NATIONAL INSTITUTE FOR FUSION SCIENCE
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Satoshi Ohdachi, Editor of the Post-CUP Workshop proceeding( Received - 2020/01/29
NIFS-PROC-116)
Feb. 21, 2020

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Contents
PrefaceTime Table
Proceedings
CFETR diagnostics design................................................................................................................ 001
HU Liqun
Recent progress of transport and confinement research on EAST ................................................... 005
LIU Haiqing
Development of advanced tokamak scenarios in HL-2A plasma ..................................................... 009
LIU YiResearch activities on the SUNIST spherical tokamak .................................................................... 020
GAO ZheExperimental results of disruption mitigation with SMBI and MGI on HL-2A............................... 026
DONG Yunbo
Detection of magnetic islands using imaging diagnostics ................................................................ 030
OHDACHI Satoshi
Nonlinear MHD simulation of plasma termination events in stellarators ........................................ 034
SUZUKI Yasuhiro
Overview of the LHD neutron diagnostics ....................................................................................... 038
ISOBE Mitsutaka
1 MeV triton orbit analysis in Experimental Advanced Superconducting Tokamak........................ 044
OGAWA Kunihiro
Spectroscopic study of impurity behavior in the edge stochastic magnetic field layer of Large Helical
Device ............................................................................................................................................... 048
OISHI Tetsutarou
EMC3-EIRENE modeling of impurity transport in neon-seeded EAST plasma.............................. 052
LIU B.
A particular method for measuring low edge ion temperature based on VUV spectroscopy........... 056
MORITA Shigeru
ICRF heating experiments on EAST ................................................................................................ 061
ZHANG Xinjun
Conceptual design of CFETR ECRH system ................................................................................... 065
WANG Xiaojie,ISimulation of energetic particle driven geodesic acoustic mode channeling in LHD using MEGA code........................................................................................................................................................... 070
WANG Hao
Evaluation method of effective ion charge in the LHD plasma........................................................ 076
KAWAMOTO Yasuko
Observation of tungsten spectra from W5+ to W45+ in EAST........................................................ 081
ZHANG Ling
Tungsten measurements at LHD using visible M1 lines .................................................................. 085
KATO Daiji
Wall recycling experiment in EAST ................................................................................................. 089
ASHIKAWA Naoko
Laser-based diagnostic of LIAS and LIBS for the plasma wall interaction on the EAST tokamak. 093
HU Zhenhua
The deuterium permeation behavior in W studied by gas driven permeation and plasma driven
permeation methods .......................................................................................................................... 097
ZHAO Mingzhong
Current status of NIFS-SWJTU joint project for CFQS................................................................... 103
SHIMIZU Akihiro
Current status of engineering aspect for CFQS .............................................................................. 109
KINOSHITA Shigeyoshi
Finite element analysis of elctromagnetic force and stress on CFQS modular coils and support structure........................................................................................................................................................... 118
XIONG Guozhen
Engineering analysis of vacuum vessel for CFQS............................................................................ 130
NAKAGAWA Sho
Micro-instabilities and turbulence transport in the CFQS ................................................................ 136
NAKATA Motoki




,,


II

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,,,

III

Program of the Workshop

24 July 2019
(Wednesday)
09:00 Registration open
ISOBE Mitsutaka
09:30

Session 1(Opening & device

09:40

overview)
Chair:(Deviceoverview&collaboration

CFETR diagnostics design

LIU Haiqing(ASIPP)

Recent progress of transport and confinement research

10:30

LIU Yi(SWIP)Development of advanced tokamak scenarios in HL-2A

11:10

Coffee Break 10:55-11:10
NAGASAKI
Recent experimental results from the Heliotron J
Kazunobu
helical device(Kyoto Univ.)
GAO ZheResearch activities on the SUNIST spherical tokamak(Tsinghua Univ.)

11:35

strategy)
Chair:

Liqun (ASIPP)
HU Liqun(ASIPP)

Opening remarks

10:05

OHDACHI Satoshi

Session 2(NIFS) & HU

12:00

on EAST
plasmaDONG Shaohua(ASIPP)

New cooperation opportunities in the thirteenth-fiveyear project time

MORITA Shigeru
Lunch Break12:25-13:50
Session 3(Control & MHD)
Chair: YE Minyou

13:50
14:15
14:40

DONG Yunbo(SWIP)

Experimental results of disruption mitigation with

OHDACHI
Satoshi(NIFS)SUZUKI
Yasuhiro(NIFS)

Detection

SMBI and MGI on HL-2A
of

magneticislands

using

imagingdiagnostics
Nonlinear MHD simulation of plasma termination
events in stellarators

Group Photo + Coffee Break 15:05-15:30
15:30
Session 4(Neutron & EPs)
Chair: LIU Yi

15:55
16:20
16:45

FAN Tieshuan(Peking Univ.)
ISOBE Mitsutaka(NIFS)MURAKAMI
Sadayoshi(Kyoto Univ.)
OGAWA
Kunihiro

Fusion neutron spectroscopy on magnetic confinement
devices
Overview of the LHD neutron diagnostics
Estimation of DD/DT fusion neutrons of LHD plasmas
by GNET
1 MeV triton orbit analysis in Experimental Advanced,9IV

(NIFS)

Superconducting Tokamak

25 July(Thursday)
Session 5(Edge plasmas)
Chair: LIU Haiqing

09:00
09:25
09:50

YE Minyou(USTC)The influence of magnetic flux expansion on the in-out

OISHI
Tetsutarou(NIFS)KAWAMURA
Gakushi(NIFS)

Spectroscopic study of impurity behavior in the edge

divertor asymmetry in snowflake configuration
stochastic magnetic field layer of Large Helical Device
EMC3-EIRENE modeling of impurity transport in neonseeded EAST plasma

Coffee Break 10:15-10:35

Session 6

10:35

ZHANG Xinjun(ASIPP)ICRF heating experiments on EAST

11:00

WANG Xiaojie(ASIPP)

Status of ECRH system on EAST

11:25

WANG Xianqu(SWJTU)

Fishbone-like modes excited by energetic particles in

11:50

WANG Hao(NIFS)

Simulation of energetic particle driven geodesic acoustic(Heating & EPs)
Chair:SUZUKI Yasuhiro

weak reversed shear plasmas

mode channeling in LHD using MEGA code

Lunch Break 12:15-13:45
MORITA
13:45

Shigeru(NIFS)KAWAMOTO

Session 7(Diagnostics)

14:10

Yasuko(NIFS)

Chair: HU Liqun

14:35
15:00

Session 8

15:45

(PWI)
16:10

A particular method for measuring low edge ion
temperature based on VUV spectroscopy
Evaluation method of effective ion charge in the LHD
plasmaZHANG Ling

Observation of tungsten spectra from W5+ to W45+ in(ASIPP)EASTKATO Daiji(NIFS)

Tungsten measurements at LHD using visible M1 lines

Coffee Break 15:25-15:45
ASHIKAWA
Naoko
Wall recycling experiment in EAST
(NIFS)DING Fang(ASIPP)

ELM-induced W source in EAST divertor
9



V

Chair: KATO Daiji

Laser-based diagnostic of LIAS and LIBS for the plasma

16:35

HU Zhenhua(ASIPP)

17:00

ZHAO
Mingzhong(Shizuoka Univ.)

wall interatction on the EAST tokamak
The deuterium permeation behavior in W studied by gas
driven permeation and plasma driven permeation
methods

26 July(Friday)
09:00
09:20
Session 9(CFQS & closing)
Chair:ISOBE Mitsutaka

SHIMIZU
Akihiro(NIFS)KINOSHITA
Shigeyoshi(NIFS)

Current status of NIFS-SWJTU joint project for CFQS
Current status of engineering aspect for CFQS
Finite element analysis of elctromagnetic force and

09:40

XIONG Guozhen(SWJTU)

10:00

NAKAGAWA
Sho(NIFS)

10:20

NAKATA Motoki(NIFS)

Micro-instabilities and turbulence transport in the

10:40

ISOBE Mitsutaka(NIFS) & Hu
Liqun (ASIPP)

Summary reports, closing and others

stress on CFQS modular coils and support structure
Engineering analysis of vacuum vessel for CFQS

CFQS
(Proceedings with format, schedule after session etc.)

Lunch Break 11:00 - 13:00
13:00-17:00

Session 10
Discussionson

Post-CUP

andJSPS-CAS Bilateral
Joint

Research

Projects
Chair:ISOBE Mitsutaka
HU Liqun
OHDACHI Satoshi

Discussions on next workshop plans
Education of young scientists
Discussion on coordinator meeting
Discussion on annual report of the collaboration
Discussion on web site of the collaboration
Discussions on machine operation schedules
Discussions on new toroidal devices
Discussion on budget and administrative works"Some key issues in international cooperation among institutes"
by Ms. Shaohua DONG and Prof. Mitsutaka ISOBE et al.


9,


VI

CFETR Diagnostics Design
L.Q. Hu1, on behalf of CFETR diagnostic team1, 2
1

Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031, P.R. China
2Southwestern Institute of Physics, P.O. Box 432, Chengdu, Sichuan 610041, P.R. China

Abstract
The China Fusion Engineering Test Reactor (CFETR) is the next generation device aiming to bridge thegaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO) and help therealization of fusion energy in China. According to the roadmap of China magnetic confinement fusion
development, the primary mission of CFETR is to demonstrate fusion energy production in self-sufficientburning with different-advanced operation scenarios. CFETR will be proposed to be operated in two phases[1]. Phase I emphasizes on fusion power P fusion ı200 MW with fusion power gain Q=1-5, tritium breeding
ratio (TBR)>1.0, neutron irradiation effects ~10 dpa, and steady-state operation with a duty cycle of about
0.3–0.5. Phase II focuses on DEMO validation with P fusion ı1 GW and Q˚10, duty cycle ı0.5, neutron
irradiation effects in level of 50 dpa. Auxiliary heating and current drive systems with up to 80 MW including
N-NBI, ICRF, EC and LHCD will be employed to achieve steady-state advanced operation. Plasma
diagnostics aiming to machine protection, plasma control and burning physics research will provide a keyrole in plasma performance evaluation and study to help selection of operation regime and performance
optimization [2]. Due to very harsh nuclear environment in CFETR which is much worse than that in ITER,
most of diagnostics adopted by ITER cannot survived. Therefore new strategy and regulation need be setup
for reasonable selection of diagnostics, right development of diagnostics and relevant components to avoid
unnecessary waste of budget and human resource. Engineering design progress of CFETR diagnostics andthe first integration have been presented and discussed in this paper.
1. Introduction
CFETR task is to explore options for DEMO blankets and divertor solutions, and to discover a technicalsolution for licensing DEMO. As diagnostics in CFETR will in the severe nuclear environment with neutron
irradiation effects up to 50 dpa, which is several tens of times higher than that in ITER, consequentially arange of new phenomena coming from radiation-induced performance change and resulting radiation damage
must be considered and handled for physical properties intact of the material, including conductivity (RIC),
electrical degradation (RIED), electromotive force (RIEMF), thermo-electric sensitivity (RITES), absorption(RIA), nuclear heating, Radioluminescence (RL or RIE). Changes in other properties such as activation,transmutation and swelling have to be dealed with as well. Those will bring new and significant challenges
to diagnostic design and development, in aspects of structural metal material (physical property, mechanicstrength, volume stability, electrical and thermal conductivity, dielectric loss and permittivity), signal pickup
and transmission (optical properties, luminescence, coloration), electrical insulation (ceramics), signal
process (electronics, cable), and maintenance. Nuclear environment sets stringent demands on theengineering of the diagnostic system: neutron shielding, tritium containment, vacuum integrity, RH
-1-

1

001

compatibility. Moreover, existence of tritium blanket module and accessibility limited by ports availability
will greatly confine view and number of the diagnostics, diagnostics selection and engineering design will
be a compromising result among physics requirement, engineering realizability and environment available.
2. Diagnostics selection
Five years’ conceptual design of CFETR has produced fruitful results and enhanced understanding ofphysics design, integrated modeling of operation scenarios, advanced operation regime and engineering
requirement [1,2]. Published expected plasma performance is shown in Table 1, in which machine and plasma
parameters are I p =6-14 MA, B t0 =6.5 T, R 0 =7.2 m and a=2.2 m as top input parameters for engineering design.
The machine has 16 upper ports, 16 lower inclined ports and 6 equatorial ports, and 1.2m thick tritium blanket
module (TBM). Nearly 800 people from more than 30 research institutes and universities are involved in
CFETR engineering design, research and development (R&D) activities. Remote handling scheme andmaintenance strategy have been studied and setup. In vessel components maintenance is preferred for divertor,and divertor module will be removed through lower port during machine maintenance. TBM in multi-module
segment is changed and maintained through vertical port.
Table 1 CFETR plasma performance with 0D basic operation mode

3. Effects
3.
Effe
Ef
fect
fe
ctss of thect
the RMPRMP onon thethe energeticthener
nerrgegeti
ticc particle
ti
paart
rtic
icle
ic
le drivendririve
ven MHDven
MHD instabilities.
inst
in
nst
stab
abil
ab
illit
itie
ies
ies.
ie
(A-Steady State operation mode; B-Hybrid operation mode)
In past decade, ITER diagnostics have got significantly progress, however they are prepared only for 0.3
dpa nuclear circumstance and are not ready to be put into about service. In the other hand, port resource fordiagnostics in CFETR is relatively poor, and only two equatorial ports (EP) can be expected to diagnostics
even if more upper and lower ports are available (UP, LP). Therefore, plasma diagnostics in CFETR could
not be rich, and coverage range and spatial resolution would be very limited. Considering those issues, plasma
diagnostics in CFETR will be developed in two phases. In Phase 1 with Pf=0.5 GW (D-Phase 1), diagnostics
are designed and developed based on mature ones from ITER achievement together with a few prospective
and promising technology, the target of which will be to help choose most suitable operation region/scenario
from several candidates. In Phase 2 with Pf=1 GW or above (D-Phase 2), nuclear domain diagnostics plus
-2-

2

002

endurable and survival diagnostics from Phase 1 will be operated to approach target of plasma performance
optimization in given operation region. The missions of those diagnostics are of priority firstly machine
operation and protection, then plasma operation and control, and finally burning plasma understanding
mainly on investigation of Į particle behavior and performance. CFETR diagnostics should be kept inminimum but with reliable and achievable in techniques, small port resource consuming, long service time,high maintainability and nuclear radiation hardness. Individual diagnostic is better of multi-function to keep
enough redundancy. List of diagnostics in D-Phase 1 is given in Table 2, in which nearly half are port
dependent. They will provide plasma parameters of Te(r), Ti(r), Ne(r), V ȍ , impurity species and flux, fusion
products, etc., as much as possible from core to edge for burning control study. Nuclear techniques domain
diagnostics and few in Table 2 with very limited channels for Te/Ti/Ne, impurity, etc., and with high neutron
hardness and port independent could be employed for stable optimized plasma performance study in D-Phase
2.
Table 2 Diagnostics list in Phase 1
No.

Diagnostics

Objects
Ip, Vp, W, Ȍ, ǻ, ȕp, Halocurrent, Dsep; MHD
locking/precursor, etc.
Ne(r), j(r) (line averaged)

3

Magnetics (flux loop,
Rogowiskis/Diamagnet
ic/Mirnov/Sad. coils)
Polarimetry andInterferometry POINT
MW reflectmeter

Ne(r) (physics)

14

4

Metal Bolometer

15

5

LIDAR TS

Radiated power(total & distribution)Te(r), Ne(r)

6

CXRS

1
2

7
8
9
10
11

No.

