pedestal temperatures reaching 400 eV. ELM mitigation

was achieved with n=1 RMP coils. MAST-U pedestals are peeling mode limited which may aid

access to ELM-free H mode scenarios.

High energy, type-1 ELMy H-mode plasmas were sustained for 1 second using both beams

to inject an average of 3 MW of power. A super-Aflvénic fast particle population was found to

excite a range of instabilities, including suspected toroidal Alfvén eigenmodes, fishbones and

compressional or global Alfvén eigenmodes.

Acknowledgements

This work has been funded by the RCUK Energy Programme [grant number

EP/T012250/1]. To obtain further information on the data and models underlying this paper

please contact PublicationsManager@ukaea.uk.

References

[1] W. Morris et al, IEEE Transactions on Plasma Science, vol. 46, no. 5, pp. 1217-1226, 2018.

[2] E Havlíčková et al Plasma Phys. Control. Fusion 56 075008 (2014)

[3] D. Moulton et al,. Proc 44th EPS Conf. 2017

Overview of the SMall Aspect Ratio Tokamak - SMART

M. Garcia-Munoz1,2, A. Mancini1,2, J. Segado-Fernández2, D. J. Cruz-Zabala1,2, J.

Garcia- Dominguez2, J. Ayllon-Guerola2,3, E. Viezzer1,2, J. Galdon-Quiroga1,2 and the PSFT

team

1 Department of Atomic, Molecular and Nuclear Physics, University of Seville, Seville, Spain

2 Centro Nacional de Aceleradores (U. Sevilla, CSIC, J. de Andalucia), Sevilla, Spain

3 Department of Mechanical Engineering and Manufacturing, University of Seville, Seville,

Spain

Spherical Tokamaks (STs) offer an attractive path to compact Fusion Power Plants (FPP)

with high power densities. Compact STs are, however, subject to specific challenges inherent

to their tight aspect ratio and high wall loads. Experiments at the TCV and, more recently, DIII-

D conventional tokamaks show that Negative Triangularity (NT) shaped L-mode plasmas

feature H-mode like confinement properties without Edge Localized Modes (ELMs) emerging,

thus, as a promising scenario for a tokamak-based FPP. NT shaped plasmas might become even

more attractive for a ST- based FPP due to their confinement and power exhaust properties. The

investigation of NT plasmas in present devices is, however, somewhat cumbersome and

hampered by the geometry of the device which is mostly optimized for Positive Triangularity

shaped plasmas.

The SMall Aspect Ratio Tokamak (SMART) is a new ST device designed to operate at Bt

up to 1T and Ip up to 1 MA with the magnetic axis Rmag=0.45m for standard ST configurations

with aspect ratios A (Rmag/a) = 2. The vacuum vessel as well as the coils and power supply

systems have been designed to allow flexible plasma shaping with aspect ratios 1.4 ≤ A ≤ 3

and triangularities -0.6 ≤ δ ≤ 0.6. An overview of the SMART device together with its status and

physics basis will be presented. Implications for future ST-based FPP with NT shaped plasmas

will be discussed.

References

[1] Mancini et al., “Mechanical and electromagnetic design of the vacuum vessel of the

SMART tokamak,” Fusion Engineering and Design, vol. 171, 10 2021.

[2] M. Agredano-Torres et al., “Coils and power supplies design for the SMART tokamak,”

Fusion Engineering and Design, vol. 168, 7 2021.

[3] S. J. Doyle et al., “Magnetic equilibrium design for the SMART tokamak,” Fusion

Engineering and Design, vol. 171, 10 2021.

[4] S. J. Doyle et al., “Single and double null equilibria in the SMART tokamak,” Plasma

Research Express, vol. 3, 12 2021.

Physics drivers of the STEP divertor concept design

S S Henderson1, S L Newton1, R T Osawa1, D Moulton1, J Harrison1, L Xiang1, F Militello1,

M Kryjak2, C Cowley2, B Lipschultz2, B Dudson2, C Ridgers2, A Hudoba1, G Voss1, G

Cunningham1, S Kahn1, A Tarazona1, B Chuilon1, A Barth1, S Wang1 and J Farrington1

1UKAEA-CCFE, Culham Science Centre, Abingdon, OX14 3DB, UK

2York Plasma Institute, Department of Physics, University of York, Heslington, York

YO105DD, UK

The STEP programme aims to demonstrate viability of a GW-scale spherical tokamak

reactor, with parameters around: geometric axis 3.6m, aspect ratio 1.8, elongation 2.9, plasma

current 20MA, on-axis magnetic field 3.2T, core radiated power 350MW, with steady state

power crossing the separatrix 150MW. Here we give a physics overview of the studies driving

forward the STEP divertor concept design, focussing on operational phases during current ramp-

up and flat-top. The inner divertor in a spherical tokamak is challenging, with total flux

compression along the leg and small strike point major radius suggesting unmitigated surface

power loads of GW/m2. The down-selected operational point thus focused around a double-null

configuration, with surface area and poloidal variation of transport weighting the core power

outflux towards the outboard. The outer divertor leg was extended to large major radius to

increase the target wetted area, with unmitigated loads up to ~200MW/m2, managed by seeding.

Our recent studies indicate how synergy between connection length and flux expansion can be

used to improve the detachment threshold and window, as well as front location sensitivity.

Advantages of double-null geometry will only be realised with vertical position control

precision to 1-2 millimetres, based on anticipated heat flux decay widths, so we have examined

the impact of separatrix disconnection, dRsep, on target fluxes. Up-down power sharing was

found to be consistent with previous experimental modelling [Brunner et al (2018) Nucl. Fusion

58 076010], but the strong flux compression from outboard midplane to inboard target

contributed to weak sensitivity of in-out power sharing, with the impact of drifts on this

conclusion under investigation. This motivates testing of dynamic double-null (full power

alternating from upper to lower divertor) for practical vertical stability control, with modelling

indicating strong frequency and amplitude dependence of divertor component fatigue by

temperature cycling. Partial detachment was recovered in a vertical inner target configuration,

with high far-SOL temperatures driving unacceptable erosion. Full detachment required direct

extrinsic impurity seeding or the introduction of an advanced divertor geometry, with the former

giving unacceptably high core fuel dilution.

A toolkit has been developed which allows optimisation of divertor magnetic geometries

in free boundary magnetic equilibria, in combination with compatible first wall geometries,

while holding constraints within allowed margins. Constraints include core profiles, informed

by stability and transport analysis, and field coil positions, defined by neutron shielding and

current limitations, while divertor performance measures, such as connection length and

poloidal flux expansion, are returned. Performant geometries are passed to SOLPS-ITER, to

assess trends in power handling ability. The detached operating space is identified in terms of

deuterium puffing and argon seeding, while minimising gas throughput, and maintaining low

upstream density and impurity concentration. The inner divertor now focuses on a design

approaching an X-divertor, finding substantial improvements in performance allowing reduced

seeding levels, with ionisation profiles capturing the benefits of angled targets. The effect of

pump location and inner target geometry on helium pumping efficiency is being optimised. The

potential impact of gas puff location and pumping speed are also studied.

Ken.McClements@ukaea.org.uk

Plasma Scenario Development in the Spherical Tokamak for Energy

Production (STEP)

K.G. McClements for the STEP Plasma, Control and Heating & Current Drive Team

United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon, OX14 3DB, UK

The UK-based STEP programme aims to develop by 2040 a prototype reactor based on

the spherical tokamak (ST) concept, thereby establishing a basis for developing commercial

electricity production from fusion [1]. The compact design restricts the possible inductive flux,

hence the flat-top plasma current will be entirely non-inductive, enabling long-pulse operation.

External current drive will be delivered using electron cyclotron and electron Bernstein waves.

Modelling using the transport code JETTO indicates that the time required to ramp up the

current will be around 103s or longer, in part to avoid large back-EMFs. To demonstrate the

viability of an ST-based commercial reactor, the flat- top phase will ultimately need to be much

longer than the ramp-up, while the energy confinement time will be a few seconds. STEP will

therefore need to be a steady-state device, with a fuel cycle and control system that keep the

plasma close to a target scenario. The ST concept makes it possible to maximise fusion power

and bootstrap fraction in a compact device at relatively low toroidal field by allowing operation

at high normalised pressure and elongation, but it also poses unique challenges. Double null

operation will be used, with divertors that are resilient to losses of up-down symmetry.

Operating in a highly self-organised, high beta scenario with large radiation and bootstrap

fractions will require novel control techniques. The prediction of confinement is particularly

challenging. For a high beta ST such as STEP, presently-available scaling laws and reduced

transport models are well outside their domain of experimental validation. Moreover, parameter

dependencies differing from those in scaling laws derived for conventional tokamaks have been

observed in present day STs, and reduced models need to be modified due to the

electromagnetic nature of the turbulence expected to dominate. Three flat-top operating points

with different assumptions have been developed, with input from the systems code PROCESS.

Linear gyrokinetic (GK) modelling shows that the turbulence in these flat-top plasmas is

dominated by micro-tearing modes (MTMs) and kinetic ballooning modes (KBMs).

Diamagnetic flow shear is stabilising for KBMs and other modes at high wave number k, and

the turbulence is likely to be dominated by MTMs at low k: this makes the non-linear GK

modelling very challenging. Actuators to control the transport are being investigated to seek

routes to optimised confinement. Reduced models capturing the magnetic flutter-driven

electron heat transport are being tested using the integrated modelling suite JINTRAC. Results

so far support assumptions made for the simpler Bohm-gyro-Bohm model used for the concept

evaluation. We have identified a preferred scenario based on the exploration of several

prototype concepts with different plasma and technology assumptions. In this scenario fusion

alpha-particle-driven toroidal Alfvén eigenmodes are predicted to be suppressed due to high

beta. I will give an overview of STEP and discuss the physics design of the preferred scenario.

I will present the key challenges and assumptions leading to the scenario choice and discuss the

modelling framework being used to reduce uncertainties. This work suggests that a compact

fusion reactor may indeed be feasible.

Acknowledgements

This work has been funded by STEP, a UKAEA program to design and build a prototype

fusion energy plant and a path to commercial fusion.

References

[1] H.R. Wilson et al., STEP – on the pathway to fusion commercialization, Ch 8 of

Commercialising Fusion Energy, IOP Publishing (2020).

Plasma shaping effects on the neoclassical tearing mode threshold physics

A V Dudkovskaia1, J W Connor2, D Dickinson1, P Hill1, K Imada1, S Leigh1, T Shi3 and H R

Wilson1

1York Plasma Institute, School of Physics, Engineering and Technology, University of

York,Heslington, York YO10 5DD, UK

2UKAEA-CCFE, Culham Science Centre, Abingdon Oxon OX14 3DB, UK

3Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China

Neoclassical tearing modes (NTMs) degrade tokamak confinement and sometimes can

terminate plasma discharges in a disruption. Understanding the NTM stabilisation techniques

requires knowledge of the NTM threshold magnetic island width below which tearing modes

are self-stabilised [1]. A new drift kinetic nonlinear theory of small magnetic islands relevant

to the NTM threshold has been developed [2, 3] in the limit of a low beta, large aspect ratio,

circular poloidal cross section tokamak geometry. In [4] this theory is extended to finite beta,

arbitrary magnetic field configurations, relevant to spherical tokamak plasmas.

Small magnetic islands of width  are assumed (is the trapped ion banana

orbit width and  is the radius of the rational surface associated with the magnetic island). This

ensures that the finite orbit width effects of trapped ions are captured and provides a small

expansion parameter  to reduce the 5D drift kinetic equation to a 4D problem. When

collisions are rare (i.e. the particle collision frequency is much smaller than the characteristic

drift frequency), the particle distribution function is found to be flattened across drift islands

[3], i.e. phase space islands shifted radially by a value proportional to the poloidal Larmor radius,

relative to the location of the magnetic island. This then supports a contribution to the density

gradient within the magnetic island associated with passing ions

when (with  being the tokamak major radius). When , ions

and electrons respond differently to the magnetic island which generates a perturbed

electrostatic potential to ensure quasi-neutrality and thus restores the electron density gradient

across an island of width . This model provides a physics basis for the NTM threshold.

The threshold island width is predicted to be which is in close quantitative

agreement with experiment [1]. Inclusion of finite aspect ratio effects in [4] shows a

relatively strong impact of plasma triangularity on the threshold physics: negative triangularity

is found to enhance stabilising effects of drift islands, while positive triangularity, in contrast,

reduces the NTM threshold, making the higher triangularity plasma more prone to NTMs.

Finally, the threshold island width dependence on poloidal beta was obtained in [4] and

benchmarked against the EAST threshold island width experimental measurements: they are

found to be in good agreement.

References

[1] R.J. La Haye, R.J. Buttery, S.P. Gerhardt et al. Phys. Plasmas 19 (2012) 062506

[2] K Imada, H R Wilson, J W Connor, A V Dudkovskaia and P Hill Nucl. Fusion 59 (2019)

046016

[3] A V Dudkovskaia, J W Connor, D Dickinson, P Hill, K Imada, S Leigh and H R Wilson

Plasma Phys. Control. Fusion 63 (2021) 054001

[4] A V Dudkovskaia, L Bardoczi, J W Connor, D Dickinson, P Hill, K Imada, S Leigh, N

Richner, T Shi and H R Wilson, to be submitted to Nucl. Fusion (2022)

Presenter e-mail: jmenard@pppl.gov

PPPL Low-Aspect-Ratio Next-Step Facility Scoping Studies

J.E. Menard1, T. Brown1, R. Maingi1, W. Guttenfelder1, C. Swanson1, C. Rana1, Y. Zhai1

1Princeton Plasma Physics Laboratory

Compact tokamaks have been proposed as a means of potentially reducing the capital cost

of a fusion pilot plant (FPP). However, compact tokamak FPPs with higher magnetic field and

reduced major radius and surface area face the challenge of integrating high core confinement,

plasma pressure, and high divertor parallel heat flux and wall loading. This integration is

sufficiently challenging that construction and operation of a Sustained High Power Density

(SHPD) / EXhaust and Confinement Integration Tokamak Experiment (EXCITE) facility has

been proposed by the U.S. community to close this integration gap. This presentation will

review SHPD/EXCITE in the context of present and planned facilities and describe recent

systems studies and design activities for superconducting steady-state FPP and EXCITE

facilities. Specific core-edge integration challenges including edge density operating windows

and choice of impurities for detached divertor operation, the potential to access turbulence-

driven broadening of the scrape-off-layer heat fluxes, and H-mode access and sustainment with

high core radiation fraction will be described as a function of facility size and aspect ratio A [1].

Low-A “spherical tokamaks” (STs) have potentially favorable properties for such facilities

including reduced toroidal field and central solenoid magnet mass and cost, enhanced energy

confinement through elevated pedestal pressure, and increased turbulence broadening of the

scrape-off-layer heat flux. A low-A fusion power plant preliminary concept leveraging

knowledge gained from the SHPD/EXCITE studies with aspect ratio near 2 and R = 4 to 4.5m

and capable of generating 200-500 MWe will also be described.

Acknowledgements

This research was supported by the U.S. Department of Energy (DOE) Contract No. DE-

AC02- 09CH11466 and the DOE Laboratory Directed Research and Development (LDRD)

program.

References

[1] J.E. Menard, B.A. Grierson, et al, Nucl. Fusion 62 (2022) 036026

Pre-conceptual Design of Liquid Li Plasma Facing Components in NSTX-U

R. Maingi1, E. Emdee1, R. J. Goldston1, A. Khodak1

1Princeton Plasma Physics Laboratory, 100 Stellarator Road, Princeton NJ 08540 USA

Liquid metal plasma-facing components present a transformative opportunity for

enhanced power exhaust and possibly improved energy confinement in future fusion devices,

e.g. fusion pilot plants. Central to their consideration for this use is near-term deployment and

evaluation in high power tokamaks. In this work we performed pre-conceptual design

calculations for two concepts in NSTX-U: a lithium vapor box (LVB)1, 2, and a fast-flowing

liquid lithium coolant in a capillary porous system (CPSF)3.