Ti edge, Rotation (NBI)nHe/n(r), Zeff
XICS
Ti core, rotation (nonNBI), Heavy impurity
Visible spectrometer
Impurity species, Zeff,
H-mode&ELM, nHe/n(r)Fission chamber & Neutron yield, Pt, P(p&t)
microfission chamber
distribution, power loss,
stability and uniformity
Neutron camera
Neutron and Į source(Vertical & Radial )profile, flux&emissivity,
core fuel ratio and Ti(r)Gamma Ray
Core fuel ratio, REs, Įspectrometer GRS
particle performance,

12
13

Diagnostics
NeutronActivation System(NAS)
Energy-Res. Farad
Cup
LVIC camera

18

Visiblesurvey/imaging systemIRsurvey/imaging system
Divertor Erosion
monitor
Thermal couple

19

Dust monitor

16
17

20

21

TrtiumSurfaceconcentration
intensity, Hall
Dosage inspection
on outflow of themachine(solid,gas, liquid), T in
the air

Objects
2.45/14MeV,
absolute yield
Escaped Įparticle, T
H&ELM precursor,
pl.Equilibrium, REs
Stable operation andcontrol, shape
Wall
and
inner
component, rs
Vis.
CCD,
IR,
Erosion rate
Tmax, surface andinner components
Collection tube
HS CCD
Contamination, n/Ȗ
dosage rate, Dosage;
distribu.
Radioactivity andSpecific activity inoutflow, T intensity,Įȕsurface
contamination

3. Diagnostics integration
Similar to diagnostics integration scheme on ITER, the first round of integration on CFETR has been
carried out and port layout is shown in Fig.1. Equatorial port (EP) in size of 3071 mm(W)×2860
mm(H)×8070 mm(L) is divided to three DSM (diagnostic sheilding module). Each DSM with size of 926mm(W)×2860 mm(H)×8070 mm(L) is full of labyrinth to stop nuclear radiation going outside straightly. Up
to now, all diagnostics in Table 2 can be settled down in the integration. Three IR and visible CCD cameras

-3-

3

003

cover most inner surface of the blanket and heating systems. Ten sets of m-NFC (micro-nuclear fission
chamber) uniformly monitor the neutron yield from the divertor (toroidal 3 sets), inner blanket wall in HFS(high field side, 2 sets) and LFS (low filed side, three sets), and cryostat (2 sets). Nine poloidal channels of
POINT (Polarimetry and Interferometry) will be adopted to measure electron density and current density
pofile, and a single toroidal channel will be ready for density feedback control. Three sets of MW (microwave
interfemeter) provide local and fast variation of elctron density (one on HFS, one on top edge, another one
on the external blanket). LIDAR Thomson scattering system will be used to get electron temperature profile.
Two sets of neutron camera, one in EP (RNC) and the other in UP (VNC), supply neutron and Į particalsource profile, and core fuel ratio. Twelve sets of Neutron Activation System (NAS) in two EPs provide
absolute local 2.45 MeV or 14 MeV neutron yield. Gamma Ray spectrometers (GRS) locate after neutral
camera offer confined Įparticle performance and runaway beam (REs) behavior. Six sets of energy-resolved
farad cup (ERFC, EP1&2, each with three in vertical) give lost Į particle and energetic particle message.
Corresponding to two co-NBI injection beamlines, one Charge exchange recombination spectroscopy(CXRS) and one combination of CXRS and Visible spectrometer provide plasma rotation profile, edge Ti,
He density and impuirty information. X-ray Imaging Crystal Spectrometer (XICS) can give core Ti, rotation
profile in non –NBI case and heavy impurity information. Besides, those diagnostics will make compensation
with each other to provide supplementary or backup for primary one. But, it should be noted that as detailed
design of individual diagnostic moves on, much crash among them and more view space requirment fromphysics may come, therefor relocation of some diagnostics are probably unavoidable.
4. Summary
CFETR engineering design has been in progress well and the first integration allows detailed diagnostics
design moving on. Load analyses of the structure and assessment of diagnostics function will be done andbased on detailed design it will help to confirm or revise integration design. In the meantime, evaluation of
redundancy and robustness, remote assembly and maintenance of the diagnostic are needed to be considered
together with design. R&D activities of key components, especially common issues, are suggested to go inparallel to support the engineering design.
Acknowledgments
This work is supported by the National
Magnetic Confinement Fusion Program ofChina under Contract No. 2017YFE0300503.
References[1] Y.X. Wan, J.G. Li, Y. Liu, et, al., Overview
of the present progress and activities on the
CFETR, Nucl. Fusion, 57 (2017)102009.
[2] Y.T. Song, S.T. Wu, J.G. Li, et al., Concept
Design of CFETR Tokamak Machine, IEEE

Figure 1. Port layout of the first round diagnostics

Trans. Plasma Sci. 42 (2014) 503

integration on CFETR
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4

004

Recent progress of RF heating plasma confinement research on EAST
Haiqing Liu1, EAST team and collaborator*
1Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, China, 230031
Abstract
In tokamak plasma, a weak or reversed magnetic shear configuration is an important candidate for the
future fusion reactor, it is helpful to build a hollow current profile, establish an internal transport barrier (ITB),
improve the plasma confinement, and realize the long-pulse discharge. In experiments, auxiliary heating
systems, including NBI (Neutral Beam Injection), ECRH (Electron cyclotron resonance heating), ICRH (Iconcyclotron resonance heating), and LHCD㸦Lower hybrid current drive), are used to control the plasma current
as the Ohmic current driving efficiency is limited, consequently, achieving the weak or reversed magnetic
shear configuration. An improved confinement plasma operation can be obtained with these heating systems.
On EAST (Experimental Advanced Superconducting Tokamak), both of the four heating systems are applied
to heat plasma. In recent experiments, optimized confinement plasma is obtained with the interaction between
LHCD and ECRH, the simulation and experiments results are in good agreement.
1. Introduction
Advanced tokamak operation scenarios being explored for future fusion reactors must have highconfinement and stability which are closely linked to the internal plasma current density distribution.
Candidates scenarios can be either a weak-shear or a reversed-shear plasmas as the current profile in bothplays an important role in achieving and maintaining confinement and stability[1].To obtain weak or reversed
shear discharges, plasma current must be driven off-axis combining various non-inductive techniques, suchas neutral beam current drive(NBCD), electron cyclotron current drive(ECCD) and lower-hybrid current
drive(LHCD). LHCD has proved to be one of the most effective ways to drive the plasma current so far [2].
Numerous experiments have been conducted on the subject of LHCD in various tokamak devices like JT60U[3, 4] and JET[2, 5, 6] and so on. The results have shown that LHCD is an effective way to achieve thecurrent distribution required for advanced, high-performance, discharge scenarios.
Since 2015, EAST has been equipped with all ITER-relevant auxiliary heating and current drive systems,
enabling the investigation of plasma profile control by coupling and integration of various sources.
Table 1summarizes the designed power level of auxiliary heating systems and the recently achieved injection
power on EAST. LHW(Lower Hybrid Wave) system is the steadiest and most effective way to drive plasma
current. Recent experiments on EAST have demonstrated access to broad plasma current profiles by
increasing the plasma density in fully non-inductive lower hybrid current-driven discharges[7].

-1-

5

005

Heating/CD
schemeDesigned/projected
power (MW)Injected power and duration

4 MW/CW

2.8 MW/6 s; 1.2 MW/410 s

6 MW/CW

3.5 MW/2 s;1.4 MW/101 s

12 MW/CW

3.8 MW/6.0 s; 0.8 MW/60 s

2–4 MW/10–100 s

3.2 MW/2.0 s; 0.4 MW/100 s

2–4 MW/10–100 s

2.6 MW/2.0 s

4 MW/100–1000 s

0.6 MW/5 s; 0.5 MW/101 s

2.45
GHz
LHW

4.6 GHz

ICRF

Cocurrent
CounterNBI

ECRH

current

Table 1 summary of heating systems on EAST
2. RF heating optimized plasma on EAST
1)

Plasma current driven by LHCD broadens as density increases

Fig.1 Discharge parameters for LHCD plasmas.
Lower hybrid wave on EAST plays a very important role in heating plasma, steady operation and high
-2-

6

006

efficiency makes it a candidate in EAST long-pulse plasma operation. In experiments, 4.6GHz lower hybrid
wave was set to close heating input power, but with the different plasma density, loop voltage decreases tonearly zero as LHW inputs, implying the plasma current was driven effectively by LHCD. The LHCD profile
is expected to become more off-axis with higher density, because radial penetration of the wave is reduced
at higher density, thus the wave is absorbed closer to the plasma edge. Therefore, with only LHCD andbootstrap as sources of plasma current, the total current profile should become broader at higher density.
Combing EFIT and POlorimeter-INTeferometer (POINT) measurement together, plasma current profile canbe obtained, with LHCD current and bootstrap current only, the current driven by LHCD can be estimatedwith J LHCD = J tot – J BS .

Fig.2 Selected radial profiles for the four discharges of figure 8 at t ~ 4.9 s. (a) Fit to electron density
measurements, (b ) reconstructed current density and calculated bootstrap current.
Results of the reconstructions are shown in figure 2, confirming that broader current profiles are obtained
at higher density. Note that the q-profile goes from monotonic at lower density to negative central magnetic
shear sustained in steady-state at higher density. Thus, these results show a sort of ‘high performance synergy’
between broad current profile and high density in LH-driven plasmas: on one hand, higher density is needed
for higher performance, and on the other hand, higher density naturally gives a broader current profile, which
also can improve performance.
2)

Interaction between LHW and ECH provides an approach to optimize current density profile
LHCD is an effective way to drive plasma current in tokamak, the interaction between LHCD and otherheating system is also an important topic in plasma research. On EAST, the interaction between LHW andECH was studied, results are shown in figure 3.

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Fig.3 (a) internal inductance decreases slightly after ECRH off, (b) simulated LHCD current with by
GENRAY and CQL3D codes
In the recent EAST experiment campaign, we observed a decrease in internal inductance in an electron
heating dominant (LHW+ECRH+ICRF) H-mode plasma after ECRH termination. Simulations using
GENRAY&CQL3D codes reproduced the flattening of LHCD profile after ECRH termination (note that
LHCD is the major current drive source). Detailed analyses showed that thH current profile peaking 㸦blue
color in Fig.3 b㸧was related to the interaction between LHW and ECRH.

Acknowledgements
This work was partly supported by the Post-CUP program. This work was supported in part by the
Collaborative Research Program of the Research Institute for Applied Mechanics, Kyushu University andwas also partially supported by the National Key R&D Program of China under Grant No. 2017YFE0301205.

References[1] T. Suzuki, et al. 2008, Nuclear Fusion, 48:045002[2] A. Ekedahl, et al.1998, Nuclear Fusion, 38:1397[3] S. Ide, et al. 1996, Plasma Physics and Controlled Fusion, 38:1645[4] S. Ide, et al. 2000, Nuclear Fusion, 40:445[5] D. Mazon, et al. 2003, Plasma Physics & Controlled Fusion, 45:47[6] D. Moreau, et al. 1991, Plasma Physics & Controlled Fusion, 33:1621[7] A. M. Garofalo, et al. 2017, Nuclear Fusion, 57

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Role of the MHD activities in triggering H-mode in HL-2A tokamak
Yi Liu1, J. Cheng1, M. Jiang1, Y. G. Li1, L. M. Yu1, Y. B. Dong1, D. Li1, L. Liu1, W. L.
Zhong1, J. Q. Xu1, Z. H. Huang1, X. Q. Ji1, S. D. Song1, D. L. Yu1, M. Xu1, Z. B. Shi1, O.
Pan1, Q. W. Yang1, X. T. Ding1, X. R. Duan1, Yong Liu1 and HL-2A team
1

Southwestern Institute of Physics, P. O. Box 432, Chengdu, Sichuan 610041, China

The impact of a low frequency MHD mode crash on triggering the H mode has beenstudied in detail on the HL-2A tokamak. The mode manifests fishbone characteristics with aprocession frequency f |14-19 kHz. The abrupt mode crash evokes substantial energy releasefrom the core to the plasma boundary and hence increases the edge pressure gradient andE r uB flow shear, which further suppresses turbulence and leads to confinement improvement
into the H-mode. Under the same NBI heating (~ 1 MW), the I-phase plasma transits into
H-mode with a rapid mode crash while it returns to the L-mode without the presence of themode in the I-phase. With increasing heating power by the ECRH added to the NBI, thekink-like mode disappears. The statistical result shows that with the MHD mode crash theheating power for accessing the H mode is significantly lower than that without the mode
crash. All these facts reveal that the MHD mode crash in the I-phase plays a critical role intrigging the IoH transition at marginal heating power. In addition, it has been found that withthe same NBI power heating, the magnitude of the mode (crash) increases with increasingplasma density, implying larger energy release being needed to access the H-mode for higher
density plasmas. The results may have significant implication on ITER operation scenarioswith tight margin of L-H threshold power.

1. Introduction
Accessing and maintaining H mode performance are of crucial importance for the ITER
operation [1]. At present, the most widely used power scaling law, Pscale [MW] = 0.049  QH
[1020m3]0.72  Bt [T]0.8  S[m2]0.94, is valid for standard magnetic configuration with favorable
ion ‫׏‬B drift (pointing towards X-point) [2]. This scaling law indicates that the three dominantdependences are plasma density ( QH ), toroidal magnetic field (Bt) and the plasma surface area
(S). However, ongoing experiments from many devices discovered that other variables may
also strongly affect the L-H transition threshold power. For example, the DIII-D experiments
demonstrated that the plasma rotation is an important parameter affecting the L-H threshold
power [3]. Investigations at ASDEX Upgrade revealed a decrease of the L-H power by about
25% for the plasmas after the transition from a full carbon wall to a full tungsten wall [4].
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Findings in JET [5], DIII-D [6], MAST [7] and C-Mod [8] pointed out that the divertor
magnetic geometry (e. g., the X-point height and the length of the divertor leg) has strong
influence on the L-H power. Recently the key role of edge ion heat flux in the L-H transition
has also been reported on AUG [9]. Up to now, the underlying physics behind theses hidden
parameters has not yet been identified. Further research should be centered on identifying the
key ingredients of the L-H transition in order to reduce the power threshold.
Considering the present ITER power capabilities in the range of 70 MW [1], thereduction of the L-H power threshold would have great impact on ITER plasma operation
scenarios. For the tight margin of ITER L-H threshold power, an improved understanding of
the L-H transition at reduced power is highly desirable. In this paper, we firstly report thefavorable role of a kink-like MHD mode crash in triggering the transition into the H-mode
with reduced heating power in the HL-2A tokamak.
The rest of the paper is organized as follows. The experimental setup is described in
section 2. The experimental results and discussions are presented in section 3 and a summaryis given in section 4.
2. Experimental setup
The H-mode experiments were performed on the HL-2A tokamak with a lower single-nulldivertor configuration and following parameters: R=1.65 m, a=0.4 m, Bt=1.4 T, Ip=170-180
kA, QH | (1.4-2.8)u1019 m-3, PNBI=0.8-1.0 MW and PECRH = 0.3-1.5 MW. For measuring edge
equilibrium density (ne), electron temperature (Te) and floating potential (Vf) as well as
fluctuating quantities, a four-step Langmuir probe array (connected in a triple probe model)
has been employed with a sampling rate of 1 MHz [10]. The probe array was installed on areciprocating manipulator with a scanning speed of 1.0 m/s. The radially scanned distance isabout 8.0 cm, covering both the plasma edge and the scrape-off layer (SOL) [11]. The radial
electric field Er is deduced from the radial derivative of the plasma potential Vp = Vf + 2.8Te(for D plasmas). The magnetic fluctuations were detected by Mirnov coils at a sampling rateof 1 MHz. In addition, a multi-channel ECE diagnostic was used to measure the electron
temperature profile and a soft x-ray (SXR) system, consisting of two pinhole cameras, tomeasure the soft x-ray emission intensity.
3. Experimental results
3.1 Evolution of plasma parameters across transitions
In the HL-2A tokamak, with a marginal NBI heating power (PNBI | 1.0 MW) and theplasma central line-averaged density ranged at QH =(1.52.8)u1019m-3, an intermediate phase(I-phase) has been routinely observed during the LoH transition. The I-phase is characterized
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by strong oscillations in the edge recycling (DD emission) signal, and usually has better
plasma confinement than in the L-mode. At HL-2A, in addition to a single LoIoH transition,
multiple LoIoLoIoH transitions also occur frequently. Figure 1 illustrates the typical
temporal evolution of plasma parameters, including DD emission, the central line-averaged
density, NBI heating power, magnetic fluctuations and its root-mean-square (RMS) level
during the time period of LoIoLoIoH in shot 19380. It is clear to see that before each
LoI (t1=747ms, t2=785 ms) or IoH (t3=824ms) transition a strong MHD mode is developed,
and it collapses rapidly just prior to the transition time. The MHD mode frequency is around1419 kHz. These results explicitly reveal that the MHD mode collapse plays a critical role intriggering the confinement improvement, i. e., the LoI or IoH transition. It is interesting to
notice that the central line-averaged density abruptly reduces at each MHD mode crash andthen gradually recovers. This reduction is associated with the variation of the density profile
induced by the mode crash, as shown in figure 3 (f). The I-phase dynamics have been widely
discussed in terms of turbulence-zonal flow interaction (predator-prey model) and thebifurcation model [12-20]. In the present paper, our focus is nevertheless put on investigating
physical processes of the MHD mode crash and its favorable role in inducing the IoH
transition at relatively low auxiliary heating power on the HL-2A tokamak.
3.2 Impact of the MHD mode crash on edge plasma parameters
In order to understand the underlying physics of the MHD mode collapse in triggeringthe LoI or IoH transition, we look first on the influence of the MHD mode crash on theevolution of edge plasma parameters in two types of discharges with the same NBI heating
power. One is the LoIoL, while the other is the LoIoH shot. Figure 2 plots the time
history of the DD signal, the central line-averaged density, magnetic fluctuations and their
RMS level, the edge radial electric field (Er), the edge pressure gradient (rPe) and theambient turbulence level in density fluctuations ( QaHUPV filtered in 20-100 kHz) during the
LoIoL (#19274) and LoIoH (#19380) transitions, respectively. For the measurements ofEr, rPe and QaHUPV , the probe position is located at 'r |8 mm inside the last closed flux
surface (LCFS). In both cases, the transition from the L to I phase is accompanied by an
abrupt crash of the MHD mode, as shown in figures 2(c1, d1) and (c2, d2). However, during
the I-phase, the MHD behaviors are quite different between these two type discharges. In caseof LoIoL there is nearly no MHD activity in the I-phase, whereas for the LoIoH a strong
MHD mode is developed during the I-period and it drastically collapses prior to the IoH
transition. The impact of the abrupt MHD mode crash on edge plasma parameters can be seenin figures 2 (e1, f1, g1) and (e2, f2, g2). For the discharge of LoIoL (left column), after thequick MHD collapse (or LoI transition) the strength of edge Er (~Er ~) and ~rPe~
gradually increase and then keep constant for a while during the I-period, along with
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reduction in ambient turbulence level ( QaHUPV ). Finally, all of these quantities return to the
L-mode level without presence of any large MHD mode. In contrast, for the IoH transition(right column) the rapid mode crash in the I-phase excites sharp increases in ~Er~ and~Pe~, and concurrently, a sudden drop in QaHUPV before entering into the H-mode.
Across the IoH transition, the influence of the quick MHD mode crash on theequilibrium profile has also been surveyed and depicted in figure 3. Plotted in figures 3(a)(d)are time evolutions of the DD signal, the RMS level in magnetic fluctuations together withmulti-chord-averaged density ( QHO ) and electron temperature (Te) measured by HCN and ECE
diagnostics at different radial locations. From figures 3 (b), one can see that during theI-phase an MHD mode is developed and it grows rapidly. At about t=507.3 ms, the mode
collapses and concurrently dramatic changes occur in the density and temperature signals. In
the inner region of the core plasma, both QHO and Te quickly reduce, as seen in green curves
in figures 3(c) and (d), while in the outer area both of them increase (see brown curves).
These results clearly show a release of plasma energy from the core to the edge region due tothe MHD mode crash. As such, both of the density and temperature profiles are alteredconsiderably from the peaked to flattened one (see figures 3 (e) and (f)). The change of the
profiles results in a significant increase in the edge pressure gradient and the radial electricfield, as shown by the sudden jump in the ~rPe~ and ~Er~ signals in figures 2(e2) and (f2).
Then, the enhanced EruB flow shear will suppress the micro-turbulence magnitude, as
indicated in figure 2(g2). To further verify the role of the EruB flow shearing on thedecorrelation of turbulence eddies, we have compared the time evolutions of the EruB flow
shear rate (JEuB) with the local turbulence decorrelation rate (JD) measured at 'r |8 mm byprobes across the IoH transition. The results are shown in figure 4. Here the JEuB is definedas JEuB =wEr/(wrBt) and the JD is calculated from the inverse of the autocorrelation time ofdensity fluctuations. The data are taken from 10 shots with similar magnitude in the MHD
mode oscillations. The time 't=0 is defined at the time when the edge radial electric field andpressure gradient start to jump up during the I-phase. The error bars are estimated from thestandard deviation about the mean over the 10 similar discharges. In the figure one can clearly
see that before the MHD mode collapse ('t<0) the JD is larger than JEuB during the I-phase,whereas after the mode crash ('t>0) the EruB shear rate quickly rises up and the local
turbulence decorrelation rate reduces significantly so that JEuB > JD, resulting in turbulence
suppression, and consequently, the transition from the I-phase into the H-mode. These results
are generally in agreement with the turbulence decorrelation theory for the L-H transition [21].
It should be pointed out that for many years the effect of sawtooth crash on triggering the L-H
transition has been widely observed [22-24]. However, the detailed physical process has not
yet been reported. In this work, we provide for the first time the clear evidence showing the
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physical connection between the energy release evoked by the MHD collapse and edge flow
shearing on turbulence, which eventually excites an improved confinement into the H-mode.
3.3 Properties of the MHD mode
In this section, our attention focuses on exploring the nature of the MHD mode prior to
the IoH transition. For this purpose, we have made an expanded time window in figures 5(a)(c) for the time history of the DD and magnetic fluctuation signals drawn in figures 2(a2)and (c2), along with the time evolution of the frequency spectrum of the magnetic fluctuations.
It can be seen that the MHD mode appears twice with a lifetime of 25 ms and a chirping
frequency of 1419 kHz before the IoH transition. The first mode grows and decays in amild way. But for the second one the mode amplitude increases dramatically at the late stage
and then crashes abruptly. The influence of the mode magnitude and decay rate on plasma
parameters will be further illustrated in section 3.4. To investigate the spatial location of themode, figure 5(d) compares the time evolution of soft x-ray (SXR) signals measured at threedifferent radial positions. It is found that the oscillation magnitude is quite high in the coreplasma and it quickly diminishes toward the edge region, suggesting that the MHD mode islocalized in the plasma core area. For further identifying the spatial structure of the mode, wehave made tomographical reconstruction of the SXR emission by using a Bayesian tomographymethod [25]. The Singular-Value-Decomposition has been implemented on these reconstructions
to abstract the perturbation modes from the total emission. Figure 6(a) plots the temporal