For the LVB design, we performed SOLPS-ITER calculations of a Li vapor box design

and compared in depth to a slot divertor design, for an NSTX-U equilibrium that can in principle

generate ~ 100 MW/m2 perpendicular heat flux on the PFCs4, 5. The goal was to obtain a peak

heat flux < 10 MW/m2. Design and sensitivity calculations showed this was achieved with an

upstream Li concentration typically less than 10%. When an additional deuterium gas puff was

added upstream and pumped at the target, the upstream Li concentration was typically less than

5%. In general the LVB was much more effective at retaining Li to the divertor than the slot

divertor for the same cooling requirements. Unlike the slot, the LVB could cool the target heat

flux to <10 MW/m2 without a fuel puff and without significant deterioration of the midplane

electron temperature. With a fuel puff, the slot and LVB results become more similar, though

the LVB still has divertor impurity retention advantages.

For the CPSF system design, we performed analysis of heat transfer and MHD drag from

flowing Li PFCs with COMSOL and ANSYS, with a variety of assumed heat flux profile

shapes6. In general 10 MW/m2 perpendicular heat flux can be exhausted in NSTX-U with flow

speeds ~ 5-10 m/s, which is driven by imposed electrical current into the Li that forces radial

Li movement, while maintaining surface temperature < 450 C. In addition, allowing the surface

temperature to increase to 500-550 C reduces the required drive speed to 1-2 m/s. Under this

condition the CPSF system functions as an evaporator, which can be used as the Li source for

a LVB.

These two sets of calculations show a self-consistent operational window between liquid

Li flow, power dissipation, and upstream Li contamination from the plasma and fluid dynamics

perspectives, and bode well to initiate a conceptual design of a LVB system, with a CPSF

module in the bottom, for NSTX-U.

Acknowledgements

Sponsored in part by U.S. DoE contract DE-AC02- 09CH11466.

References

[1] R. J. Goldston et al., 2016 Phys. Scr. T167 014017

[2] R. J. Goldston et al., 2017 Nuclear Materials and Energy 12 1118

[3] A. Khodak and R. Maingi, 2021 Nuclear Materials and Energy 26 100935

[4] E. Emdee, "Modeling and Physics Design of a Lithium Vapor Box Divertor", 2022 Ph. D.

Thesis, Princeton Univ.

[5] E. D. Emdee et al., 2021 Nuclear Materials and Energy 27

[6] A. Khodak and R. Maingi, 2022 Phys. Plasmas 29

*tanyi@sunist.org

Progress of the SUNIST-2 spherical tokamak

TAN Yi1*, GAO Zhe1 and the SUNIST group1

1Department of Engineering Physics, Tsinghua University, Beijing 10084, China

The SUNIST-2 spherical tokamak (R/a: 0.525/0.325 m, BT: <1 T, IP: < 500 kA) is the

successor of SUNIST. The major objective of SUNIST-2 is to study the properties of spherical

torus plasmas with higher magnetic field and different shapes. Lots of minor modifications were

applied to the vacuum vessel, the central cylinder, the internal coils and the compensation coils

after the 2019 version of initial engineering design. Additional configurations such as negative

triangularity were analyzed and designed for SUNIST-2. Due to multiple reasons, the first

plasma of SUNIST-2 has delayed about two years, but the progress has been accelerated very

much recently. Xi’an will be the first location of SUNIST-2, where a building with 2400 square

meters has been renovated for it. SUNIST-2 is being assembled in the building, as well as the

development and the final test of power supplies, control and data acquisition systems, basic

diagnostics. When the workshop begins, SUNIST-2 is expected to be assembled.

Acknowledgements

This work is supported by the National Natural Science Foundation of China under Grant

No. 12175113, 11827810, 11875177, and12105189, International Atomic Energy Agency

Research under Grant No. 22733.

Properties of anomalous ion heating in lower hybrid wave sustained

plasmas

K. Iwasaki1, A. Ejiri1, N. Tsujii1, K. Shinohara1, O. Watanabe1, S. Jang1, P. Yi1, Y. Ko1, Y.

Lin1, Y. Shirasawa1, T. Hidano1, Y. Tian1, F. Adachi1

1The University of Tokyo

In the TST-2 spherical tokamak device, non-inductive plasma current start-up experiment

by lower-hybrid waves (LHW) is conducted. Typical plasma current is 20 kA. The LHW can

couple with electrons, but not with ions due to the difference between the particle velocities and

the wave phase velocity. Thus, the LHW can directly heat electrons, not ions. Plasma current is

considered to be generated mainly by fast electrons, whose energy is up to about 100 keV.

Bulk electron temperature has a hollow profile, about 10 eV at the magnetic axis and about 50

eV near the last closed flux surface, and electron density is low (< 1×1018 m-3) and has a center

peaked profile. Power flow in LHW sustained plasmas is important to confirm whether the

LHW drives current efficiently. For that purpose, the ion temperature measurement and

investigation of the heating mechanism are necessary.

Ion temperature is obtained from the Doppler broadening of a line spectrum emitted from

impurity ions. In TST-2, the emission is measured by a visible spectrometer, which has a 16-

channel photomultiplier tube array, and its wavelength range is about 0.3 nm. The line of C2+

(CIII, 464.74nm) is mainly measured. By changing line of sights for each shot, eight radial

positions can be measured. Typical ion temperature is 10 eV, and the profiles are flat or hollow.

Power modulation experiments were conducted to investigate the ion response. The

modulation period was 6 ms and the LHW power decreases from 50 W to 0 W during a period.

Plasma size and density is kept constant. Since the emission from LHW plasma is weak, period

averaged signal is analyzed. Ion temperature decreases after LHW is OFF and increases after

LHW is ON near the plasma edge ( = 0.25, 0.49 m) and the response time is less than 3ms for

both cases. It does not change near the magnetic axis ( = 0.35 m). From this result, ion heating

seems to occur on the peripheral region. Electron temperature is higher than ions’ in any time

and any position, so power flow by collision is always from electrons to ions. Heat diffusion

equation, in which thermal diffusivity is estimated by neoclassical diffusion and heating term

is only collision, shows that ion temperature couldn’t be over 2 eV in equilibrium state. Since

the included cooling power is the minimum, the calculated temperature of 2 eV is the maximum

if no other heating term is assumed. Thus, there should be other mechanisms to heat ions to

explain the measured temperature of 10±3 eV.

In TST-2, Parametric Decay Instability (PDI) is often observed. LHW decays into two

waves in a PDI, and one wave is ion cyclotron quasi-mode, whose frequency increases with the

toroidal magnetic field strength. Ion cyclotron damping or ion Landau damping of a wave

generated by PDI is one possible mechanism for the ion heating.

yhwang@snu.ac.kr

Research in Versatile Experiment Spherical Torus (VEST) for the

Development of Compact Fusion Reactor

Y. S. Hwang1 and VEST team1

1Department of Nuclear Engineering, Seoul National University, Seoul, Korea

Compact fusion reactor is emerging as a strong candidate for the realization of Carbon

Neural 2050 with aggressive investments from private sectors. Spherical torus (ST) with its

high beta and high bootstrap current capability is a critical component to reduce reactor size for

earlier fusion commercialization. Center for ST magnetic fusion metaware research was

established to solve stability and confinement challenges in ST by developing a virtual ST

fusion reactor of Versatile Experiment Spherical Torus (VEST) [1] with 6-d full Particle-In-Cell

(PIC) code.

Recent ohmic operation of VEST (R = 0.4 m and a = 0.3 m) has been focused on high-

current, long-pulse operation that can ba critical for efficient beam heating from the neutral

beam injection (NBI) system (<15kV, <600kW). Vacuum and wall-conditioning systems are

improved to control base pressure and impurity influx, so plasma current and pulse length can

be increased. With limited volt seconds plasma currents of up to 180kA at toroidal magnetic

field of 0.18T are obtained, but they are still too low to keep beam-heated fast ions to be

confined. Plasma current needs to be increased further more for high beta operation with

efficient beam coupling.