evolution of the Mirnov signal before IoH transition, The graphs below (Figs. 6(b) and (c))
show the reconstructed mode structures at the time of “t” prior to the mode crash. One canclearly see that this mode has a dominant structure of m/n=1/1 and a harmonic m/n=2/2,
which rotates around the q=1 surface. Similar structures have been found for all of theperturbative MHD modes depicted in figures 1 and 2 in the L and I phases.
Further studies reveal that this mode is associated with the energetic-particle-driven
fishbone, whose frequency is close to the procession frequency of trapped ions. As shown infigure 5(c), the experimentally observed mode frequency is about 14-19 kHz. The procession
frequency of trapped ions is estimated by fpro = qEH(k,s)(1+0(H))/eBtRr, where q, E, H and s
represents the safety factor, the ion energy, inverse aspect-ratio and magnetic shear,
respectively. The H(k,s) is the geometrical factor [26] . For the HL-2A discharge parameters,
the Bt is 1.4 T and the pitch angle X///X | 0.15. Assuming that the energetic particles are
attributed to the NBI energy (ion source 40 keV), the procession frequency of theenergetic-particle-driven fishbone is about fpro |13 kHz near the q=1 surface, as marked by thecross symbol in figure 6 (d). Here, the effect of Doppler-shift frequency has been taken into
account. Using kI |1/qR | 0.6 m-1, VI=(30-50) km/s measured by the
charge-exchange-recombination spectroscopy (CXRS) [27], the Doppler-shift frequency
fDop=kIVI/2S |3-5 kHz. Thus, the mode frequency in the lab frame is flab=fpro+fDop |16-18
kHz, in good agreement with our observations. This result indicates that the MHD mode
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mentioned above is the trapped-ion-driven fishbone. In fact, at HL-2A another sort of
kink-type MHD mode, the so-called long-lived mode, has also been frequently observed andit plays a similar role in triggering the IoH transition.
3.4 Influence of the mode crash rate on plasma parameters and the IooH transition.
In figures 1 and 2, one may notice that the kink-type mode of fishbone is routinely
developed during the L-mode or I-phase for more than one time. However, only an abrupt
crash of the MHD mode with a large amplitude and a fast decay rate can trigger the LoI or
IoH transition. To understand this effect, we plot in figure 7 the time history of two fishbone
modes with different decay rates in the I-phase before the IoH transition, along with themulti-channel line-averaged density and electron temperature. One can see that the first mode
decays rather slowly (starting at t|817 ms from the climax) within ~ 3 ms, while the second
one collapses (starting at t|825.5 ms) much faster in a time scale of 0.5 ms, as shown in figure7(c) in the RMS level of the MHD modes. For the first slow crash at t|817 ms, there is nearlyno influence on the electron temperature and density at various radial locations. But for thesecond abrupt mode crash, both of the temperature and density are affected remarkably. These
results explicitly indicate that only the fast decay of the MHD mode, i. e., a sufficiently quick
release of plasma energy, can effectively modify the pressure profile, and hence, increase theedge pressure gradient and radial electric field, which further suppresses edge turbulence andleads to an improved confinement.
4. Summary
In this article, a kink-like MHD mode, routinely developing and abruptly crashing prior to
the LoI and IoH transitions with low heating power in the HL-2A tokamak, has beenreported. It has been found that the mode has fishbone characteristic with the procession
frequency f|1419 kHz and spatial mode structures m/n=1/1 and 2/2 around the q=1 surface.
The rapid mode crash evokes substantial energy release from the core to the plasma boundaryand hence increases the edge pressure gradient and EruB flow shear, which further suppresses
turbulence and leads to confinement improvement into the H-mode. Under the same NBI
heating, the I-phase plasma transits into H-mode with a rapid mode crash while it returns to
the L-mode without the presence of the mode in the I-phase. With increasing heating powerby the ECRH added to the NBI, the kink-like mode disappears and the plasma enters into the
H-mode with an overall rising in density and temperature profiles. Meanwhile, the lifetimeand magnitude of the I-phase oscillation are both reduced. The statistical result shows thatwith the MHD mode crash the heating power for accessing the H mode is significantly lower
than that without the mode crash. All these facts reveal that the MHD mode crash in theI-phase plays a critical role in trigging the IoH transition at marginal heating power. In
addition, it has been found that with the same NBI power heating, the magnitude of the mode(crash) increases with increasing plasma density. This observation suggests that larger energy
release is needed for accessing the H-mode regime for higher density plasmas, in accordance
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with the empiric scaling of the L-H threshold power.

Acknowledgment
This work is partly supported by Chinese National Fusion Project for ITER under grant Nos.
2015GB104000, and the National Natural Science Foundation of China under Grant Nos. 11375054,
11575057. It was also partially supported was partly supported by the Post-CUP program.

References:[1] Martin Y R et al 2008 J. Phys. Conf. Ser. 123 012033[2] Doyle E J et al 2007 Nucl. Fusion 47 S18[3] McKee G R et al 2009 Nucl. Fusion 49 115016[4] Ryter F et al 2013 Nucl Fus 53 113003[5] Andrew Y et al 2004 Plasma Phys. Control. Fusion 46 A87[6] Gohil P et al 2011 Nucl. Fusion 51 103020[7] Meyer H et al 2011 Nucl. Fusion 51 113011[8] Ma Y et al 2012 Plasma Phys. Control. Fusion 54 082002[9] Ryter F et al 2014 Nucl. Fusion 54 083003[10] Cheng J et al 2014 Nucl. Fusion 54 114004[11] Yan L W et al 2005 Rev. Sci. Instrum. 76 093506[12] Itoh S -I et al 1991 Phys. Rev. Lett. 67 2485
[13] Kim E J et al 2003 Phys. Rev. Lett. 90, 185006[14] Estrada T et al 2011 Phys. Rev. Lett. 107 245004[15] Conway G D et al 2011 Phys. Rev. Lett. 106 065001[16] Schmitz L et al 2012 Phys. Rev. Lett. 108 155002[17] Xu G S et al 2011 Phys. Rev. Lett. 107 125001[18] Kobayashi T et al 2013 Phys. Rev. Lett. 111 035002[19] Cheng J et al 2013 Phys. Rev. Lett. 110 265002[20] Shesterikov I et al 2013 Phys. Rev. Lett. 111 055006[21] Burrell K H et al 1997 Phys. Plasma 4 1499
[22] Tynan G R et al 1994 Phys. Plasma 1 3301
[23] Ido T et al 2002 Phys. Rev. Lett. 88 055006[24] Martin Y R et al 2004 Plasma Phys. Control. Fusion 46 A77[25] Li D et al et al 2013 Rev. Sci. Instrum. 84 083506[26] Yu L M et al 2013 Nucl. Fusion 53 053002[27] Wei Y L et al 2014 Rev. Sci. Instrum. 85 103503
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Fig. 1 Temporal evolutions of (a) DD emission; (b) the central line-averaged density; (c) NBI
heating power; (d) the magnetic fluctuation signal and (e) the RMS level of magnetic fluctuations
across the time period of LoIoLoIoH mode transitions.

Fig. 2 Temporal evolutions of DD emission (a1, a2); the central line-averaged density (b1, b2); themagnetic fluctuation signal (c1, c2); the RMS level of magnetic fluctuations (d1, d2); the radial
electric field (e1, e2); edge electron pressure gradient (f1, f2) and the RMS level of density
fluctuations (g1, g2) in the LoIoL (left column) and LoIoH (right column) discharges under
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the same NBI heating power.

Fig. 3 Left: time evolutions of (a) DD emission; (b) the RMS level of magnetic fluctuations; (c)the chord-averaged density detected by different HCN channels and (d) the electron temperaturedetected by multi-channel ECE diagnostic at various radial locations across the IoH transition.
Right: comparison of radial profiles of (e) electron temperature and (f) plasma density estimatedby Abel inversion from multi-channel HCN signals before (blue curves) and after (red curves) the
kink-like mode crash during the I-phase.

Fig. 4 Time-resolved evolutions of (a) the MHD mode magnitude indicated by the RMS level ofmagnetic fluctuations and (b) the EruB flow shear rate (red symbols) and turbulence
decorrelation rate (blue symbols) measured at 'r |8 mm by probes across the IoH transition.
The 't=0 is defined as the starting time of the MHD mode collapse. The data are averaged over
10 similar shots.
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Fig. 5 Temporal evolutions of (a) DD emission; (b) magnetic fluctuation signal and (c) the
frequency spectrum of magnetic fluctuations during the IoH transition. Shown in (d) is thecomparison of SXR signals measured at three different radial locations.

Fig. 6 (a) Temporal evolutions of the Mirnov signal along with contour-plots of thetomographically reconstructed SXR emission around the time of “t” before the mode collapse; (b)estimation of procession frequencies of trapped ions with different NBI energies at Bt =1.4T andthe pitch angle =0.15.
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Fig. 7 Temporal evolutions of (a) DD emission; (b) the frequency spectrum of magnetic
fluctuations; (c) the RMS level of magnetic fluctuations; (d) chord-averaged density detected by
HCN channels and (e) and electron temperature detected by multi-channel ECE at different
radial locations across the IoH transition.

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Recent Research Activities on the SUNIST Spherical Tokamak
Z. Gao1, and the SUNIST group1, 2
1

Department of Engineering Physics, Tsinghua University, Beijing 100084, China
2

Institute of Physics, Chinese Academy of Sciences, Beijing, China

Abstract
Recent research activities on the SUNIST spherical tokamak are introduced. Experimental researches arefocused on MHD behaviors during plasma current (Ip) ramp up /down phases of ohmic discharges, including
toroidal Alfven eigenmodes (TAEs) during minor disruption, internal reconnection events (IREs) in Ip rampdown phase, and tearing mode(TM) instabilities in Ip ramp-up phase.
1. Introduction
In tokamak discharges, the magneto-hydrodynamic (MHD) stability strongly relays on the profile of theplasma current and pressure. Therefore, various MHD activities appears in the ramp-up and ramp-down stageof plasma current.
During the ramp-down stage, the MHD activities have strong impacts on the operation safety of fusion
plasma device. The avoidance of major disruption and/or the emigration/suppression of runaway electrons
during disruption is of great importance in tokamak plasmas. However, major disruption seldom occurs in
spherical tokamak (ST), instead of minor disruptions or IRE relaxation. Therefore, ST provides a uniqueplatform for studying MHDs during Ip ramp-down phase. On the other hand, since the current torus in a
spherical tokamak has a smaller inductance than that of traditional tokamaks due to small major radius aswell as the low aspect ratio of the torus, the plasma current in a spherical tokamak will generally exhibit afaster current rise speed, which is accompanied with heavier skin effect and increases the potential for
detrimental MHD activities. Although the inherent MHD stability of spherical tokamaks reduces the tendency
of disruptions, fast confinement deteriorations have been experimentally observed during the fast current rise,
which deters the plasma from reaching the high beta regime.
The SUNIST device is a typical university-scale spherical tokamak. Its typical parameters are as follows:major radius/minor radius ~ 0.3m/0.23m; magnetic field at the axis < 0.3T, plasma current < 80kA; plasma
density ~ 1*1019m-3; pulse durations ~ 10~20ms. Recently, SUNIST focuses on MHD behaviors during IP
ramp up /down phases of ohmic discharges, including toroidal Alfven eigenmodes (TAEs) identification
during minor disruption and internal reconnection events (IREs) in Ip ramp-down phase, and tearing
mode(TM) instabilities in Ip ramp-up phase.
2. Toroidal Alfven eigenmode during minor disruption in Ip ramp-down phase
Toroidal Alfven eigenmodes (TAEs) excited in purely ohmically heated plasmas without any auxiliary
heating have been identified for the first time in the SUNIST spherical tokamak. [1] The experiments were
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performed in purely ohmic plasmas with Ip ~ 40 kA, toroidal magnetic field Bt ~ 0.15 T. The line-averaged
density was measured by a 94 GHz microwave interfer- ometer within the range of 0.2–6 *1018 m3. The
hard x-ray (HXR) was detected by CdZnTe with a narrow collimating aperture, which was used to estimate
the energy of REs. 
Typical discharge is shown in figure 1. A series of minor disruption was displayed in the Ip ramp-downphase. The MHD activities were observed during minor disruptions with a frequency range of 150–500 kHz.
Figure 2 shows that the mode structure analysis indicated the existence of both m/n~-3/-1 and -4/-1 harmonics,
propagating in the electron diamagnetic direction in the laboratory frame of reference. When the density ormagnetic field varied, the frequency also changed. Figure 3 shows that the observed mode frequency scales
linearly with the calculated TAE frequency. Also, TAE calculated by the GTAW code [2] agrees withexperimental results. These TAEs appear simultaneously with the generation of runaway electrons in thecurrent quench phase, accompanying with the density sweeping during the minor disruption. Recent in the
MAST, J-TEXT and HL-2A,[3] there were also observed some high-frequency Alfven-like MHD oscillations
during minor or major disruptions. All of these indicate that be some stories between Alfven instabilities anddisruption/RE. However, the driving mechanisms of these TAEs are still open.