To extend operating stability boundary, Internal Reconnection Event (IRE), one of the

relaxation events driven by internal magnetohydrodynamic (MHD) instabilities in ST, has been

studied intensively in VEST. A strongly MHD-correlated blob at the edge is found to play a

significant role in the onset and the strength of IREs, suggesting that boundary plasmas can

impact internally driven relaxation events via MHD-correlated edge phenomena. When IRE

occurs, acceleration and deceleration of impurity ion rotation as a global phenomenon are

observed following the well-known ion heating in magnetic reconnection, indicating that this

phenomenon may be due to a neoclassical toroidal viscosity torque rather than usual magnetic

reconnection. [2]

While a 6-d full PIC code is under development in energy-conserving, semi-implicit

schemes, extensive VEST upgrades are ongoing to provide databases for the virtual fusion

reactor as well as high performance operation. Main upgrade of coil power system is doubling

ohmic volt seconds with an appropriate poloidal field coil power upgrade, enabling sufficient

plasma currents. NBI heating power will be doubled as well for high beta experiments. Existing

and upcoming new diagnostics will be integrated as a full package for extracting tokamak

plasma information consistent with PIC simulation results. High field, low-collisionality

reactor-relevant regime will be studied with the virtual fusion reactor for the development of

compact fusion reactor.

Acknowledgements

This work was supported by National R&D Program through the National Research

Foundation of Korea (NRF) funded by Ministry of Science and ICT (NRF-

2021M3F7A1084418).

References

[1] K.J. Chung et al., Plasma Science and Technology 15, 244-251 (2013).

[2] Seongcheol Kim et al., Nucl. Fusion 61, 126011 (2021).

ST Reactor – new considerations.

M Gryaznevich for the Tokamak Energy team

Tokamak Energy, 173 Brook Drive, Milton Park, Abingdon, OX14 4SD UK

Increase in the pulse duration in Fusion devices coincides with performance degradation

and at present there are no steady-state tokamaks or stellarators, although relevant regimes have

been developed. There are several possible causes for the limitation of the pulse duration. Even

if technical limitations will be removed (heating and current drive systems, thermal loads on

magnets due to neutrons, thermal loads on divertor and other structures), there are other possible

causes:

dust,

flaking,

hot

spots;

they

may

cause

degradation

termination

discharge

any

moment

increase in recycling, timescales depend on the wall material and condition, from few

to tens of minutes

accumulation of chemistry in the wall, from tens of minutes to hour(s).

Some of these issues can be (and being) addressed on present long-pulse devices: EAST,

KSTAR, W7X and even on low-performance devices (HIDRA etc). Boronisation, Li

conditioning and other technologies also contribute to resolution of performance degradation

with pulse duration, but all these have not been proved in high performance discharges with

pulse duration relevant to those required in a reactor.

That is why most of recent reactor designes assume pulsed operations. Advantages of

pulsed reactors, including economics, have been discussed recently in detail in Segal et al, NF

2021, Frank et al, Ph. Plas. 2022 for conventional tokamaks, Gryaznevich et al, Plasma 2022

for ST. It was shown that ST reactors have even more advantages due to possibility of high

bootstrap current, stronger increase in the confinement with the toroidal field, good stability,

and lower requirements on the volt-second capability of the central solenoid for the plasma

current sustainment.

We will show examples of pulse limitations and discuss ongoing and planned exeriments.

Unless these issues are resolved, we can’t confirm feasibility of a steady-state Fusion reactor

and we are progressing in resolving these issues.

*michele.romanelli@tokamakenergy.co.uk

ST40: Advancing the Physics Basis of Spherical Tokamak Reactors

M. Romanelli1*, S. McNamara1, P. Buxton1, A. Sladkomedova1, J. Varje1, C. Marsden1,M.

Moscheni1, S. Janhunen1, M. Anastopoulos1, J. Bland1, M. Sertoli1, A. Dnestrovskii1, J.

Wood1,M. Gryaznevich1, S. Medvedev1, S. Kaye2, A. Diallo2, C. S. Chang2, M. Barnes3, Y.

Zhang3,S. Sharapov4 and the ST40 Team

1Tokamak Energy Ltd., 173 Brook Drive, Milton Park, Oxfordshire, OX14 4SD, UK

2Princeton Plasma Physics Laboratory, Princeton, NJ, USA

3Rudolf Peierls Centre for Theoretical Physics, University of Oxford, Oxford OX1 3NP, UK

4CCFE Culham Science Centre, OX143DB, UK

The spherical tokamak ST40 [1] built and operated by Tokamak Energy Ltd, has recently

demonstrated record ion temperatures above 100M K by operating at 0.5MA plasma

current,1.9 T magnetic field and 1.8MW neutral beam heating power [2]. On the way to the

100M K milestone, ST40 has also produced a wealth of experimental data that allowed for

further advances in the physics basis of spherical tokamaks. To make the most of these first

phases of operation of ST40, in parallel to standard diagnostic analysis methods, forward

models of the various diagnostic systems have been developed and integrated in a single

workflow together with modelling tools such as FIDASIM and TRANSP. This has enabled the

detection of unexpected errors in various diagnostic systems, correction of geometric

uncertainties of line- of-sight mapping, and, above all, has allowed to constrain the kinetic

profile shapes and their central values despite the absence of profile diagnostics. The above

analysis provided important data that helped to uncover the impact of the large population of

suprathermal ions from NBI on both MHD and microturbulence stabilization. Gyrokinetic

analysis has shown that energetic ions reduce turbulent ion transport in ST40, although

remaining above neoclassical levels. Moreover, the coupling between turbulence and MHD is

shown to lead to improved sawtooth control. The interplay between beam-driven Alfvénic

modes of sweeping frequency (chirping modes) and H-L and L-H transitions in the global

confinement of plasmas has been investigated together with the effect of beta and the ST

geometry on toroidal Alfvén instabilities and ITG turbulence. Injection of Ar and Ne has led to

improved confinement, confirming previous theoretical results [3]. Transport analysis has

allowed the assessment of transport models including TGLF, Bohm gyro-Bohm, CDBM and

CPTM. All models show predictions of the ion temperature close to the experimental values,

however they fail to reproduce the scaling of confinement with toroidal magnetic field observed

in STs [4]. The impact of the ST configuration on the width of the SOL is being studied with

the support of the XGC1 global gyrokinetic code.

References

[1] M. Gryaznevich, O. Asunta (2017) Fusion Eng. Des. 123 177-180

[2] S. McNamara et al (2022), submitted to PRL

[3] M. Romanelli et al (2011), Nuclear Fusion 51 (10), 103008

[4] P F Buxton et al (2019) Plasma Phys. Control. Fusion 61 035006

*paolo.micozzi@enea.it

Status of the PROTO-SPHERA experiment: evidences of plasma tori

production and sustainment and progress towards Phase-2

P. Micozzi1* and PROTO-SPHERA Group

1Centro Ricerche ENEA, Frascati C.P. 65, Rome, Italy

The PROTO-SPHERA experiment is based upon a new 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, that substitutes 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=10 kA longitudinal current level, was

successfully ultimate at the begin of 2018. During the spring/summer 2018 the kink

destabilization of the PROTO-SPHERA Plasma Centerpost and the plasma tori production has

already been preliminarily obtained. 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. Therefore, now

PROTO-SPHERA is definitely not a Spheromak, since no flux conserver is present around the

central spherical plasma.

Interesting results were obtained in 2020-2021 campaign, despite the fact that the Ie was

still on the 10 kA level. Clear evidence of nested magnetic surfaces inside the tori was obtained,

both with visible light images from a fast camera (3600 fps) located outside the new (transparent)

vacuum vessel and with a 3D reconstruction from 6 medium-speed cameras (600 fps). The Hall

equatorial probes, that were inserted on the back of the movable Langmuir probes, have

collected signals, which show that the tori observed in visible light are associated with a toroidal

current inside the plasma tori; the value of the toroidal current estimated from the Hall probes

agrees with the value that can be derived from the distance between the X-points, as observed

on the visible images. The injection of current inside the plasma tori from the central screw

pinch looks like a quasi-periodical sawtooth, a succession of relaxation oscillations whose

quasi-periods lasts from 1 to 10 msec.

The Phase-2 of PROTO-SPHERA, planned to start at the end of 2023 in a new

experimental location in Frascati, will increase the plasma centerpost current by a factor of 7.