Figure 1. Typical discharge with minor disruptions in the SUNIST

Figure 2. Toroidal and poloidal mode structure analysis

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Figure 3. Observed mode frequency scales linearly with the calculated TAE frequency
3. Internal reconnection events (IREs) in Ip ramp-down phase
In SUNIST, two typical phenomena were observed in Ip ramp-down phase. One is the minor disruption
described in last section. The other is so-called IREs, which is frequently observed in STs, and the IREs canbe observed as a collection of various MHD activities in different phases, which is shown in figure 4.[4] Thecross-bispectrum during the TQ phase shown in figure 5 indicated that the multi-field energy transfer fromelectrostatic to magnetic fluctuations. This behavior can also be seen in figure 6 as the time history of the
wavelet spectra. In the CI phase, it results in mode coupling in magnetic fluctuations, growing to a large scalewith a broad frequency spectrum, finally leading to the reconnection of local field lines, which is just the socalled IREs. However, in the CQ phase the magnetic fluctuations returns energy to electrostatic one and thesystem gradually comes back to a stationary state.
The mechanism displayed in the excitation of IREs in the SUNIST is different to the Hayashi’s
simulation picture.[5] Although the IREs was due to the mode coupling, the mode with high f magnetic
fluctuation grew due to nonlinear coupling in the SUNIST experiment rather than the linear instability
simulated in Hayashi’s simulation.

Figure 4. (left) typical discharge with IREs in the SUNIST; (right) three phases of IRE evolution:
thermal quench (TQ), current increment (CI) and current quench (CQ)

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Figure 5. (left) Cross-bispectrum during the TQ phase of IREs; (right) Time history of the wavelet spectra.
4. Tearing mode(TM) instabilities in Ip ramp-up phase.
Figure 6 displayed a typical plasma discharge with strong TM activities in the current ramp up stage. [6]TheTM activities started at 40.5 ms and weakened spontaneously at 43 ms with the measured MHD frequency from~10 kHz rising to ~20 kHz. Although the TM activities disappeared finally after the ramp up, it is clearly shownthat it caused a significant loss of plasma confinement. Consequently, the plasma exist with a rather low density.
The TM instabilities was carefully measured with the magnetic diagnostic system, which includes a 12channel toroidal probe array, a 14-channel poloidal probe array at the low field side, and a 30-channel radial
probe array with high spatial resolution. It is easily found that, the n αΫ1 mode was dominant throughout the
MHD stage for most of the discharges. the poloidal mode number appeared to be decreasing throughout the
MHD evolution, from |m| γ6 to |m| γ3. The radial mode structure is also of particular interest when considering
the position and size of the magnetic island. The radial distribution of BSș as measured by the radial probe arrayis illustrated in figure 7, where a conspicuous phase reversal layer could be observed. Such a phase reversal layer
extracts the approximate position of the magnetic island chain. Figure 7 also indicated that as Ip was rising, therational q surfaces generally shifted outward, which agree with the observation of the poloidal mode number todecrease during the current rise.
Figure 8 shows the correlation of density confinement, indicated by nemax, tothe radial position of themagnetic island chain, Rinv, and the amplitude of TM, ‫ۦ‬į Bș /Bș ۧmax. Although the variation in ‫ۦ‬į Bș /Bș ۧmax
could explain some observations, it failed to explain the results when Rinv was close to the edge. When Rinv
approached the edge, ‫ۦ‬į Bș /Bș ۧmax became weak while nemax still decreased. In fact, figure 8(b) shows therelation between nemax and Rinv, neglecting the variation in the perturbation level. It can be found that theconfinement deterioration effect could be approximately described by Rinv alone in both OF control and VF
control. It indicated that Rinv plays a dominant role in the confinement deterioration effect rather than themagnetic perturbation amplitude. Therefore, during the current rise, the confinement deterioration effect of theTM activity might be reduced effectively by controlling the magnetic island chain position. 

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Figure 6. Typical discharges with TM instabilities

Figure 7. The radial distribution of Bࡆșduring current rise. Calculated q positions also labelled.

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(a)

5.5

8
5

6

4.5

4

4
3.5

2
6

5

4

3

(b)
40

45

50

55

Figure 8. The corelation of density confinement to the amptitute of TM and the radial position of themagnetic island chain
5. Summary
MHD activities during the startup, ramp-up and ramp-down of plasma current have important impacts
on plasma performances as well as the operation safety of plasma device. Here the MHD behaviors in the
SUNIST spherical tokamak, including tearing modes during ramp-up phase, REs and TAEs during rampdown phase were investigated preliminary. It is hoped to explore an effective way to control the MHD activity
during plasma startup by multiple methods and following fast current ramp-up, and, meanwhile, to find a
ramp-down senarios to increase the resilence of plasma and to decrease the energy and percentage of runaway
electrons. Further understanding of these MHD behaviors are still necessary. The development of advanced
diagnostics with high spatial and temporal resolutions are needed as well.
Acknowledgements
This work was supported by NSFC under Grant Nos. 11827810 and 11875177, Science Challenge
Project with No. TZ2016005, and National Ten Thousand Talent Program. This work was also partly
supported by the Post-CUP program.

References[1] Yangqing Liu, Yi Tan, Zhe Gao, et al, Phys. Plasmas 23, 120706 (2016)[2] Y. J. Hu, G. Q. Li, N. N. Gorelenkov, et al, Phys. Plasmas 21, 052510 (2014).
[3] M. Gryaznevich, private communication (2016); Z. Y. Chen, private communication (2016); Y Liu,
IAEA FEC 2016 EX9-3[4] Song Chai㸪Yuhong Xu㸪Zhe Gao㸪et al, Phys. Plasmas 24, 032503 (2017).
[5] T. Hayashi, N. Mizuguchi, T. H. Watanabe, Y. TOdo, and T. Sato, Nucl. Fusion 40, 721 (2000).
[6] Zhong, H., Tan, Y., Gao, Z.,Nucl. Fusion 58, 126013 (2018)
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Experimental study of disruption mitigation by shattered pellet injector on HL-2A
Y.B. Dong, H. B. Xu, C.Y. Chen, J.M. Gao, Y.P. Zhang, X. Q. Ji, Yi Liu
Southwestern Institute of Physics, Chengdu, China
Abstract
A shattered pellet injector (SPI) based on in situ technology was developed in the HL-2A tokamak. In
this paper, the basic principle of pellet formation is introduced, t the mechanism of a SPI is described in detail,and test results are presented. Bench test experimental results show that reliable formation and accelerationof a large nitrogen pellet (diameter of Ң3.5 × 4 mm) were obtained. A double-impact tube was developed
and tested for effective and reliable shattering of the pellet.
1. Introduction
Plasma disruptions cannot be avoided during tokamak operations. Plasma disruptions damage thetokamak by heat loads deposited on plasma-facing components (PFCs). Additionally, large electromagnetic
faces acting on the device and runaway electrons will also impact and damage PFCs. So, it is necessary tolook for some effective method to mitigate plasma disruptions.[1–4]
Usually, pellet injection and massive gas injection (MGI) are used to mitigate damage caused by theplasma disruptions. Experimental results from fusion devices indicate that MGI can mitigate the heat loadand Halo current. But, the critical Rosenbluth density for the suppression of runaway electrons cannot be
obtained. However, shattered pellet injection as an effective strategy for mitigating plasma disruptions has
been investigated in DIII-D (Ref. 5). The experimental results show that shattered pellet injection can
improve penetration depth, enhance assimilation, and lead to higher plasma density when compared with
MGI with the same quantity of injection atoms. The results also show that thermal quench (TQ) and current
quench (CQ) are controlled in mitigated disruptions.[6] This cools the plasma to reduce the TQ heat load,
speed up the current decay, and reduce the halo currents. A shattered pellet injector (SPI) technique has beenshown to have several advantages relative to MGI. This observation indicates that a SPI should be a goodcandidate to mitigate plasma disruption in ITER.
In order to verify the effects of shattered pellet injection for plasma disruption mitigation, a SPI basedon in situ pellet formation technology has been developed in the HL-2A tokamak.
2. HL-2A Shattered Pellet Injector
A shattered pellet injector based on the in-situ technology was developed for HL-2A. Pellet is produced
by in-situ technology in cryogenic vacuum chamber. The technology is very mature. Compared with theextruder, it has a simpler structure and lower cost. Fig.1. is the shattered pellet injector (SPI) on the HL-2A
tokamak. The size of the shattered pellet injector is 2700mm by 945mm by1300mm.
The pellet injector is composed of two-stage vacuum differential pumping system, cryostat, gas vacuum
system, diagnostic system, etc. An advanced 2-stage GM-cycle refrigerator (Sumitomo Heavy Industries,
Model RDK-415D) is used. This concept has demonstrated a high reliability, an easy maintenance, a
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facilitated control of temperature and the turn-around time of an operation and a standstill time are shortened.
It is easier to meet the condition for the formation of a pellet.
Parker's Iota One

Compound molecular pumpcontroller

Guide
Cryogenic
temperature
Support frame

Gas-vacuum
Cryogenic
vacuum chamberTwo-stage vacuum differential pumping

Fig. 1. The shattered pellet injector on the HL-2A Tokamak
The pellet injector can provide frozen pellets of different species, such as, N2, Ar or Ne, etc. The
diameter and the length of pellet are 3.5mm and 4mm respectively. The pellet is accelerated by high-pressure
helium gas (0.5MPa-6MPa, pure 99.999 %).The pellet injector can supply one shattered pellet during plasma
disruption. The pellet velocity can be changed from 150 to 400 m/s under different propellant gas pressures.
The injection reliability is required to be greater than 90%.
The pellet will be shattered before entering the plasma by Double-Impact-Tube at the entrance to thetokamak vacuum chamber. The basic design of the double impact tube is shown in Fig.2. Shattered pellets
can achieve good effect by Double-Impact-Tube through the experiment. Normally, the time interval between
injection cycles is about8 min.
Pelle

Fig.2. Schematic of double-impact tube
3. Experimental setup
HL-2A is a middle-size tokamak (major radius R =1.65 m and minor radius r = 0.4 m) with a closed
divertor[7, 8]. The current direction in deuterium plasma discharges is counter-clockwise, but the toroidal
magnetic field is clockwise.

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Double-impact tube

SPI

Fig. 3. Schematic view of the shattered pellet injector installed on HL-2A(major radius R =1.65m, minor radius a = 0.4 m)
The pellet injection system is installed in a horizontal position. A series of diagnostics have been
employed on HL-2A for SPI experiments in order to evaluate the effectiveness of the SPI system. Being
constrained by the around space of HL-2A, the installation position of SPI is not reasonable, as shown in
Fig.3. Shattered pellet is tangent injected. It is expected that particle numbers of shattered pellet will be
reduced into plasma. Next, we will readjust the SPI installation location and implementing core injection.
4. Experimental results

Fig. 4 The comparison of disruption mitigation for SPI versus MGI with 1.0×1021 atoms of Ar injection.
Disruption mitigation with MGI and SPI is designed to rapidly inject a large amount of material into theplasma in order to radiate the thermal energy of the plasma. A typical example of such comparison on
identical scenarios with identical quantities of argon is shown in Fig.4. The typical plasma parameters during
these experiments were a toroidal field of 1.6T, plasma current of 160kA, line-averaged electron density ne= 0.8×1019 m-3, safety factor qa = 3.5.
The line-averaged electron density is on average about 1.4 times as large for SPI (3.15×1019 m-3) than it
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is for MGI (2.29×1019 m-3). We think that since SPI injects material in the form of small cryogenic fragments,
the resulting deposition is expected to be deeper by analogy with fueling pellets that have been observed tohave a deeper deposition than gas puffing thanks to shielding effects from the cold gas and plasma
surrounding the pellet that regulate the ablation process. The rate of current quench is about 80 MA/s by SPI,
but the rate of current quench is about 52 MA/s by MGI. The difference between these two methods is about
50%. The plasma current decay time is about 7.6ms. The plasma current decay time is about 7.2ms. The
results of the evaluation show that they have similar effects on current decay time. This indicates more
efficient particle assimilation for SPI, which could result in better thermal energy dissipation.
5. Discussion
A SPI based on in situ technology was developed in HL-2A. Reliable formation and acceleration of alarge nitrogen pellet (diameter of ‫׽‬3.5 × 4 mm) were demonstrated in laboratory testing. In addition, adouble-impact tube was developed and tested for effective and reliable shattering of the pellets. A SPI will
be installed in HL-2A to carry out related disruption mitigation research.
During the commissioning of the injector, we found that the given temperature of the heat sink wasmuch lower than the triple-point temperature of the nitrogen for producing a good pellet. Maybe this resultsfrom the large thermal resistance between the barrel and the heat sinks. To solve this problem, we will modify
the connection structure between the barrel and the heat sinks. Meanwhile, a thin-walled steel barrel (diameterof 3.5 × 0.7 mm) replaces the existing barrel. It was also found that the shattered pellet through the doubleimpact tube was better at the speed of more than 300 m/s than that at the speed of less than 200 m/s. In
addition, more tests should be done to study the performance of the Parker’s pulse valve.
Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by National Key
Research and Development Program of China (No. 2017YFE0301300), and was also partially supported by
the National Key Research and Development Program of China (No. 2017YFE0300405, 2017YFE0300504).

References[1]. M. SHIMADA et al., “Progress in the ITER Physics Basis—Chapter 1: Overview and Summary,” Nucl.
Fusion, 47, S1 (2007).
[2]. V. RICCARDO and JET EFDA CONTRIBUTORS, “Disruptions and Disruption Mitigation,” Plasma
Phys. Control. Fusion, 45, 12A (2003): A269.
[3]. S. K. COMBS et al., “Development and Testing of Plasma Disruption Mitigation Systems Applicable for
ITER,” International Atomic Energy Agency (IAEA) (IAEA-CN–197) (2012).
[4]. M. BAKHTIARI et al., “Fast Plasma Shutdown Scenarios in the JT-60U Tokamak Using Intense Mixed
Gas Puffing,” Nucl. Fusion, 42, 1197 (2002).
[5]. L. R. BAYLOR et al., “Disruption Mitigation Technology Concepts and Implications for ITER,” IEEE
Trans. Plasma Sci., 38, 419 (2010).
[6]. D. SHIRAKI et al., “Thermal Quench Mitigation and Current Quench Control by Injection of Mixed
Species Shattered Pellets in DIII-D,” Phys. Plasmas, 23, S128 (2016).

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Detection of Magnetic Islands Using Imaging Diagnostics
S. Ohdachi1, Y. Suzuki1, 2
1National Institute for Fusion Science, Toki 509-5292, Gifu, Japan

2

Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan

Abstract
Plasma response to the externally applied magnetic field is fairly complicated issue. The physics of theshielding and penetration of the external field has not been fully understood. A new kind of diagnostics using
two tangentially viewing SX camera system is proposed. Island detection from the reconstruction of 3D
emission profile related to the magnetic field structure with L1 regularization is discussed in this study [1].
1. Introduction
In order to sustain the high-performance plasma having an edge transport barrier, instabilities related tothe steep pressure gradient in the edge region is a critical issue. For example, the heat flux released by the
excitation of the edge localized mode (ELM) is hazardous for the divertor of the next-generation Tokamak
devices, such as the ITER. It is thus urgent to find a way to control the edge MHD activities. Though theresonant magnetic perturbation (RMP) is the most promising candidate to control the ELMs, the detailed
mechanism how the RMP affects the MHD activities, has not been fully understood. In order to study thechange of the behavior of the MHD instabilities, the determination of the equilibrium field is required.
If the externally applied magnetic field penetrate the plasma, the toroidal symmetry of the torus plasma
is broken as shown in Fig. 1. From the difference of the two images, we can evaluate the wave number andthe location of the magnetic island induced by the RMP [2]. However, it is not always possible to performthe forward modelling assuming a theoretical profile. If the direct reconstruction of the two camera data canbe made, many physical phenomena can be studied. In this article reconstruction using orthogonal basis
patterns and L1 regularization is discussed in detail.
2. Regression with orthogonal pattern and L1 / L2 regularization
Consider the problem that the local emission profile g is reconstructed from the line-integrated signals f. With
column vectors f and g, f can be represented as, ࢌ = ‫ ࢍ ܪ‬+ ݊‫݁ݏ݅݋‬. , using the geometrical relationshipexpressed as the matrix H. If the least square fitting method is used to solve, the following minimization
may give the solution g,g=‫݊݅݉݃ݎܣ‬| (݂௜ െ σ௜ ݄௜ ‫)ࢍ ڄ‬ଶ
(ࢍ)

(1)

where ݄௜ is the i-th row vector of the geometrical matrix H. If this solution is ill-behaved and unstable tothe noise, an additional penalty function is often introduced. When the penalty function is concerned with thetotal magnitude of the local emission, the function to be minimized becomes Eq. (2).

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ࢍ=

‫݊݅݉݃ݎܣ‬| (݂௜ െ σ௜ ݄௜ ‫)ࢍ ڄ‬ଶ + ߣ σ௜|݃௜ |ఈ (2)(ࢍ)

When Ƚ = 2, the penalty function is the squared Euclid norm of the vector ࢍ. This regularization scheme iscalled the L2 regularization (Ridge regression). When Ƚ = 1, the scheme is called the L1 regularization(least absolute shrinkage and selection operator (Lasso) regression). Recently the algorithm for L1
regularization is significantly improved and widely used for the regression problems in various fields research.