Some technical problem arisen in the 2020-2021 campaign must still be solved in 2022, before

the reallocation of the machine, but surely a new hollow cathode design shall be necessary as

well as a new and more robust design of the divertor plates.

*alex.prokopyzyn@ukaea.uk

Steady-State Alpha-Particle Flux on the Walls of the Spherical Tokamak for

Energy Production

A. P. K. Prokopyszyn1*, H. J. C. Oliver1, K. G. McClements1, D. A. Ryan1

1United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon, Oxfordshire

OX14 3DB, UK;

The UK-based Spherical Tokamak for Energy Production (STEP) programme aims to

develop a prototype fusion reactor by 2040 [1]. The plasma (including fusion alpha-particles)

must be confined to protect the tokamak first wall from unacceptable damage. The melt limits

of a tokamak reactor's main chamber wall and divertor have been estimated to be ~1.3 MW m-2

and 10 MW m-2 respectively. Our aim in the work reported here has been to predict the alpha-

particle power flux on the first wall in conceptual ST prototype reactors with a range of

parameter values to inform the STEP design decision-making process. Alpha-particle orbits are

assumed to be determined solely by the equilibrium magnetic field, static three-dimensional (3D)

magnetic field perturbations due to toroidal field (TF) ripple and resonant magnetic perturbation

(RMP) coils designed to mitigate edge localised modes (ELMs), and Coulomb collisions. We

have varied the following parameters: the TF coil outer major radius (picture- frame coils are

assumed), the number of TF coils, the TF field, the plasma current, and the location, number

and current of the RMP coils. The alpha-particles are modelled using the Monte Carlo Lorentz

Orbit Code for Use in Stellarators and Tokamaks (LOCUST) [2], which solves for individual

kinetic markers representing Monte Carlo samples of the alpha-particle distribution function.

We track the markers from birth until they collide with the wall or thermalise. Using these data,

we calculate the energy flux distribution on the wall using the kernel density estimation method

described in [3]. In the absence of both ripple and RMPs, the peak flux on the wall is about 10-2

MW m-2 on the upper divertor. For 16 TF coils with a major radius of 7.25 m, the peak alpha-

particle energy flux is about 0.5 MW m-2 on the outer main chamber wall, which is unacceptably

high. However, the maximum flux falls exponentially with the TF coil outer radius before

saturating at about 7.75 m, dropping to an acceptable value of below 0.1 MW m-2. As in the case

of TF ripple, alpha-particle losses due to RMPs are very sensitive to the coil parameters. RMPs

with dominant toroidal mode number n = 3 have been calculated using the MARS-F code [4]

with coil sets designed for error field (EF) correction and resistive wall mode (RWM) control.

Coil currents of 91 kAt and 47 kAt have been assumed in these two cases (there are large

uncertainties in the currents that would be required to suppress ELMs). Taking the plasma

response into account, the peak alpha-particle energy fluxes in the main chamber (on the inner

wall this time) are found to be 0.1 MWm-2 in the EF case and 0.05 MWm-2 in the RWM case.

These figures increase rapidly with coil current, however, reaching 0.5 MWm-2 in the EF case

when the current is increased by a factor of three. If active ELM control is required in STEP, the

choice of RMP coil parameters will thus be constrained by the need to avoid unacceptable hot

spots on the first wall arising from alpha- particle losses.

Acknowledgements

This work has been funded by STEP, a UKAEA program to design and build a prototype fusion

energy plant and a path to commercial fusion.

References

[1] H.R. Wilson et al., STEP – on the pathway to fusion commercialization, Ch 8 of

Commercialising Fusion Energy, IOP Publishing (2020).

[2] S.H. Ward et al., Nucl. Fusion 61 (2021) 086029.

[3] Y.-C. Chen, Biostatistics and Epidemiology 1 (2017) 161

[4] Y.Q. Liu et al., Phys. Plasmas 7 (2000) 3681.

Studies of fast electron dynamics in rf start-up plasmas on the TST-2

spherical tokamak

N. Tsujii1, A. Ejiri1, K. Shinohara1, S. Jang1, O. Watanabe1, Y. Peng1, K. Iwasaki1, Y. Ko1, Y.

Lin1,T. Hidano1, Y. Shirasawa1, Y. Tian1, F. Adachi1 and C.P. Moeller2

1The University of Tokyo

2General Atomics

Non-inductive plasma current start-up is considered to be essential in a spherical tokamak

fusion reactor. Fully non-inductive tokamak start-up has been studied on TST-2 using the outer-

midplance launch and top launch lower-hybrid current drive systems. The top antenna is

predicted to double the current driven by the conventional outer-midplane antenna from current

drive analysis assuming zero electron orbit width. However, only 26 kA has been achieved with

the top antenna compared to 21 kA of the outer-midplane antenna experimentally. Hard x-ray

radiation was substantially greater during the top antenna pulse than the outer-midplane antenna

pulse even at the same plasma current. Parametric dependences of the measured hard x-ray

radiation indicate that fast electrons may be lost by orbit expansion towards the low-field side

due to parallel electron acceleration that occurs mostly on the high-field side of the magnetic

axis in the top launch scenario.

Plasma current of the rf start-up plasmas is believed to be dominated by the non-thermal

fast electron current. Equilibrium reconstruction based on extended MHD that includes

collisionless kinetic electrons was developed to describe such a nonthermal plasma more

accurately. By using the radial electron density profile measured by the tangential Thomson

scattering diagnostic in addition to the standard magnetics, an equilibrium that better matches

the experimental measurements was reconstructed. The current drive prediction for the outer-

midplane launch scenario agreed well when the magnetic equilibrium from the extended MHD

reconstruction was used instead of the standard MHD reconstruction.

A new off-midplane launch antenna is being developed to overcome the limit of the present

lower-hybrid current drive system and achieve ramp-up up to higher current. The off-midplane

antenna was designed to fill the spectral gap of the presently installed midplane antenna and

boost the plasma current without causing orbit diffusion that leads to fast electron losses in the

top launch scenario. The new antenna has been fabricated and tested on bench. The antenna will

be installed in TST-2 this September. The results of the synergistic current drive experiment

will be reported if available.

Corresponding author email: yhwang@snu.ac.kr

Systemic error analysis of Thomson scattering system on VEST with rotational

Raman calibration

Jung-Hwa Kim1, Young-Gi Kim2, Doyeon Kim3, Jong-ha Lee2, and Y. S. Hwang1

1Seoul National University, Korea, 2Korea Institute of Fusion Energy, Korea, 3Institute for

Basic Science, Korea

Thomson scattering system is one of the main plasma diagnostic systems on Versatile

Experiment Spherical Torus (VEST). Recently, upgrade of the 1 kHz Thomson scattering

system has been successfully conducted to provide the time evolving plasma parameters in core

plasma, such as the rapidly changing radial profiles of the electron temperature and the density

[1]. Also, it allows for rotational Raman calibration, a more accurate method in the VEST

compared to Rayleigh calibration, by blocking the stray light. The Raman calibration can ignore

the effect from stray light and does not require any system modification. However, even small

systemic errors in the spectral responsivity measurement and the effective cross section of the

rotational Raman scattering may lead to a large change in the calibration result. By investigating

the signal anomaly of the Raman calibration with three filters, the causes of systemic errors are

analyzed and corrected. As a result, the calibration factors of the Raman calibration and their

uncertainties are obtained to measure the absolute electron density.

Acknowledgements

This work was supported by National R&D Program through the National Research

Foundation of Korea(NRF) funded by Ministry of Science and ICT(NRF-

2021M3F7A1084418)

References

[1] J.-H. Kim et al., Rev. Sci. Instrum. 92, 043549 (2021).

The New PEGASUS-III Experiment

S.J. Diem, M.W. Aslin, M.W. Bongard, M.T. Borchardt, A.L. Ferris, R.J. Fonck, J.A.