Fig. 1 Schematic view of the concept of the dual tangentially viewing SX camera system.
On the other hands, the series expansion method using global orthogonal basis patterns was a method
proposed in the early stage of fusion research [3]. This method was quite effective for reconstructing theemission profile, especially, of the Tokamak plasmas in circular cross section. Since this method is based on
the global patterns, it might be possible to reconstruct the emission profile from the information that is
acquired by the measurements of much limited coverage of the entire objective region.
In the series expansion methods, the emission profile g is expanded by a series of patterns ࢈࢏ asࢍ = σ୧ ߚ௜ ࢈࢏(3)that is, we have a series expansion of f with basis patterns ࢞࢏ = ‫࢏࢈ܪ‬ࢌ = ෍ ߚ௜ ࢞࢏(4)

୧

The relation can be written as ࢌ = ߚ ‫࢞ ڄ‬, where ࢞ = H࢈ .When the coefficients ߚ௜ are determined by themeasurement, by minimizing the ࢼ =‫݊݅݉݃ݎܣ‬| σ௜(݂௜ െ ࢼ ‫) ࢏࢞ ڄ‬ଶ + ߣ σ௜ ߚ௜ఈ
(ࢼ)

(5)

L1 regularization is used for solving Eq. (5). After the coefficients ߚ௜ is determined, the local emission
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profile g can be easily composed using Eq. (4).
3. Reconstruction of the tangentially viewing camera data
In this numerical test of reconstruction of the dual tangentially viewing camera system,
Fourier-Bessel type basis patterns having n = 0 components and n = 1 components are usedas,
Ȳ௟௠ (ߩ, ߠ, ߶) = ݁‫ܬ )ߠ݉݅(݌ݔ‬௠ (ߣ௟௠ ߩ) (݊ = 0)௠Ȳ௟ (ߩ, ߠ, ߶) = ݁‫ܬ )ߠ݉݅(݌ݔ‬௠ (ߣ௟௠ ߩ)exp(݅߶) (݊ = 1)

(5)Here, ‫ܬ‬௠ is the m-th Bessel function of the first kind where ߣ௟௠ is the l-th zero location of them-th Bessel function. ߠ and ߩ are the coordinate in poloidal direction and the averaged minor
radius, respectively. ߶ is the toroidal angle. Schematic image of this 3D basis patterns areshown in Fig. 2.

Fig. 2 Schematic image of the 3D basis patterns. n=0 example is shown in the left figure and n=1 in theright.
First, reconstruction having n=0 type emission profile are tested with one camera system [1].
The reconstructed pattern is quite similar to the assumed profile when the peaked emission
profile and hollow emission profile are assumed. For the reconstruction, the Scikit-learn(https://scikit-learn.org/) library using the LARS algorithm [4] is used for solution. The hyper
parameter is optimized by the cross validation method. In the case in which only the part (1/2~
1/4) of the tangential image is used, reconstruction is still fairly good. It is a great advantageof this type of method using global patterns since the viewing field in the fusion experiment isoften limited severely.
When the emission profile is superposition of the n=0 component and n=1 component, which
is closer to the real experiments, the performance of this reconstruction method is examined.
Fig. 3(b) is the tangential emission image where both the toroidally constant ( n=0 ) peaked
emission profile of Fig. 3 (a1) and the magnetic island like structure of the Fig. 3 (a2) (m/n =
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2/1 type) are assumed. Here, m and n are the poloidal and toroidal mode number, respectively.
Reconstruction of the n=0 and n=1 component are shown in Fig. 3 (c1) and Fig. 3 (c2). Though
the low amplitude noises can be seen in Fig. 3 (c2), fundamental characteristics of the n=1
mode is adequately reconstructed. L1 regularization is a key of this kind of reconstruction.
The noise of the reconstruction image using L2 regularization is much larger especially in thecore region of the image. One of the reasons of the merit of the L1 regularization is that lesscontributed components are neglected and not used. Most of the coefficient for L1
regularization is zero and only the coefficients for m=2 patterns are not zero and sufficiently
large.
In summary, it is shown that dual SX camera system can detect asymmetric component of
the SX emission. Penetration of the RMP or the locked-islands can be studied by using thereconstruction of the emission profile using L1 regularization.

Fig.3 Assumed profile (a) and tangentially projected synthetic image(b) and reconstructed image(c) are
shown.
Acknowledgements
This work was partly supported by the Post-CUP program. It is also supported by Japan / U. S.
Cooperation in Fusion Research and Development.

References[1] S. Ohdachi, Plasma Fusion Res., 14 (2019) 3402087[2] X. D. Du, et. al., Phys. Plasmas, 26 (2019) 042505[3] Y. Nagayama, Journal of Applied Physics, 62 (1987) 2702
[4] B. Efron, J. Am. Statist. Assoc., 99 (2004) 619
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Nonlinear MHD simulation of plasma termination events in stellarators
Y. Suzuki1,2, S. Futatani3, and J. Geiger4
1

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

2

Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan
3

Polytechnic University of Catalonia, Barcelona 08034, Spain

4 Max-Planck Institute for Plasma Physics, Greifswald D-17491, Germany

Abstract
two types of plasma termination events, which are MHD instabilities driven by the pressure gradient andplasma current, are studied, respectively. In the Large Helical Device (LHD) experiment, many MHD
instabilities are observed. Especially, if the peaked pressure profile was sustained by the pellet injection, acollapse event, so-called the core density collapse (CDC), was happen. In nonlinear MHD simulations, it isexpected the CDC is driven by the resistive ballooning mode. On the other hand, in Wendelstein 7-X (W7X) experiment, collapse events simultaneously happened in the plasma core, if the strongly localized plasma
current, which is ECCD for an example, is driven in the plasma core. If the rotational transform, ɇ, increases
by ECCD and crosses the rational of ɇ = 1, the magnetic island opens on ɇ = 1 rational surface. However,in nonlinear simulations, MHD event does not happen on ɇ = 1 rational surface. At the outer region of ɇ =1 rational surface, where is the negative shear region corresponding to the positive shear region in tokamaks,
the ballooning mode appears and destroys the flux surfaces completely. That is a mechanism to terminate theplasma.
1. Introduction
Nonlinear dynamics of plasma termination events in stellarators are studied using a 3D nonlinear MHD
simulation code. In this study, two types of plasma termination events, which are MHD instabilities driven
by the pressure gradient and plasma current, are studied, respectively. In the Large Helical Device (LHD)experiment, many MHD instabilities are observed. Especially, if the peaked pressure profile was sustainedby the pellet injection, a collapse event, so-called the core density collapse (CDC), was happen. In nonlinear
MHD simulations, it is expected the CDC is driven by the resistive ballooning mode [1]. Recently, a newimaging diagnostic of the two-dimensional soft-X ray arrays is installed in the LHD. Using the new
diagnostics, perturbations localized at the outward of the torus. That is a characteristic of the ballooning mode.
So, it seems the ballooning mode is observed in the LHD experiments. However, to interpret the experimental
observation, we need to know what kind mode patterns should be observed.
On the other hand, in Wendelstein 7-X (W7-X) experiment, collapse events simultaneously happenedin the plasma core, if the strongly localized plasma current is driven in the plasma core. That expects an MHD
event driven by the plasma current. However, we did not identify what kind mode makes the collapse eventin such a case.
In this study, we study 3D MHD equilibria with reconstructed pressure profile using a 3D MHD
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equilibrium code, which does not assume nested flux surfaces [2]. And then, we will study nonlinear MHD
simulations based on the 3D MHD equilibrium with the magnetic island [3]. In this study, we note nonlinear
saturation to compare with the experimental observation.
2. Numerical Model
Nonlinear MHD simulation is carried out by using the MIPS code (MHD Infrastructure for Plasma
Simulation) which solves the following full MHD equations in the cylindrical coordinates (R, Ԅ, Z),

The ɏ is the mass density, ‫ ݒ‬is the plasma velocity, p is the pressure, B is the magnetic field, Ʉ isthe resistivity, ɋ is the viscosity, ‫ܦ‬ୄ LV WKH SDUWLFOH GL൵XVLRQ FRH൶FLHQW ߯‫ צ‬and ߯ୄ ? are parallel andperpendicular thermal GL൵XVLRQFRH൶FLHQWVUHVSHFWLYHO\DQG Ȟ is the adiabatic constant. In these equations,the variable v = E × B/‫ܤ‬ଶ + ‫ צݒ‬B/B represents the MHD velocity, and ‫ݒ‬௜‫ = כ‬B × ‫݌׏‬௜ /ߩ‫ܤ‬ଶ is thenormalized ion diamagnetic velocity.
The quantities ܵఘ = െ‫ ׏‬ή ߯ୄ ߩ௘௤ and ܵ௣ = െ‫ ׏‬ή ߯ୄ ‫݌׏‬௘௤ are the sources of density and pressure,
where ߩ௘௤ and ‫݌‬௘௤ are the equilibrium density and pressure respectively. The normalization is as follows:
the magnetic field is normalized to the field magnitude on the magnetic axis ‫ܤ‬଴, the density to the density
on the magnetic axis ߩ଴, the velocity to the Alfven velocity ‫ݒ‬஺ = ‫ܤ‬଴ /ඥߤ଴ ߩ଴, and the pressure to ɏ‫ݒ‬஺ଶ.
The time is normalized to the Alfven time ߬஺ = ‫ݔ‬଴ /‫ݒ‬஺ , where ‫ݔ‬଴ is the distance normalization. As aresult of this normalization, all the diamagnetic terms are multiplied by the parameter ߜ௜ = 1/(߱௜ ߬஺ ) =݀௜ /‫ݔ‬଴ = ‫ܭ‬/(‫ݔ‬଴ ඥ݊଴ ), where ߱௖௜ = ݁‫ܤ‬/݉௜ is the cyclotron frequency, ݀௜ is the ion skin depth and K =ඥ݉௜ /(ߤ଴ ݁ ଶ ) is a constant. In these studies, all extended MHD terms were switched off. That means all
simulations are resistive MHD simulations.
The original MIPS code uses the fourth-order fiQLWH GL൵HUHQFH VFKHPH WR approximate the spatial
GL൵HUHQFH7KHfourth order Runge-Kutta method is used for the time integration.
3. Core Density Collapse (CDC) in LHD
In LHD experimentKLJKSHUIRUPDQFHFRQ¿QHPHQW with a superdense core (SDC) inside the internal
GL൵XVLRQEDUULHU ,'% KDVEHHQ achieved using the pellet injection scheme. In the re-heating stage in such adischarge, where the plasma beta increases gradually after ceasing the pellet injection, an DEUXSWÀXVKLQJRI
the central density occurs in some cases. Such a collapse phenomenon, which is named “core density collapse”(CDC), should be avoided since it often limits the density increase. Although the physical mechanism of
CDCs remains unclear, several models based on the magnetohydrodynamics (MHD) instability or the
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equilibrium limit have been proposed. Revealing the mechanism of CDC is one of the key issues in further
GHYHORSPHQW RI WKH FRQ¿QHPHQW WRZDUG WKH KLJK-density branch. This DUWLFOH UHSRUWV WKH ¿UVW DWWHPSW WR
conduct a nonlinear MHD simulation in helical plasma with a large pressure gradient, which corresponds to
the IDB-SDC state of LHD [58, 59]. The result obtained by the simulation would provide us with basic
understandings for the collapse phenomena, including the CDC, from the viewpoint of MHD.
4. Collapse event in core by ECCD current drive in W7-X
Figure 1 shows the time history of two shots with co- and counter-ECCD. For the counter-ECCD case,small crashes of the electron temperature appeared but crashes are not huge. However, for the case of coECCD case, crashes frequently appeared and drops of the electron temperature are huge. Since co-ECCD
increases the rotational transform, it is expected the rotational transform reached ɇ = 1 rational surface or
crossed the rational surface.

Fig.1 Time histories of discharges with co-ECCD and counter-ECCD. Temperature crashes
happed when ECCD is evolved.
The 3D equilibrium is studied by HINT code, which does not assume perfectly nested flux surfaces. The
target configuration is the low-iota configuration, which is used in OP1.1. This configuration has 5/6 islandsin the plasma edge, but those islands are limited by the material limiter. For co-ECCD of 43kA, the rotationaltransform reached ɇ = 1 rational surface at ɏ = 0.4. And, flux surfaces strongly deviated by the natural
resonances. In addition, if the beta effect is induced, the magnetic island is opened. However, clear flux
surfaces are keeping and crashes from the equilibrium effects cannot be expected. Thus, these crash events
will be caused by nonlinear MHD behavior.
To confirm that hypothesis, nonlinear MHD simulations are studied by MIPS code. nonlinear MHD

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036

simulations are initialized by HINT equilibrium in above mentioned. From MIPS simulations, linear andnonlinear behaviors are observed. In linear phase, the ballooning mode clearly appears in the outside of ɇ =1 rational surface, where the magnetic shear is negative corresponding to the positive shear in tokamak, inspite of the magnetic well. However, flux surfaces are keeping clearly. For the beginning of the nonlinear
phase, the ballooning mode destroys the flux surface. Only flux surfaces on the ballooning mode aredestroyed and then the region in the outside of the ballooning mode keeps nested flux surfaces. In the later
phase of nonlinear simulation, flux surfaces are completely destroyed and then the plasma pressure diffuses
along the stochastic field line. That is a reason the electron temperature drops by nonlinear MHD.

Fig.2 magnetic fields and perturbed pressure distributions in nonlinear simulations. The
bolloning strucutre of the perturbation appeared in the outside of ɇ=1 rational surface. In thelater phase of the nolinear state, flux surfaces are completeley broken.
5. Summary
In this study, two types of the plasma collapse events observed in LHD and W7-X are studied
by nonlinear MHD simulation code, MIPS.
Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by NIFS
budget code KEJC001 and was also partially supported by JSPS (the Japan Society for the Promotion of
Science) Grant-in-DLGIRU6FLHQWL¿F5HVHDUFK % +.

References[1] N. Mizuguchi, Y. Suzuki and N. Ohyabu, Nucl. Fusion 49 (2009) 095023.
[2] Y. Todo, N. Nakajima, M. Sato and H. Miura, Plasma Fusion Res. 5 (2010) S2062.
[3] Yasuhiro Suzuki, et al., Nucl. Fusion 46 (2006) L19.

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Overview of the LHD Neutron Diagnostics
M. Isobe1,2, K. Ogawa1,2, E. Takada3, T.S. Fan4, M. Kobayashi1,2,
S. Yoshihashi5, T. Nishitani1, M. Osakabe1,2, and the LHD Experiment Group1
1

National Institute for Fusion Science, National Institutes of Natural Sciences, Toki 509-5292, Japan

2SOKENDAI (The Graduate University for Advanced Studies), Toki 509-5292, Japan
3

National Institute of Technology, Toyama college, Toyama 939-8630, Japan

4

Peking University, Beijing 100871, People’s Republic of China

5

Nagoya University, Nagoya 464-8603, Japan

Abstract
A comprehensive set of neutron diagnostics has been successfully operated in the Large Helical Device(LHD) deuterium discharges, playing an important role in assessing fusion gain (Q), confinement propertyof energetic particles, and/or radiation safety. So far, maximum neutron emission rate (S n ) of 3.3×1015 n/s,
maximum shot-integrated neutron yield of 5.0×1015 n/shot, Q DD of 4×10-4, and equivalent QDT assumed for
deuterium-tritium reaction of 0.11 have been recorded. In addition, energetic-particle confinement studies
have been significantly enhanced through measurements of S n , DD neutron emission profile, and secondary
triton burnup 14 MeV neutron emission rate. The triton burnup ratio defined as the ratio of DD neutron yield
to secondary DT neutron yield has reached 0.45%.
1. Introduction
The Large Helical Device (LHD) has been operated with deuterium gas since March, 2017 [1-3]. The
LHD deuterium operation began with a comprehensive set of neutron diagnostics [4]. Since the LHD isequipped with five intensive neutral beam (NB) injectors, most of neutrons are generated by so-called beamplasma reactions [5]. Therefore, neutron measurement can work as a new tool of energetic-particle (EP)diagnostics in the LHD. Neutron diagnostics have been playing an important role in the LHD deuterium
experiments in enhancement of confinement study of EPs [6-10], assessment of fusion gain (Q) [11], andstrict observance of neutron yield management [12].
2. Neutron diagnostics in LHD
Conceptual design of neutron diagnostics in the LHD was initiated in the late 1990's [13,14]. After longterm preparation, a comprehensive set of neutron diagnostics was established before the start of the deuterium
operation. The LHD is now equipped with (1) three sets of ex-vessel neutron flux monitor (NFM) with wide
dynamic range capability [15-17], providing time evolution of total neutron emission rate (S n ) and/or shotintegrated neutron yield (Y n ), (2) neutron activation system (NAS) having two irradiation ends at horizontal
and vertical ports for Y n measurement with indium disk and/or triton burnup study by use of silicon disk
[4,18], (3) two vertical neutron cameras (VNCs) for radial profile of neutron, i.e., beam ion [19-21], (4) three
scintillating-fiber (Sci-Fi) detectors developed by Los Alamos National Laboratory (LANL) in United States
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of America [22, 23], National Institute for Fusion Science (NIFS) in Japan [24], and National Institute ofTechnology, Toyama College (TOYAMA) in Japan [25, 26] for time-resolved measurement for triton burnup
14 MeV neutron emission rate, (5) neutron fluctuation detector (NFD) for observation of rapid change ofneutron flux due to unusual events such as magnetohydrodynamics (MHD) instabilities [27], and (6) single
crystal chemical vapor deposition (CVD) diamond detector [28] for evaluation of neutron field and/or
development of neutron field suitable for boron neutron capture therapy called BNCT. These have beensuccessfully working and have provided plenty of new information for EP’s behaviors in LHD plasmas.
To increase understanding of confinement property of EPs, the LHD neutron diagnostics will be largely
enhanced in 2019. Third VNC characterized by high-detection efficiency will be installed at the basement
beneath the 1.5-L port, having a vertical line of sight (LOS) at vertically elongated poloidal cross section. To
increase performance of neutron fluctuation measurement, an EJ-200 plastic scintillator characterized by fastresponse capability will be also installed as the second NFD. In addition, three different neutron energy
spectrometers (NESs) will be operated from the 2019 campaign. The first one is time-of-flight enhanced
diagnostics called TOFED developed in Peking University in People’s Republic of China [29]. This task isundertaken as an international collaboration between NIFS and Peking University. The overview of TOFED
on the LHD is schematically depicted in Fig. 1a). The TOFED in the LHD will have a vertical LOS. Fig. 1b)shows expected DD neutron energy spectrum at the end of collimator embedded in the concrete floor [16].
The second one is a fast-neutron scintillation detector based on 7Li-enriched Cs 2 LiYCl 6 :Ce (CLYC) crystal[30]. The 7Li-enriched CLYC is a recently developed scintillation crystal and can provide an opportunity fordirect observation of neutron energy spectrum without unfolding technique. The third one is a conventional
detector based on liquid fast-neutron scintillator designated as EJ-301. Existing neutron diagnostics together
with future detectors in the LHD are listed in Table 1.

Fig. 1a) Overview of TOFED in the LHD, b) DD neutron energy spectrum at the collimator end. The DD
neutron spectrum is calculated by the MCNP6 code [16].
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Table 1. Neutron diagnostics in LHD
System
NFM[15-17]
NAS[4,18]VNC

Sci-Fi
NFD

NES
CVD-D

Ch.
1
2
3
1
2

Detector

In, Al, and Si foils
In, Al, and Si foils

1

Stilbene [19-21]

2-1

EJ-410 (ZnS:Ag + Plastic)

2-2

EJ-410 (ZnS:Ag + Plastic)

1
2
3
1
2

LANL Sci-Fi detector [22, 23]
NIFS Sci-Fi detector [24]TOYAMA Sci-Fi detector [25, 26]
EJ-410 (ZnS:Ag + Plastic) [27]EJ-200 (Plastic)

1

TOFED [29]231

7Li enriched CLYC [30]235U FC, 10B counter
235U FC, 3He counter
235U FC, 3He counter

EJ-301 (Liquid)
Single crystal CVD diamond [28]LocationTop of LHD
10-O port
4-O port
2.5-L port
8-O port
2.5-L port/
vertical LOS
1.5-L port/
Diagonal LOS
1.5-L port/
vertical LOS
8-O port
2.5-L port
2.5-L port
2.5-L port
2.5-L port
Basement beneath
2.5-L port
6-T port
6-T port
9.5-L port

Primary purposes

Status͌Fusion output and Q DD
͌Time-resolved S n

2017~

͌Shot-integrated Y n for DD andsecondary DT neutrons

2017~
2017~

͌Time-resolved DD neutron
emission profile

2018~
2019~

͌Time-resolved DT neutron
emission rate

2017~

͌Neutron fluctuation

2017~
2019~

͌DD and DT neutron
spectrometry2019~͌DD neutron spectrometry
͌Evaluation of neutron field

2019~
2019~
2018~

3. Representative results in the FY2017 and FY2018 campaign
3.1 Neutron emission rate and yield
Dependences of S n on electron density (n e ), magnetic axis position (R ax ), and toroidal magetic fieldstrength (B t ) were systematically investigated in NB-heated plasmas in the early phase of deuterium
experiment. S n in high-B t plasmas is higher than in low-B t plasmas as expected. Also, it increases as R ax is
shifted inwardly, and vice versa as expected. The measurement indicated that S n increases in low-n e regime(<2×1019 m-3), reaches the peak around n e of (2-3)×1019 m-3, and decreases as n e increases [6, 11]. Thistendency

matchesthat

predicted

bycalculation in advance based on steady-state
Fokker-Planck equation [5]. So far, themaximum S n has reached 3.3×1015 (n/s) in
R ax /B t of 3.55 m/2.89 T in the FY2017
campaign with full power NB heating [6].
Fig. 2 shows waveforms of maximum Y nshot obtained in the FY2018 campaign in
R ax /B t of 3.55 m/2.89 T. Full power NB
heating with five NB injectors lasts for 2 s
and electron cyclotron resonance heating(ECRH) is superposed in this shot. n e
increases gradually from 2×1019 m-3 toFig. 2 Maximum neutron yield shot obtained in the

3×1019 m-3 in time. S n almost keeps constant

FY2018 campaign of the LHD.
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040

during NB injection. In this shot, Y n reached 5.0×1015 n/shot which is the maximum record value in the LHD.
3.2 Fusion gain
The fusion gain (Q) is an intrinsic interest
in nuclear fusion experiments. The Q value
defined as the ratio of fusion output divided
by heating power has been systematicallyinvestigated with negative-ion-source-based
NBs (N-NBs) characterized by high injectionenergy of 180 keV. Note that positive-ionsource-based NBs (P-NBs) injected withrelatively low energy of 60-80 keV andECRH are not used in this study because
these are inefficient in terms of neutron
production compared with N-NBs. Q DD andequivalent Q DT as a function of P N-NB_dep areshown in Fig. 3. Here, P N-NB_dep stands fordeposition power of the N-NBs. The Q value

Fig. 3 Fusion gain QDD and equivalent QDT as a functionof deposition power of negative-ion-source-based
neutron beam [11].

increases as P N-NB_dep increases up to ~3 MW, and saturates when P N-NB_dep is over ~3 MW. The maximum
Q DD value has reached 4×10-4 with P N-NB_dep of ~3.5 MW. The equivalent Q DT value is assessed with a helpof the FBURN code [10], assuming the classical slowing down of beam ion without loss. The maximum
equivalent Q DT value is obtained in the FY2017 campaign and is evaluated to be ~0.11. Note that this value
is quite similar to that obtained in large tokamaks with NB power of 5 MW [31].
3.3 Extension of energetic-particle confinement study
Here, the result of triton burnup experiment is described. 1
MeV tritons are produced in deuterium plasmas. 1 MeV tritons
undergo secondary DT reaction through nuclear fusion
reaction with background plasma deuterons. Because kinetic
parameters such as Larmor radius and precessional drift
frequency of 1 MeV triton are quite similar to those of DT-born
3.5 MeV alpha particles, 1 MeV triton can be recognized as asimulation particle of 3.5 MeV alphas. We investigated thetriton burnup ratio defined as the ratio of DD neutron yield tosecondary DT neutron yield measured with NAS and Sci-Fi
detector calibrated with NAS. Fig.4 shows triton burnup ratios

Fig. 4 Triton burnup ratio as a functionof Rax [6].

as a function of R ax obtained in LHD deuterium plasmas withn e of (1~4)×1019 m-3. The triton burnup ratio becomes higher as R ax is shifted inwardly as expected sincehelically trapped energetic-ion orbits tend to match magnetic flux surfaces in the inward shifted configuration.
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041

The maximum triton burnup ratio has reached 0.45% in B t /R ax of 3.55 m/2.89 T in the FY2017 campaign. In
addition to triton burnup, understanding of beam ion transport and/or loss due to MHD instabilities has been
largely increased through measurements of DD neutron emission profile [8, 9].
4. Summary
A highly integrated fusion neutron diagnostics has been established in the LHD, including ex-vessel NFMs
with wide dynamic range capability, a NAS having two irradiation end, two VNCs, three Sci-Fi detectors, a
NFD. To enhance EP physics studies, third VNC and three different types of neutron energy spectrometers
will be installed before the start of FY2019 campaign. So far, maximum S n of 3.3×1015 n/s, maximum Y n of
5.0×1015 n/shot, Q DD of 4×10-4, and equivalent Q DT assumed for deuterium-tritium reaction of 0.11 havebeen achieved. In addition, the triton burnup ratio defined as the ratio of DD neutron yield to secondary DT
neutron yield has reached 0.45%.
Acknowledgements
This work was partly supported by the Japan-China Post-Core-University-Program called Post-CUP. It
was also supported by the LHD project budget (ULHH003, ULHH034, ULHH802 and ULGG801) and NIFS
collaboration research programs (KOAH033 and KOAH037).

References[1] Y. Takeiri, IEEE Trans. Plasma Sci., 46 (2018) 1141.
[2] Y. Takeiri, IEEE Trans. Plasma Sci., 46 (2018) 2348.
[3] M. Osakabe et al., IEEE Trans. Plasma Sci., 46 (2018) 2324.
[4] M. Isobe et al., IEEE Trans. Plasma Sci., 46 (2018) 2050.
[5] M. Osakabe et al., Fus. Sci. Technol. 72 (2017) 199.
[6] M. Isobe et al., Nucl. Fusion 58 (2018) 082004.
[7] K. Ogawa et al., Nucl. Fusion 58 (2018) 034002.
[8] K. Ogawa et al., Nucl. Fusion 58 (2018) 044001.
[9] K. Ogawa et al., Plasma Phys. Control. Fusion 60 (2018) 044005.
[10] K. Ogawa et al., Plasma Phys. Control. Fusion 60 (2018) 095010.
[11] K. Ogawa et al. Nucl. Fusion 59 (2019) 076017.
[12] M. Kobayashi et al., Fus. Eng. Des. 143 (2019) 180.
[13] M. Sasao et al., Fus. Eng. Des. 34-35 (1997) 595.
[14] M. Sasao et al. Diagnostics for Experimental Thermonuclear Fusion Reactor 2, edited by Stott et al.,Plenum Press, New York, 1998, pp.583.
[15] M. Isobe et al., Rev. Sci. Instrum. 85 (2014) 11E114.
[16] T. Nishitani et al., Fus. Eng. Des. 123 (2017) 1020.
-5-

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[17] T. Nishitani et al., Fus. Eng. Des. 136 (2018) 210.
[18] N. Pu et al., Rev. Sci. Instrum. 88 (2017) 113302.
[19] K. Ogawa et al., Rev. Sci. Instrum. 85 (2014) 11E110.
[20] K. Ogawa et al. Rev. Sci. Instrum. 89 (2018) 113509.
[21] H. Kawase et al., IEEE Trans. Plasma Sci. 47 (2019) 462.
[22] G.A. Wurden et al., Rev. Sci. Instrum. 66 (1995) 901.
[23] W.C. Sailor et al., Rev. Sci. Instrum. 66 (1995) 898.
[24] N. Pu et al., Rev. Sci. Instrum. 89 (2018) 10I105.
[25] E. Takada et al., Plasma Fus. Res. 11 (2016) 2405020.
[26] E. Takada et al., Rev. Sci. Instrum. 90 (2019) 043503.
[27] K. Ogawa et al., Plasma Fus. Res. 13 (2018) 3402068.
[28] M. Kobayashi et al., submitted to Journal of Instrumentations.
[29] X. Zhang et al., Rev. Sci. Instrum. 85 (2014) 043503.
[30] A. Giaz et al., Nucl. Instrum. Meth. A 825 (2016) 51.
[31] D.L. Jassby et al., Phys. Fluids B 3 (1991) 2308.

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1 MeV triton orbit analysis in EAST
K. Ogawa1,2, G. Q. Zhong3, R. Zhou3, K. Li3, M. Isobe1,2, and L. Hu3
1) National Institute for Fusion Science, National Institutes of Natural Sciences, Toki-city,
Gifu, Japan
2) SOKENDAI, Toki-city, Gifu, Japan
3) Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, China
1 MeV triton orbit analysis is performed in order to understand the triton confinement
property in EAST. Collisionless Lorentz orbit of 106 1 MeV triton having random velocity
is followed from the birth position calculated by NUBEAM code. Prompt loss is dominant
when the orbit following time is set to be 10-3 s. 1 MeV tritons existing in a wider pitch
angle range are confined in the case of high plasma current conditions.
1. Introduction
Confinement of energetic particles has been intensively studied in magnetic confinement
fusion machine because a fusion burning plasma is sustained by deuterium (D)-tritium(T) born alpha particles. Therefore, energetic alpha particles must be confined well for
realizing a fusion reactor. In a deuterium plasma experiment, following two reactions aremainly occurred: D(d,n)3He and D(d,p)T. By former reaction, 2.45 MeV neutrons are
created, whereas 1 MeV tritons are created by latter reaction. It is worth noting that the
number of two reactions is almost the same. DD-created 1 MeV triton can undergo
secondary reaction with bulk deuteron if the triton slows down in the plasma.
Confinement property of 1 MeV tritons is intensively studied as DT born alpha particles
because kinetic parameters such as Larmor radius and precession frequency of the tritons
are almost the same as that of the alpha particles. In those studies, triton burnup ratio,
which is DT neutron yield divided by DD neutron yield, is one of the common index.
In EAST, integrated neutron diagnostics such as neutron flux monitor, radial neutron
camera, and neutron spectrometers are installed and working successfully [1]. From 2018
experimental campaign, the neutron activation system (NAS) was newly installed in orderto evaluate shot-integrated neutron yield. Triton burnup measurement was initiated with
the NAS. Secondary DT neutron was measured using 10 g silicon foil. Initial
measurements showed that DT neutron yield was around 1011 n/shot. The measurement
has been continued in order to obtain neutron yield on various configurations. In thispaper, 1 MeV triton orbit analysis is reported in order to understand 1 MeV triton

44

044

confinement in EAST.
2. Experimental Setups
In this calculation, collisionless Lorentz orbit of 1 MeV tritons is followed by LORBIT
code [2] in various plasma current (I p ) conditions. The orbit following time is set to be 1
ms, which is quite shorter than the Spritzer slowing down time. Toroidal field ripple is
not considered. Velocity of triton is randomly
distributed using random numbers. We
randomly launched 106 1 MeV triton according
to the birth profile of 1 MeV tritons calculatedusing NUBEAM code [3] as shown in Figure 1.
Here, only co-inject neutral beam injector isconsidered. The equilibrium is given by EFIT
result. Electron temperature, and electron
density in the plasma center are 1.64 keV, and3.2×1019 m-3, respectively. Radial profiles ofelectron temperature and electron density are
assumed to be parabolic profile.

Fig.1 Emissivity profile of 1 MeV triton
calculated by NUBEAM code.

3. Calculation Results
Figure 2 shows the typical 1 MeV triton orbit in I p of 300 kA, 450 kA, and 1 MA cases.
Here, toroidal magnetic field strength is around 2.3 T, and ion grad-B drift is directed toFig.2 Typical 1 MeV triton orbit calculated by LORBIT code.

45

045

be upward. Co-going transit 1 MeV triton having pitch angle of 175 degrees is launchedat (R, Z) of (2.25 m, 0.0 m). The triton is lost at I p of 300 kA, whereas the triton is confinedat I p of 450 kA and 1 MA. Counter-going transit 1 MeV tritons having a pitch angle of 5
degrees is launched at (R, Z) of (2.05 m, 0.0 m). The triton is confined at I p of 450 kA and1 MA. Banana 1 MeV triton having the pitch angle of 80 degrees is launched at (R, Z) of(2.15 m, 0.0 m). The triton is only confined at I p of 1 MA. Although the deviation of orbit
from the magnetic flux surface becomes smaller as the increase of I p , the deviation isrelatively large even in 1 MA case because of the high energy.
Confinement of 1 MeV triton is surveyed at I p of 300 kA, 450 kA, 600 kA, 800 kA, and
1 MA cases. Figure 3 shows the time evolution of 1 MeV triton loss. Number of lost 1
MeV tritons increases rapidly at 10-6 s, then almost saturated at 10-5 s in all the cases. This
time evolution shows that prompt loss is dominant up to 10-3 s. The loss ratio at I p of 300
kA, 450 kA, 600 kA, 800 kA and 1 MA reached 85%, 73%, 56%, 48%, and 33%,
respectively. Number of lost ions at 10-3 s decreased with the increase of I p . Pitch angle
distribution of confined 1
MeV tritons are plottedon Figure 4. Black line
shows the pitch angle
distribution of launchedparticles. Confinement ofco-going transit particles
are better than that ofcounter-going transit andbanana particles. 1 MeV
tritons existing in wider
pitch angle range can be
confined in higher I p
cases. Some of co-goingparticles are confinedeven if I p of 300 kA.
Most of banana particles
are lost even in I p of 1
MA case because thebananawidth
iscomparable with a minor
radius of a plasma.

Fig.3 (a) Time evolution of number of lost 1 MeV tritons. (b) Pitch
angle distribution of confined 1 MeV tritons.

46

046

Strike points of 1 MeV tritons in I p of 300 kA, 600 kA, and 1 MA cases are plotted in
Figure 5. Although number of particles decrease with the increase of I p , strike points of 1
MeV tritons are not so changed. Note that the strike points of 1 MeV tritons are mainly
located at the upper panel of the vessel because the ion grad-B drift is directed to be
upward. Strike points of 1 MeV triton are located at the lower panel if we changed
direction of toroidal magnetic field. The location of strike points are consistent with thedirection of ion grad-B drift, as expected.