Goetz,A.K. Keyhani, B.A. Kujak-Ford, B.T. Lewicki, M.D. Nornberg, J.K. Peery, C. Pierren,

J.A. Reusch, A.T. Rhodes, C. Rodriguez Sanchez, C.E. Schaefer, A.C. Sontag, T.N. Tierney,

J.D. Weberski, and G.R.Winz

University of Wisconsin-Madison, Madison, WI, USA

Minimizing or eliminating the need for induction from a central solenoid during startup,

ramp-up, and sustainment of a tokamak plasma is a critical challenge in magnetic fusion energy.

Solenoid-free startup and current drive techniques potentially simplify the cost and complexity

of fusion power-plant-class devices by reducing the technical requirements of, or the need for,

a central solenoid. The PEGASUS program 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). The

PEGASUS facility has been replaced with a new solenoid-free, low aspect ratio (A > 1.22, Ip <

0.3 MA, BT < 0.6 T, pulse length ~ 50-100 ms) facility called the PEGASUS-III Experiment. The

new project is focused on developing an understanding and comparative studies of several

innovative candidates for non-solenoidal tokamak startup techniques. It is equipped with: a new

LHI system capable of Ip ~ 0.3 MA; sustained and transient coaxial helicity injection (CHI)

systems; and a microwave system for initial electron Bernstein wave (EBW) and electron

cyclotron (EC) heating. Upgrades include: a new solenoid-free centerstack assembly to provide

a 4x increase in BT; an associated new high- stress TF coil system; upgraded power supplies to

enable increased pulse length and higher BT to study LHI scaling, confinement and current

evolution; expanded power systems to support increased power LHI; enhanced diagnostics; and

an expanded lab facility. Initial experiments will establish high-Ip LHI scenarios, followed by

deployment of transient CHI, evaluation of non-circular LHI injector hardware, modest

sustained CHI and low-power microwave injection studies. PEGASUS-III will provide key

enabling reactor relevant technology to directly test proposed plasma startup and ramp-up

scenarios envisioned for larger scale STs, investigating methods to synergistically improve the

target plasma for consequent bootstrap and NBI current sustainment.

Acknowledgements

Work supported by US DOE grants DE-SC0019008 and DE-SC0020402

Presenting and corresponding author: jsegado@us.es

The SMART Central Stack Design

J. Segado-Fernandez1* A. Mancini1,2, J. Garcia-Dominguez1, J. Ayllon-Guerola1,3, D. J. Cruz-

Zabala1, L.Velarde1, M. Garcia-Munoz1,2, E. Viezzer1,2, C. Navarro3, P. Vicente-Torres1

1Centro Nacional de Aceleradores (U. Seville, CSIC, J. de Andalucia), Seville, Spain

2Department of Atomic, Molecular and Nuclear Physics, University of Seville, Seville, Spain

3Department of Mechanical Engineering and Manufacturing, University of Seville, Seville,

Spain

The inner-radial compactness of Spherical Tokamaks (ST) offers interesting physics

properties [1]. However, it also represents a technical challenge due to the reduced space for the

Toroidal Field Coil (TFC) inner legs and the Central Solenoid (CS) [2-5]. The SMall Aspect

Ratio Tokamak (SMART) is a spherical tokamak being assembled at the University of Seville.

Its operation will be divided in three phases with toroidal fields 1 T, inductive plasma

currents  500 kA and a pulse length up to 500 ms. The main geometrical parameters are 

= 0.4 m, 0.25 m and a wide range of shaping configurations (aspect ratio, 1.4 < R/a < 3,

elongation, 3, and triangularity, -0.6 0.6) [6, 7]. Within this frame, SMART will be

able to explore Negative/Positive triangularity plasmas and alternative exhaust techniques. The

magnet system of the tokamak is composed by 12 Toroidal Field Coils (TFC), 8 Poloidal Field

(PF) coils and a Central Solenoid (CS) [8]. The SMART central stack has been designed to

operate up to phase 2 (=0.4 T with inductive  up to 200 kA) and it comprises the inner legs

of the TFC, surrounded by the CS. Its design suppose a challenge due to the high electrical

currents needed to the create the magnetic fields and the limited space available. The loads

expected are 1.5 MN of centering force and 6 kNm of twisting torque. This contribution presents

the mechanical design and the structural verification of the central stack. A multiphysics finite

element model has been developed for the verification.

Keywords

SMART, Tokamak, Spherical, Central-Stack, FEM, multiphysics, structural

References

[1] A. Sykes et al. Plasma Physics and Fusion Technology, 39, 247-260 (1997)

[2] V. Thompson et al., Fusion Engineering and Design, 86, 1398-1401 (2011)

[3] Y. Zhai et al., Fusion Science and Technology, 77, 658-675 (2021)

[4] V. Belvakov et al, 17th IEEE/NPSS Symposium Fusion Engineering, 1, 273-276 (1997)

[5] V. Minaev, 25th IAEA Fusion Energy Conference, FIP/P8-25 (2014)

[6] S. Doyle et al., Fusion Engineering and Design, 171, 112706 (2021)

[7] A. Mancini et al., Fusion Engineering and Design, 171, 112542 (2021)

[8] M. Agredano-Torres et al., Fusion Engineering and Design, 168, 112683 (2021)

*samanta.macera@gssi.it

Three-dimensional Magnetic Reconnections in PROTO-SPHERA

S. Macera1*, … and PROTO-SPHERA Group2

1GSSI, Viale Francesco Crispi 7, 67100 L’Aquila, Italy

2Centro Ricerche ENEA, Frascati C.P. 65, Rome, Italy

In the last eighty years, a lot of efforts have been made in Plasma Physics in order to

understand magnetic reconnection, a physical process consisting of a topological rearrangement

of magnetic field lines in a plasma in order to minimize the magnetic energy. Reconnection

occurs in regions of strong magnetic shear, that are regions in which the direction of the

magnetic field changes significantly over a short distance. The big amount of energy released

during the process is converted into kinetic energy through acceleration or heating of charged

particles in the plasma. Magnetic reconections seem to be the key process for self-organization

of fusion plasmas in devices as tokamaks, spheromaks and reversed field pinches. In this work,

the data obtained from the PROTO-SPHERA (PROTOtype of a Spherical Plasma for HElicity

Relaxation Assessment) experiment were analyzed. PROTO-SPHERA is an innovative

magnetic confinement plasma experiment for controlled thermonuclear fusion research, located

at the CR Enea in Frascati. This experiment is characterized by a simply connected magnetic

configuration composed of a spherical torus (ST) with closed flux surfaces and a toroidal

plasma current and a screw pinch (SP) fed by electrodes with open flux surfaces and plasma

electrode current. Magnetic reconnection is the crucial mechanism which traps plasma particles

inside a magnetic island. Since the system is three-dimensional, and thanks to its rotational

invariance, this magnetic island is actually a torus in which plasma is confined. The diagnostics

present are the following: Hall sensors, one Langmuir probe, six fast cameras with a resolution

of 640x480 pixels and 600 fps, displaced on the equatorial plane at 60 degrees apart, an Ultra-

Fast camera, with a resolution of 1024x1024 pixels and 3600 fps. By exploiting the data of the

probes and the images of the fast cameras, an attempt was made to characterize reconnection

phenomena. For what concernes the geometry and dynamics of the plasma, we find that plasma

is not perfectly axisymmetrical, and appears to have a more oblate shape than a sphere. The

plasma motions are mainly two: one of rotation around its own axis of symmetry (vertical axis,

passing through the center of the plasma column) and one of rigid rotation, consisting of a

rotation of the plasma center around the axis of symmetry.

From this analysis it results that the plasma exhibits dynamic behaviors guided in part by

the magnetic reconnections: it appears that reconnection cause the plasma to expand and tighten

in a systematic way. This behaviour is accompanied by phenomena of incoming and outgoing

current from the spherical torus, which remains for the entire duration of the plasma discharge.

Furthermore, putting together the temperature signals and images it results that luminosity and

temperature peaks correspond to localized deformations around the X-shaped points present in

a stable manner in the magnetic configuration of the experiment itself. However, the observed

time-scales do not seem to be compatible with the collisional and collisionless reconnection

models. This may depend on the fact that the diagnostics in use have a sampling frequency that

does not allow direct observation of the reconnection episodes, but allow only to study the

cumulative effects of these episodes on the plasma. The long time-scales observed could,

however, be linked to the intrinsic nature of the reconnections, which show a decidedly three-

dimensional phenomenology.