Fig.4 Distribution of strike point of 1 MeV tritons.

4. Summary
Study of 1 MeV triton confinement is performed using Lorentz orbit code LORBIT inorder to understand 1 MeV triton confinement characteristics in EAST. Typical 1 MeV
triton orbit shows that the deviation of orbit from the flux surface decrease with theincrease of I p , as expected. The deviation is comparable with minor radius of a plasma
even in 1 MA case because of high-energy of tritons. Time evolution of 1 MeV triton loss
shows that the prompt loss of the triton occurs at t<10-5 s in all I p cases. 1 MeV tritons
existing wider pitch angle region are confined in higher I p cases compared with lower I p
cases. Strike points of 1 MeV triton shows that most of the particles are lost on the upper
wall which is consistent with the direction of ion grad-B drift.
References[1] G. Q. Zhong et al 2016 Plasma Phys. Control. Fusion 58 75013.
[2] M. Isobe et al 2009 J. Plasma Fusion Res. SERIES 8 330.
[3] A. Pankin et al 2004 Comput. Phys. Commun. 159 157.

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Spectroscopic study of impurity behavior in the edge stochastic magnetic field layer ofLarge Helical Device
T. Oishi1,2, S. Morita1, M. Kobayashi1,2, G. Kawamura1,2, Y. Kawamoto1, M. Goto1,2, andthe LHD Experiment Group1
1

National Institute for Fusion Science, National Institutes of Natural Sciences, 322-6, Oroshi-cho, Toki,
Gifu 509-5292, Japan2

Department of Fusion Science, SOKENDAI (Graduate University for Advanced Studies), 322-6, Oroshicho, Toki, Gifu 509-5292, Japan

Abstract
Carbon impurity transport in a thick stochastic magnetic field layer called “ergodic layer” located at theedge plasma of LHD was studied for both deuterium (D) and hydrogen (H) discharges by EUV and VUVspectroscopy, and EMC3-EIRENE simulation. By comparing emission intensities from carbon impurity
ions in high- and low- charge states, impurity screening is found to be more effective in D plasmas rather
than H plasmas. Enhancement of the bulk-impurity collision frequency in D plasmas might be one of thereasons of the effective impurity screening.

In high density discharges, the carbon impurity flow in theergodic layer is directed toward the same direction of the friction force, indicating the presence of the impurity
screening.

The carbon flow velocity measured by the VUV spectroscopy is smaller in the deuterium

plasmas. The difference of the flow velocity was reproduced by the simulation qualitatively.
1. Introduction
Stochastization of edge magnetic fields is extensively studied not only for the ELM mitigation but also
for the plasma detachment and the impurity transport.

A thick stochastic magnetic field layer called“ergodic layer” of the large helical device (LHD) consists of stochastic magnetic fields with threedimensional structure intrinsically formed by helical coils, while well-defined magnetic surfaces exist inside
the last closed flux surface [1].

It is therefore extremely important to study the impurity behavior andtransport in the ergodic layer and to compare with those in the scrape-off layer of tokamaks.

In LHD, it isfound that carbon impurities are screened by the presence of the ergodic layer [2] and iron impurities are
more effectively screened. As a result, the iron density in core plasmas of LHD is found to be extremely
low despite the stainless steel vacuum vessel [3]. A transport model for the impurity behavior in the ergodic
layer has been proposed considering the parallel momentum balance on impurity ions along a magnetic field
line connecting the core plasma and the divertor plate based on the following equation;mZ

wVZ //wtV  VZimp
wTwT
1 wTZ n Z//
 ZeE //  m Z i //
 0.71Z 2 e  2.6 Z 2 i ,n Z ws
Ws
ws
ws

(1)

where five terms in the right-hand side are contributions of impurity ion pressure gradient, parallel electricfield, friction force between bulk ions and impurity ions, electron thermal force, and ion thermal force, in theorder [4]. Among these terms, the friction force term and the ion thermal force term are the dominant terms.
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48

048

When the ion density gradient increases, the friction force increase resulting the impurity flow is directed
toward divertor plates, which means the impurity screening. On the other hand, when the ion temperature
gradient increases, the ion thermal force increases resulting that the impurity flow is directed toward the coreplasmas, which means the impurity accumulation. Based on the model, the parallel flow of the impurity
ions is considered to be a key mechanism to determine impurity distributions in the ergodic layer.
2. Impurity screening phenomena observed by spectroscopy
In the present study, spectroscopic diagnostics has been performed in discharges with a magnetic
configuration with the position of the magnetic axis, R ax , of 3.6 m and the toroidal magnetic field, B t , of 2.75
T. The discharge initiated by the electron cyclotron heating (ECH) is grown by three neutral beam injections
based on the negative ion sources (n-NBI) beams with total port-through power of 10 MW and maintainedwith a flat-top phase for 1 s. We performed similar operations for the working gas of H 2 and the beam
particle of the neutral beam of hydrogen (denoted as “H plasma”), and the working gas of D 2 and the beam
particle of the neutral beam of hydrogen (denoted
as “D plasma”).
Intensities of carbon line emissions aremonitored as an indicator of the impurity
screening. CIII (977.03 Å, 2s2-2s2p) and CIV(1548.02 Å, 2s-2p) are measured using a 20 cm
normal incidence VUV spectrometer [5], while CV(40.27 Å, 1s2-1s2p) and CVI (33.73 Å, 1s-2p) aremeasured using a grazing incidence EUV
spectrometer [6]. The ionization potential, Ei, forC2+, C3+, C4+, and C5+ is 48 eV, 65 eV, 392 eV, and490 eV, respectively. Therefore, CIII and CIV
radiation is emitted by carbon ions with low Ei
located at the outer region of the ergodic layer,
while CV and CVI radiation is emitted by carbon
ions with high Ei located at inner region of theergodic layer. Figure 1 shows the electron density
dependence of line intensity of (a) CIII, (b) CIV, (c)CV, and (d) CVI normalized by the line-averaged
electron density and (e) a line ratio of CV / CIV as
an indicator of the impurity screening effect.
Smaller values of the ratio leads to enhancement of
the impurity screening effect. The line ratio

Fig. 1decreases with the electron density, because carbon

intensity of (a) CIII, (b) CIV, (c) CV, (d) CVI

lines emitted from the outer region of the ergodic

normalized by the electron density and (e) line ratio

layer (CIII, CIV) increase, while those from inner

CV / CIV for deuterium and hydrogen discharges.
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Electron density dependence of line

region (CV, CVI) decrease. It indicates enhancement of the impurity screening in the high density regime.
Figure 1(e) also shows a comparison of the line ratio between the D plasmas compared to the H plasmas.
The impurity screening effect is more obvious in the D plasmas. Enhancement of the friction force in D
plasmas might be one of the reasons of the effective impurity screening.
3. Relationship between the impurity screening and the impurity flow
Figure 2 shows vertical profiles at the bottom edge of the ergodic layer of the flow velocity derived from theCIV line emission measured by VUV spectroscopy for an H plasma and a D plasma with a magnetic
configuration with R ax = 3.6 m and B t = 2.75 T. The observation range of the edge profile measurement of
the VUV spectroscopy is shown in Fig. 2 [7]. The flow velocity along the sightline, v R , is given by v R = c('O / O), where c is the light speed, 'O the Doppler-shift and O the wavelength of line emission. The
measured flow velocity is projection of the flow along the observation chord which can be approximately
considered to be the direction of the plasma major radius.

Therefore, a variable of v R is used to indicate themeasured flow value. Positive and negative sign in thevertical axis of Fig. 2 corresponds to the outboard an
inboard direction along the plasma major radius,
respectively. As shown in the figure, the flow velocity
toward the outboard direction develops clearly with the
maximum value at Z = -480 mm, which is a location
close to the outermost region of the ergodic layer in the
H plasma. This direction is same as the friction force
in the parallel momentum balance calculated with athree-dimensional simulation code, EMC3-EIRENE[8]. On the other hand, the maximum value of the flow
velocity in the D plasma is clearly smaller than that in
the H plasma.
The electron density dependence of the maximum
value of the observed flow is summarized in Fig. 3.
All plots in the figure have directions same as thefriction force. In the case of the H-discharge, the flow
increases with the electron density.

The result

supports a prediction by the simulation that the friction
force becomes more dominant in the force balance inhigher density regime, which results in the increase of
impurity flow causing the impurity screening. In thecase of the D-discharge, the flow has a smaller value.
In the friction force term in the equation of the
momentum balance, m Z (V i// - V Z// imp) / W s , the parallelFig. 2 Vertical profile at the bottom edge ofthe ergodic layer of impurity flow velocity
derived from the Doppler profile of the CIV
line emission measured by VUV spectroscopy
in a D plasma (red open circle) and a H plasma(blue closed circle). The observation range ofthe VUV spectroscopy is illustrated together.

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velocity of the bulk ion, V i// , and the collision time
between the bulk ion and the impurity ion, W s , might
be changed between the H-discharge and the Ddischarge.

Further experiments and simulations

for the H-D comparison are needed to clarify thedifference.
4.

Summary

Carbon impurity transport in a thick stochastic
magnetic field layer called “ergodic layer” locatedat the edge plasma of LHD was studied for both
deuterium (D) and hydrogen (H) discharges by EUV
and VUV spectroscopy, and EMC3-EIRENE
simulation.

By comparing emission intensities

Fig. 3 Observed C3+ flow at the bottom edge offrom carbon impurity ions in high- and low- charge

the ergodic layer in the H and D plasmas as astates, impurity screening is found to be morefunction of density for inward-shifted magnetic

effective in D plasmas rather than H plasmas.

configuration with Rax = 3.6 m.

Enhancement

of

the

bulk-impurity

collision

frequency in D plasmas might be one of the reasons of the effective impurity screening. In high density
discharges, the carbon impurity flow in the ergodic layer is directed toward the same direction of the friction
force, indicating the presence of the impurity screening. The carbon flow velocity measured by the VUVspectroscopy is smaller in the deuterium plasmas. The difference of the flow velocity was reproduced by
the simulation qualitatively.
Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by NIFS budget code
ULPP010 and was also partially supported by the Ministry of Education, Science, Sports and Culture Grantin-Aid for Scientific Research for Young Scientists (B) (17K14426).
References[1] T. Morisaki et al., J. Nucl. Mater. 313-316 (2003) 548.
[2] M. B. Chowdhuri et al., Phys. Plasmas 16 (2009) 062502.
[3] S. Morita et al., Nucl. Fusion 53 (2013) 093017.
[4] M. Kobayashi et al., Nucl. Fusion 53 (2013) 033011.
[5] T. Oishi et al., Plasma Fus. Res. 10 (2015) 3402031.
[6] M. B. Chowdhuri et al., Appl. Opt. 47 (2008) 135.
[7] T. Oishi et al., Appl. Opt. 53 (2014) 6900.
[8] T. Oishi et al., Nucl. Fusion 58 (2018) 016040.

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3D effects of neon injection position on the toroidally symmetric/asymmetric heat flux
distribution on EAST
B. Liu1, S.Y. Dai1, G. Kawamura2,3*, L. Zhang4, Y. Feng5 and D.Z. Wang1
1

Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education),
School of Physics, Dalian University of Technology, Dalian 116024, China
2National Institute for Fusion Science, National Institutes of Natural Sciences, Toki 509-5292, Japan3

Department of Fusion Science, SOKENDAI (the Graduate University for Advanced Studies), Toki
509-5292, Japan
4

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China
5Max-Planck-Institut für Plasmaphysik, D-17491 Greifswald, Germany*

presentation at the Post-CUP workshop

Abstract
The heat flux deposition on the divertor targets with neon impurity injection on EAST has been
investigated by the three-dimensional (3D) edge transport code EMC3-EIRENE. Impacts of the poloidal
neon injection position on heat flux deposition have been studied. It is found that neon impurity injectedaround the separatrix at the divertor (i.e. strike point) can lead to a toroidal asymmetric distribution of heatloads on the same-side targets as the injection position.

1. Introduction
One of the most critical issues in fusion devices is mitigation of the heat flux deposition on theplasma-facing components (PFCs) [1,2]. The extrinsic impurity seeding is usually utilized to achieve thepower dissipation and heat loads reduction in fusion facilities. The gas puffing such as neon, nitrogen,argon has been investigated on Alcator C-Mod [3], JET [4], LHD [5] and EAST [6]. Spatial localization of
impurity seeding may result in the toroidal asymmetry of heat flux distribution on the divertor targets. The
previous experimental results on EAST show that the argon impurity injection at the divertor region canlead to suppression in the divertor heat loads [6]. However, the toroidal symmetric/asymmetriccharacteristics of heat flux deposition under the toroidally-localized gas injection are not well understood
on EAST due to diagnostic limitations. Hence, modelling predictions of symmetric/asymmetric heat flux
distribution are important to obtain a better understanding of the underlying mechanisms involved and to beable to optimize impurity puffing position to minimize heat loads on EAST.

2. Construction of EMC3-EIRENE simulations
The EMC3-EIRENE code [7,8] can deal with arbitrary magnetic confinement configurations and has
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been applied to many fusion devices. The EMC3 code solves a set of time-independent Braginskii’s fluid
equations for the mass, momentum and energy for electrons and ions using a Monte Carlo (MC) method [7]and is iteratively coupled with EIRENE [8], which solves the kinetic Boltzmann equation for neutral atoms
or molecules and gives the particle momentum and energy source terms due to plasma-neutral interactions.
In this work, the computational grid used in the simulation is constructed based on shot number #74209 at
4.0s of EAST, which is the upper single null magnetic configuration. A two-dimensional poloidal crosssection of the computational grid is shown in figure 1. Deuterium is used as the main plasma species andneon is considered as the extrinsic impurity to be injected into the torus. The arrows in figure 1 indicate theneon impurity seeding directions and positions (named I1~I5). The red points named corner 1 and corner 2
indicate the corner of each divertor target. The purple dashed line is used to indicate the boundary between
the inboard and outboard sides of the torus for the discussion in section 3. The toroidal location of the neon
impurity injection position is defined as the toroidal angle of ɔ=0° in this study. The injected neon flux
with an initial energy of 0.03 eV is 100 A for all neon injected cases. The input power is 2.66 MW, which isequally divided into the power of electrons and ions in the simulation. The upstream density is set to 1.0×
10ଵଽ ݉ିଷ, which is regarded as boundary condition for particle transport. The cross-field transport ofparticle and heat modelled by diffusive transport coefficients are assumed as Dୄ =0.4 mଶ ‫ି ݏ‬ଵ and ɖୄ=3Dୄ . The impurity perpendicular diffusivity D୧୫୮ is assumed to be same as the background plasma, i.e.
0.4 mଶ ‫ି ݏ‬ଵ.

Figure 1. The poloidal cross section of thecomputational grid used in EMC3-EIRENE for
EAST modelling. The red balls mark the cornerof each divertor target. Five neon impurity
injection positions are indicated by arrows. The
purple dashed line indicates the boundary
between the inboard and outboard of the torus.
The black dashed line indication the vertical
positions across the X-point.

3. Results and discussion
Figure 2 shows the 2D distributions of heat flux deposition on the upper inboard divertor targets withneon impurity injected at the positions of I1, I2 and I3. The horizontal and vertical coordinates indicate thetoroidal angle and the distance from the position on the target plates to the corner 1, respectively. It is seen

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that the downstream injections at I2 and I3 can result in a lower heat flux deposition on the upper inboard
divertor targets compared with the upstream injection at I1. The most effective suppression of heat loads on
the upper inboard divertor targets is achieved at I2 with the peak value of about 1.2 MW/mଶ. Moreover, atoroidal asymmetric distribution of heat loads on the upper inboard targets is obtained with neon impurity
injected at I2. From the toroidal injection angle of 0°, the heat flux reduces gradually as the toroidal angle
increases. The heat loads in the range from about 40° to 120° is much lower than other toroidal locations asshown in figure 2 (b).

Figure 2. The 2D distributions of heat flux deposition on the upper inboard divertor targets with neon
impurity injected at I1 (a), I2 (b) and I3 (c), respectively. The horizontal and vertical axis indicate thetoroidal angle and the distance between the corner 1 and the positions on target plates.

Figure 3. The heat flux distributions on the upper inboard (a) and upper outboard (b) divertor targets at
toroidal angles of ĳ ° and 60° with impurity seeded at I2 and without neon injection.
Figure 3 shows the heat flux distributions along the divertor targets poloidally for the upper inboard
and outboard divertor targets at toroidal angles of ɔ=0° and 60°. The horizontal axis is the distance fromthe location on target plates to each corner of divertor target. The peak heat loads on the upper inboard
targets in figure 3 (a) are about 1.1 MW/mଶ and 0.6 MW/mଶ at ɔ=0° and 60°, respectively. However,
the heat flux distributions for the upper outboard targets with injection I2 are similar at ɔ=0° and 60° in
figure 3 (b), which does not show an asymmetric property of the heat flux distribution. This indicates thatthe neon impurity injected around the strike point at the divertor can lead to a toroidal asymmetric
distribution of heat loads on the same-side targets as the injection position. In addition, the heat flux
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distribution without neon impurity injection is also presented in figure 3. The heat loads on the upper
inboard and outboard without neon impurity seeded are about 2.8 MW/mଶ and 3.7 MW/mଶ ,respectively. In comparison with no neon injection, the neon impurity injected at I2 results in the
suppression of heat loads by about 61% and 48% on the upper inboard and outboard targets, respectively.