Email address of presenting author: kaye@pppl.gov

Transport Analysis of ST40 High Performance Discharges

S.M. Kaye1, M. Sertoli2, P. Buxton2, A. Dnestrovskii2, S. McNamara2, M. Romanelli2

1Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543, USA

2Tokamak Energy Ltd., Milton Park, Oxfordshire OX14 4SD, UK

The main mission element of the ST40 tokamak, owned and operated by Tokamak Energy,

Ltd., UK, is to attain fusion-relevant ion temperatures (100M Kelvin) in a small, compact

spherical tokamak. These temperature levels have been achieved at higher aspect ratio, but only

in much larger devices with very high heating power levels. ST40 has a major radius of 40 cm,

plasma currents up to 800 kA, toroidal fields up to 2.2 T, and it is heated by up to 2 MW of

deuterium neutral beam injection. 2.2 T is the highest toroidal field in a spherical tokamak to

date. Near central impurity ion temperatures in excess of 100M K (~8.6 keV) have recently

been achieved and reported in ST40 for both hydrogen and deuterium thermal plasmas.

Temperature and density profiles, and their uncertainties in these ST40 plasmas have been

inferred from experimental and modeling constraints in the absence of full profile

measurements. TRANSP has been used to analyze the transport properties of these high-

performance plasmas and confirm the hydrogenic temperature levels. While slightly lower than

the impurity ion temperature level, hydrogenic temperatures do reach and exceed the 100M K

level. Despite the high central ion temperatures, both ion and electron transport are large, and

the ion temperature exceeds neoclassical levels in the outer region of these plasmas. The

TRANSP results are being used as a basis for both gyrokinetic studies of the source of these

turbulence levels as well as for studying fast ion driven instability observations.

Acknowledgements

The work at Princeton Plasma Physics Laboratory was supported by the U.S. Department

of Energy CRADA NFE-19-07769.

*tanyi@sunist.org

Twenty years of research on the SUNIST spherical tokamak

TAN Yi1*, GAO Zhe1 and the SUNIST group

1Department of Engineering Physics, Tsinghua University, Beijing 10084, China

The Sino UNIted Spherical Tokamak (SUNIST) is the first spherical tokamak in China. It

got the first plasma with 20 kA plasma current in the first full discharge test in Nov. 2002. Since

then, many technical and scientific works were done in the twenty years of research.

Considering the limited ohmic flux of the solenoid, SUNIST has a 3D viton seal between

the toroidal sections of the vacuum vessel, which insulated the sections, but didn’t have any

leak and significantly lowered the difficulties of breakdown. The power supplies of all magnetic

field coils were upgraded at least one times and greatly improved toroidal field, plasma current,

pulse length from designed values by 80%, 120% and 100%. The control system of SUNIST

has been upgraded to be distributed and web-based. Authenticated people can control and run

it from PCs or mobile phones. Two types of high field side mini SMBI systems were developed

and showed much better fueling efficiency than gas puffing. The top and bottom feeds of the

central solenoid were blown up respectively due to fatigue introduced extra resistance after

more than 15 years of operation. However, they were quickly fixed within one week or even

within one day.

The diagnostics of SUNIST were upgraded many times too. Several sets of electrostatic

probes were installed, including an ultrafast reciprocating probe with speed up to 21 m/s. They

were used to study the edge or core turbulence and to measure parameters there. The magnetic

diagnostics in SUNIST consists of an internal floating magnetic probe and several magnetic

probe arrays in radial, poloidal and toroidal directions with sufficient effective area for

equilibrium reconstructions and frequency response up to 1 MHz. The radiation detection range

covered from several GHz (ECE / EBE), infrared (bolometer), visible (fast cameras, filter

scopes, doppler widening spectrums), vacuum ultra violet spectrum (impurities), soft X-rays

(bremsstrahlung, impurities) and hard X-rays (runaway electrons). The latest version of

interferometers in SUNIST works at about 100 GHz with time response up to 0.1 microseconds

and density resolution up to 1E-15 m-2.

Since the first plasma, the research topics of SUNIST are mainly on non-inductive startup

and current drive, turbulences, MHD activities, Alfvenic behaviors. The electron cyclotron

wave startup with 2.45 GHz microwaves often exhibits an oscillating behavior because of the

long wavelength, which was experimentally examined in SUNIST. The mode conversion from

ECW to EBW was carefully verified in the ohmic plasma. Edge and core turbulences were

extensively studied by various diagnostics. Tearing modes caused by the rapid ramp up rate of

plasma current were measured by the radial array of magnetic probes. Its effects on confinement

were checked. Alfvenic perturbations were often observed during minor disruptions of ohmic

discharges. Their features and interactions with runaway electrons were also studied.

After 20 years of operation, SUNIST will not be closed but become a test bed for new

techniques and mainly an educational device for undergraduate students in Tsinghua University.

*am.kudr95@gmail.com

Verification of toroidal field coil stress analysis based on the measurements

in experiments on tokamak GLOBUS-M2

D.I. Alekseev1, E.N. Bondarchuk1, A.A. Kavin1, A.M. Kudriavtseva1*, A.N. Labusov1,

M.V. Manzuk1, V.B. Minaev2, Yu.V. Petrov2, I.Yu. Rodin1, N.V. Sakharov2, O.G. Filatov1,

1JSC «NIIEFA», St. Petersburg, Russia

2Ioffe Institute, St. Petersburg, Russia

This paper presents the verification results of stress analysis of the Globus-M2 spherical

tokamak toroidal field coil (TFC) based on measurements carried out in experiments.

In a series of experiments with close to the maximum values of the plasma current (≈400

kA) and toroidal field (≈0.8 T), the TFC displacements in the toroidal direction under

electromagnetic (EM) loads relative to its natural position, using the Photron SA5 high-speed

camera, were measured. The measurement equipment was triggered from the tokamak

synchronizer. The objective of the Photron SA5 camera was focused on the upper part of the

outside TFC, where, according to preliminary calculations, the maximum displacements of the

TFC during the plasma discharges were expected. Video recording of the TFC displacements

in the experiments was saved with the high frequency (1000 FPS) in hardware for subsequent

comparison with the calculated data.

The EM loads on the 3D model of the Globus-M2 EM tokamak system were determined

from the currents of the TFC, poloidal coils and plasma measured in experiments.

These loads were used for the stress analysis of the Globus-M2 EM system in the ANSYS

software. For analysis, the finite element model of the ½ Globus-M2 EM system was used

which includes 8 TFC coils, poloidal coils, interblock and supporting structures.

The analysis results were compared with the experimental measurements of TFC out-of-

plane displacement and showed agreement within 20%. Therefore, the approach consisting of

two stages - the reconstruction of the EM loads and the determination of the mechanical

response of the structure, has been verified for the EM system of the Globus-M2 and can be

used for other tokamaks.

Keywords:

Spherical tokamak Globus-M2, toroidal field coil, electromagnetic loads, stress analysis.

Will the MHD model coupled with the plasma diagnostics data be able to

reconstruct out of reach core plasma parameters in Tokamak devices?

Tara Ahmadi1, Yasushi Ono1, Yunhan Cai2 and Hiroshi Tanabe1

1Graduate school of frontier sciences, University of Tokyo, Tokyo, Japan

2Graduate school of engineering, University of Tokyo, Tokyo, Japan

In this report, we introduce some results received from a novel computational tool enabling

the reconstruction of the plasma core parameters inside the Tokamak devices. Despite its

simplicity due to utilizing the MHD system of equations, this model is able to include data

received from the multiple plasma diagnostics to reconstruct the internal plasma parameters

which are out of reach of diagnostic systems because of the high temperature of plasma core.

In the current model, the data received from the magnetic diagnostic systems such as magnetic

probe arrays or flux loops are implemented in the 2D Hall-MHD model as some current

carrying points resembling the measuring systems[1].At this time, the EFIT - like models are

being used to give some information regarding the plasma shape, size , and position inside the

Tokamak devices. However, the EFIT model is based on the Grad- Shafranov equation when

the plasma is at the equilibrium state. In this case, the flow velocity is fully neglected, making

it impossible for the EFIT to provide any reliable information regarding the plasma core

parameters outside the plasma equilibrium. There is a crucial necessity to control the plasma

during its dynamic phase inside the Tokamak to improve the plasma stability, temperature , and

confinement.

The developed reconstruction tool is adaptive to various Tokamak device irrelevant of the

size and start-up method as the MHD system of equations governing the plasma-magnetic field

interactions globally and the data received from the plasma diagnostics can give a detailed view

of the plasma processes inside the Tokamak devices.

University of Tokyo TS-3U spherical Tokamak has been adopted as a test case to evaluate

the reconstruction tool. TS-3U utilizes the merging start-up to initialize plasma current and

heating. At present, there are various plasma diagnostics functioning during the merging start-

up such as a high-resolution magnetic probe system, Rogowski coils , and ion Doppler

tomography. As TS-3U is able to operate at low-temperature regimes, it is possible to compare

the reconstructed plasma core parameters with the directly measured internal values enabling a

detailed evaluation of the model. The detailed analysis during the plasma formation, merging ,

and equilibrium state revealed that the reconstruction error is directly related with spatial profile

of magnetic field, flow , and thermal pressure of the Tokamak plasmas. As an example, the 7%

reconstruction error at the mid-plane where the kinetic and 3D effects govern the process and

3% of errors at the plasma equilibrium.

The next study case is ST-40 spherical Tokamak plasmas with 1T magnetic field and keV

temperature in Tokamak energy Inc. which makes it impossible to estimate the plasma

parameters. Besides the presence of multiple active plasma diagnostics, there are a series of

external magnetic probe arrays and flux loops located on the vacuum vessel providing detailed

information regarding the external magnetic configuration. The reconstructed plasma boundary

corresponds well with the plasma boundary estimated from the visible light image measured by

CCD camera.

Acknowledgements

We would like to express our gratitude towards the ST-40 group especially Mikhail

Petrovich Gryaznevich for sharing their experimental data.

References

[1] Ahmadi, etal. (2022), Reconstruction of the Internal Magnetic Configuration of Two

Merging Spherical Tokamak Plasmas by External Probe Measurement and MHD

Simulation. IEEJ Trans Elec Electron Eng, 17: 1231-1233.

*mono@pppl.gov

X-I mode for efficient solenoid-free ECCD plasma current start-up for

fusion reactor

M. Ono1*, N. Bertelli1, V. Shevchenko2, H. Idei3, K. Hanada3

1Princeton Plasma Physics Laboratory, Princeton, NJ, 08543, USA

2Tokakam Energy, UK

3RIAM Kyushu University, Kasuga 816-8580, Japan

The elimination of OH solenoid maybe the most impactful design driver for the realization

of economical compact fusion tokamak reactor systems. However, this would require plasma

start-up and current ramp-up from zero plasma current and low electron temperature of sub-

keV to the full plasma current ~ 10 MA completely non-inductively. Among many solenoid-

free start-up approaches, the ECH approach is generally considered to be attractive from the

reactor compatible technology and well understood wave coupling, propagation, and absorption

physics. To address this challenge, an efficient solenoid-free start-up and ramp-up scenario

utilizing the low-field-side-launched extraordinary mode at fundamental electron cyclotron

harmonic frequency (LFSL X-I) is proposed to generate ~ 10 MA of plasma current for a

compact Sustained High Power Density (SHPD) tokamak facility [1] with relatively modest

ECH power of ~ 10 - 20 MW utilizing the presently available 170 GHz gyrotrons developed

for ITER [2]. The present concept of X-I ECCD is more than two orders of magnitude higher

in efficiency compared to other more conventional ECCD such as O-I, O-II, or X-II (ordinary

mode at Ωe, 2Ωe, extraordinary mode at 2Ωe) in the plasma start-up parameters in the reactor-

grade toroidal field regimes. The wave accessibility density limit improves significantly at

higher BT0 as accessible density goes up as square of wave frequency f^2 to operate in a

reasonable plasma density of ~ 0.5 – 1.0 x 10^19/m^3 with a very high ECH/ECCD efficiency

at 170 GHz for BT0 ~ 5- 6 T. The high ECCD efficiency is due to the strong wave-particle

interactions at the Doppler broadened  = Ωe resonance, due to the strong cyclotron interaction

of X-mode, together with absorption on uni-directional passing electrons which results from

the wave accessibility constraint. The quasi-linear analysis using CQL3D shows further

enhancements of ECCD efficiency over the linear calculations due to the generation of less

collisional higher energy tail electrons. The current drive efficiency remains high for a broad

range of launched parallel index of refraction nll0 (or injection angle) suggesting that a relatively

simple waveguide launcher could be used. Since the LFSL X-I ECCD region is well localized

near the = Ωe resonance region, it is also possible to control the current drive profile if

desirable by aiming the waveguide launchers on/off machine midplane. This may also lead to

more MHD stable current profiles needed for the advance tokamak operation. With the advent

of higher field ST facilities such as ST-40, the present concept could be tested in the near future.

Acknowledgements

This work was supported by US DOE Contract No. DE-AC02-09CH11466.

References

[1] J.E. Menard, et al, IAEA-CN-286/1013, IAEA (2021).

[2] M. Ono et al., Physical Review E 106, L023201 (2022).

Evolution of the D-T Spherical Torus (ST) toward the p-11B ST

Y.-K. M. Peng, Yuejiang Shi, Bing Liu, Wenjun Liu, Mingyuan Wang, Xiang Gu, Yunyang

Song, Takashi Maekawa, Akio Ishida, Jiaqi Dong, Kaiming Feng, Baoshan Yuan and

Minsheng Liu

Hebei Key Laboratory of Compact Fusion, Langfang 065001, China

ENN Science and Technology Development Co., Ltd, Langfang 065001, China

The evolution of the ST concept for D-T fusion and more recently toward p11B fusion are

reviewed, with regard to the following development stages:

1. It started with the theoretical concept of reducing A=R/a to 1.2–1.7 to improve plasma

stability and confinement while raising the utilization efficiency of B field.

2. After experimental verification of this concept at modest parameters (e.g., START),

major experiments aiming to reach multi-keV temperatures and >1019m-3 densities (e.g.,

MAST, MAST-U, NSTX, NSTX-U, GLOBUS-M2) were pursued.

3. To reduce device complexity, design concepts were introduced which feature

replaceable TF coil center post without central solenoid (CS), and modularization of

internal components for improved reliability and safety using remote-handling (e.g.,

VNS, CTF, Pilot Plant of varied plasma fusion power amplification Q).

4. DEMO plans such as UKAEA’s STEP program and Tokamak Energy’s ST fusion power

development program, relying on the HTSC magnets, began more recently.

5. To simplify design and enhance performance, experiments on CS-less current drive, to

raise ion energy confinement toward the neoclassical levels, and to test the possibility

of near-100% bootstrap current (requiring p/A~1) have been proposed.

6. Freed from neutrons and tritium, a p-11B ST could take advantage of the

abovementioned features, in exchange for the requirements of >100keV ion

temperatures and much improved confinement times beyond the ITER and ST scaling

laws.

7. To begin the pursuit toward these long-term goals, the p-11B ST R&D at ENN entered

the first two R&D stages by initiating the EXL-50 experiment where an energetic

electron component was introduced for the first time.

8. The research has seen progress [1] and challenges, in topics such as CS-free startup and

sustainment; multi-fluid rotating plasma equilibrium with hot energetic electrons and

colder ions; heating and fueling of the main plasma at a distance; confinement of the

main plasma when separated from the outer wall material; and dispersal of heat and

particle fluxes over wide areas of plasma-facing surfaces.

From a common plasma physics foundation of magnetic confinement, these results have

pointed to varied physics mechanisms relative to those commonly observed in larger tokamaks

and ST’s. The EXL-50 experiment is planned to continue through Summer of 2023, after adding

a 50-GHz gyrotron delivering 200kW source power, a tangential NBI heating system, and

improved plasma fueling and control systems. Details of the issues, challenges and likely

solutions will be presented.

Reference

[1] Yuejiang Shi et al, Nucl. Fusion 62 (2022) 086047.