4. Summary
The heat flux deposition on the divertor targets with different poloidal neon impurity injection
positions has been investigated by EMC3-EIRENE code. The neon impurity is seeded at upstream andtargets plate to study the 3D impacts of spatial impurity transport and radiation on the divertor heat flux
distribution. It is found that a lower heat flux deposition is obtained with neon impurity injection position at
the downstream divertor compared with that at the upstream. Moreover, neon impurity injected near the
strike point can lead to not only the larger reduction in heat loads but also the asymmetric distribution ofheat flux on the same-side targets as injection positions. The neon impurity injected near the strike point
leads to a larger neon ions density and radiation power at the same-side of the torus as the injection position,
which results in the lower heat loads on the same-side divertor targets in comparison with the upstreamneon injection.
Acknowledgments
This work was supported by National MCF Energy R&D Program of China Nos: 2018YFE0311100,
2018YFE0303105, 2017YFE0301206, 2017YFE0300501 and 2017YFE0300402, National Natural Science
Foundation of China under Grant Nos. 11405021, 11675037, 11575244 and 11775269, High-level talentinnovation support program of Dalian No. 2017RQ052, and the Fundamental Research Funds for theCentral Universities No. DUT18LK03, and was supported partly by Japan-China Collaboration for Fusion
Research (Post-CUP Collaboration) and JSPS KAKENHI Grant Number 16K18340.

References[1] Federici G et al. 2001 Nucl. Fusion 41 1967.
[2] Loarte A et al. 2007 Nucl. Fusion 47 S203.
[3] Loarte A et al. 2011 Phys. Plasmas 18 056105.
[4] Oberkofler M et al. 2013 J. Nucl. Mater. 438 S258-261.
[5] Tanaka M et al. 2017 Nucl. Mater. Energy 12 241.
[6] Chen J B et al. 2017 Chin. Phys. B 26 095205.
[7] Feng Y et al. 2004 Contrib. to Plasma Phys. 44 57.
[8] Reiter D et al. 2005 Fusion Sci. Technol. 47 172.

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A particular method for measuring low edge ion temperature
based on VUV spectroscopy
S. Morita
National Institute for Fusion Science, Toki 509-5292, Gifu, Japan

Abstract
Ion temperatures have been measured with high accuracy in Compact Helical System (CHS) using a
1m normal incidence VUV monochromator by carefully checking a slit function, which was obtained from
LyD spectral profile by changing an entrance slit width in steady state ECR discharges for wallconditioning. As a result, low ion temperatures could be successfully measured at T i 10eV from VUVlines located in the edge region of CHS plasmas, e.g. CIV at 1548.2Å, while high ion temperatures(Ti400eV) were observed in plasma central region using OVII at 1623.3 Å. Hopefully, the present
method will be helpful for development of edge ion temperature profile measurement.
1. Introduction
Recently, impurity/particle transport in edge plasmas including stochastic magnetic field layer has
been well studied experimentally and theoretically through 1-D and 2-D spectroscopy and EMC3-EIRENE
code calculation. However, the study of energy/heat transport in edge plasmas is entirely insufficient due
to a lack of experimental data, in particular, ion temperature distribution. Sufficient energy relaxation
between electrons and ions cannot be expected in the plasma edge, i.e. TezTi. Then, the edge ion
temperature profile measurement is important for evident progress of the edge energy/heat transport study.
Results of ion temperature measurement using VUV monochromator in Compact Helical System (CHS) arepresented, which were mainly carried during 1990-1992.
2. Spectral resolution
Spectral resolution of spectrometers is given by
O/'O a mN,(1)

where m is the diffraction order, and the value of N (=1/d) is the number of grating grooves per unit length,
is usually used for indicating the grating specification. Then, d means distance between two adjacentgrating grooves. Typical value of N is 103/mm for grating and 107/mm for crystal.
are roughly given as follows;
Visible range: 4000-7000Å
Vacuum ultraviolet (VUV) range: 300-3000Å
Extreme ultraviolet (EUV) range: 10-500Å
X-ray range: 1-10Å
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Wavelength ranges

Since the angle of incidence of EUV spectrometers is very large, e.g. 89q, the spectral resolution is nothigh. In general the ion temperature measurement in EUV range is difficult. On the other hand, visible
and x-ray spectrometers have an advantage for the ion temperature measurement due to the longwavelength and high N value, respectively.

However, disadvantages also exist in the two spectrometers,

i.e. edge visible lines are affected by Zeeman Effect and kind of noble gases useful for the central ion
temperature measurement is very limited in X-ray spectroscopy. Although Ar gas is usually used for the
X-ray diagnostic, He- and H-like Ar ions do not exist in the plasma center at Te!2-3keV. The x-ray
spectroscopy using Ar gas may work well in ITER as the edge ion temperature diagnostic.
3. Zeeman EffectZeeman separation, 'E, at magnetic field strength, B, is given by'E = gePBB,(2)

where PB and ge are the Bohr magneton and g value (a2.0), respectively.

Therefore, the Zeeman splitting

is large with increasing magnetic field strength. The visible spectrum basically arises at a transition
between two excited levels because the transition energy is very small.

Since the total number ofsublevels is described by ~2J+1~ (J=S+L at LS coupling), the excited level intrinsically involves many
sublevels, while those are degenerated at B=0. When B increases, the degeneracy is disappeared and the
Zeeman splitting is appeared. Typical examples are shown in Fig.1 (a) and (b) for HD and HeI visiblelines, respectively. The spectral line can be clearly resolved in S (parallel to B field) and V (perpendicular
to B field) components. As a result, the ion temperature cannot be obtained straightforwardly from themeasured visible spectra. On the contrary, VUV lines are not practically affected by the Zeeman Effectbecause the transition energy is relatively large and the transition is connected to the grand level of electric
configuration.

(a)

(b)

V

S

Fig. 1VS

Zeeman Effect of (a) HD (6563Å) [1] and (b) HeI (5876Å) [2] spectra. When polarizer is

not used, measured spectral profiles are broadened by a mixture of S and V components.
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4. Entrance slit function of 1m normal incidence VUV monochromator
In order to measure the ion temperature using a 1m VUV normal incidence monochromator, an
entrance slit function has to be accurately determined because the monochromator has a large spectral
aberration due to a simple focal system without use of mirror. At first the entrance slit must be carefully
checked with laser light and simple microscope. Two sharp knife edges have to be set accurately parallelto the grating axis and zero position of the entrance slit has to be determined after checking an exact
parallel position of the two knife edges along the slit height.
The slit function was obtained by scanning the wavelength in steady ECR and He-glow discharges forwall cleaning in CHS. LyD (1215.67Å) and HeI (584.3Å) lines were measured for checking the slit
function. Mercury lamp was also used in N2 flow condition to avoid air attenuation of VUV lines at HgI
2536.5Å, HgI 3125.66Å and HgI 3131.83Å. The results from LyD spectral profile measurement are
shown in Figs.2 (a), (b) and (c) at entrance slit width of 10, 20 and 30Pm, respectively.

The exit slit widthis identical to the entrance slit width. A secondary electron multiplier (R595: Hamamatsu co. ltd.) is used
for detecting the VUV line.

The detector possesses sufficient quantum efficiency in most VUVwavelength ranges, i.e. 300-1800Å.

(a)Fig. 2(c)(b)

Slit function of 1m normal incidence VUV monochromator obtained through LyD spectral

profile measurement at entrance slit width of (a) 10Pm, (b) 20Pm and (c) 30Pm.

5asymmetric, e.g. 10Pm case in Fig. 2(a). It indicates the spectral
aberrationinfocusingmechanismofthepresent

VUV

monochromator. The slit function is further checked for different
wavelength ranges using several lines mentioned above.

Wefound the slit function hardly changes against the wavelength.
The spectral resolution is also examined as a function ofentrance slit width. The result is compared between LyD and HeI

Spectral resolution (Å)

When the entrance slit width is narrower, the slit function is4

LyD: 1216Å
HeI: 584Å

3
2
1
0
0

100 200 300 400 500 600
Entrance slit width (Pm)

It is clear from the figure that theFig. 3 Spectral resolution as aspectral resolution is identical between two lines. We understandfunction of entrance slit width

the spectral resolution is almost constant against the wavelength.

at Ly (solid circles) and HeI

lines, as shown in Fig.3.

(solid triangles) lines.
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5. Results of edge ion temperature measurement
5-1 Ion temperature from OVISpectral profiles of OVI (1031.9Å) were observed from ECH plasmas (Bt=0.9T, PECHa100kW,
Te(0)a0.4keV, ne=0.6x1013cm-3) in CHS.

Since the density of ECH plasmas is very low, the ion

temperature is entirely low due to less equipartition between ions and electrons. Of course the OVI
emission is not located in the central plasma region, but the ion temperature of OVI is not so different fromTi(o) because the ion temperature profile seems to be flat along the plasma radius direction. The ion
temperature analyzed from OVI spectral profiles is plotted in Fig.4 as a function of discharge time. The
electron density shows a gradual increase during the ECH pulse of 35ms, then the electron and ion
temperatures begin to decrease during the discharge.
The ion temperature is analyzed by
I(O) = ³ [+f o -f] G(O') S(O-O') dO',

(3)

where I(O) and G(O') are the measured spectral profile and the Gauss profile, respectively. S(O-O') is theslit function obtained from Fig. 2(a). As G(O') and S(O-O') are normalized, I(O) is also normalized as³ [+f o -f] I(O) dO = 1.

(4)

The ion temperature can be obtained from analyzed G(O') profile, as shown in Fig. 5. A very low OVI ion
temperature can be obtained with small error bar at the end of ECH pulse, Ti(OVI) = 19r3eV at t=35ms.

150

Ti (eV)
from OVI

(a)

100
(b)

50
ECH
0

0

10
0

20
t (ms)

30

40

(c)

Fig. 4 Temporal ion temperature behavior
analyzed from OVI profile during ECH
discharge.

Fig. 5 Measured OVI profile I(O) (solid line) and slit function S(O-O') at (a) t=10ms, (b) t=19ms and(c) t=35ms.
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5-2 Ion temperature from CIV
Ion temperature was also measured from CIV (1548.2Å) spectral profile during NBI discharges in
CHS. The ion temperature decay after switching off the NBI pulse is plotted in Fig.6. The edge ion
temperature measured from CIV keeps a constant value of Ti(CIV)a40eV during 2ms after switching off
the NBI pulse. After switching off the NBI beam the heat input to the edge plasma is a function of plasma
confinement time, WE, and beam slowing down time, Ws. Since the energy exchange time between ionsand electrons, Wei, has a similar value to the WE in the discharge, i.e. WeiaWE=3ms, the ion temperature is
probably identical to the electron temperature. In the NBI discharge of CHS the electron temperature
immediately starts to decrease after switching off the NBI beam, while the NBI fast ions mainly heatelectrons and the value of Ws (a5ms) is longer than WE. Therefore, the constant edge ion temperature phaseof 2ms after switching off the NBI beam may be reasonable because it simply reflects WE.
Based on the present method an extremely low edge ion temperature is successfully observed, e.g.
Ti(CIV)=9eV at t=136.6ms, as shown in Fig.7. A small temperature increase during t=135-136ms may
reflect an effect of recombination from C4+ ions to C3+ ions.

(a)

(b)

Fig. 6 Temporal ion temperature behavior analyzed(c)from CIV profile at the end of NBI discharge.
Fig. 7

Measured CIV profile I(O) (solid line) and slit

function S(O-O') at (a) t=126.4ms, (b) t=131.8ms and (c)t=136.6ms.
6. Summary
The present method based on accurate determination of slit function enabled to observe low edge ion
temperature in VUV region. If a VUV spectrometer is combined with concave mirror system [3], a spatial
profile measurement of edge ion temperature may be possible.
References[1] A. Iwamae, A. Sakaue, N. Neshi et al., J. Phys. B 43 (2010) 144019.
[2] K. Mizushiri, K. Fujii, T. Shikama et al., PPCF 53 (2011) 105012.
[3] S.Morita, S.Muto and M.Sakurai, Fusion Eng. Des. 34-35 (1997) 211
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ICRF Heating on EAST
Xinjun Zhang1,a) , C.M. Qin1), L.N. Liu1), Y.P. Zhao1), H. Yang1) ,B. Lyu1),J.F. Chang1),J.
Huang1), M.H. Li1), B.J. Ding1), X.Z. Gong1), B.N. Wan1), Y.T. Song1), J.G. Li 1) , L. Ai1) ,L.L. Ping1) , and EAST Team1

Chinese Academy of Sciences, Hefei 230031, China

Abstract
Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary
auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). This paper
summarizes effects of ICRF power on LH coupling and modification of heat loads on EAST. Different
rotation behavior are observed in the LHCD discharges. ICRF can also be induced counter-current rotationin the LHCD L-modes discharges. In Hydrogen minority in Deuterium plasma, the interaction of fast wave
and fast D-NBI ions are detected on EAST. Acceleration of fast D ions are confirmed by neutron yield.
1. Introduction
The research objectives of EAST are to perform advanced tokamak research in high performance regime
and to explore methods for achieving a steady-state operation for a tokamak fusion reactor[1,2]. RF powerin the ICRF is one of the primary auxiliary heating techniques for EAST. The ICRF system[3] on EAST
provides plasma heating and current drive through various scenarios over a range of magnetic fields. A 12MW
ion cyclotron resonance heating (ICRH) system has been designed and constructed on ASIPP for EAST. The
system consists of eight independent power lines. each designed to generate and launch 1.5 MW of RF power
into the machine during a 1000-s period in the 25 to 70 MHz frequency range. In this paper, we will present
an overview of the main experimental results from recent campaign on EAST.
2. Challenges for ICRF Coupling and Heating on EAST
The performance of ICRF heating has progressed steadily in the EAST [4-11]. For the present EAST
antennas, the coupled power has been limited often by arcing in the antenna and transmission lines and bylarger reflected power due to ELMs and others. This is thought to be a result of light loading at the low
densities typically used for long pulse operation. Thus, EAST has identified increasing the coupled power
and improving the core heating effectiveness as high priority. The coupling experiments were focus onantenna phasing and gas puffing. The coupling performance was better under lower wave number. By fuelingthe plasma from antenna local gas injectors, as shown in fig.1, placed as uniformly spaced array from top tobottom at each side limiter of the B-port antenna, which works in dipole phasing, the coupling resistance of
the B-port antenna increased obviously. Furthermore, the coupling resistance of the I-port antenna wasinsensitive to a smaller rate of gas puffing but a sharp increase in the coupling resistance of the I-port antenna
occurred when the gas injection rate was higher. which was mainly caused by the toroidal asymmetricboundary density arising from gas puffing.

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Antenna B
Antenna I(Rc-Rc0)/Rc0

0.3
0.2
0.1
0.0
7

8

9
10
N (1020/s)

11

12

Figure 1. Relative improvement of antenna resistance with gas puff
3.

Observations of impurities behavior and Modification of LH coupling and heat load

As shown in Fig. 2(a) the application of ICRF power can reduce LH coupling efficiency.
Previous experiments on Tore Supra[12], EAST[13], Alcator C-Mod [14] and JET[15] havedemonstrated ICRF power can depletes the density in front of the LH antenna when the active
ICRF antenna is magnetically connected or adjacent to the LH antenna. As shown in Fig.2.,
spectroscopic observations revealed significant increases in metal impurity concentrations
during the RF pulse, with tungsten levels increasing by as much as a factor of four at the RF
powers about 1.0MW. This indicates that the tungsten limiters of LH are the major impurity
sources in the ICRF heating phases.

Figure 2 Observations of impurities behaviour are presented from ICRF heating experimentson the EAST, using graphite limiters and stainless steel antenna Faraday shields.

Figure 3 Surface temperature evolution on the right side limiter
The surface temperature elevation on the actively cooled side limiters of the LH launcher
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is monitored by infrared cameras. When the 4.6GHz LH launcher was powered alone. Hot
spots appeared in front of the waveguide rows on the left and right side limiter. When ICRH
was added from B-port antenna next to the LH launcher, both LH coupling and heat loads
were modified. The total power reflected coefficient averaged over all the LH modules increase
from 3% to 14%. Furthermore, as shown in Fig.3, the LH coupling and heat loads showed
different evolutions on the top and bottom parts of the launcher.

4. Interaction of fast wave and fast ions
On EAST, NBI and ICRF are usually combined in experiments requiring a large amountof heating power. The interaction between fast waves and the fast particles has been observedin two different types of experiments. The first type is the second harmonic of deuterium withhigher hydrogen concentration. The second type of experiments is third harmonic heating of a
D plasma. As shown in figure 4, when adding the ICRF in NBI heated plasma, the increase ofneutron can reach 150. But operating ICRH only, the increase in neutron is less than 30. A
possible candidate explanation would be that the beam ions are absorbing the RF power.

Figure 4 Examples of the EAST shots with combined NBI and second harmonic RF Heating
5.

Plasma Rotation driven by ICRF in LHCD target plasma

Figure 5 Counter Rotation driven by ICRH in the LHCD target plasma: small change inouter gap(left) and in the core plasma density(right).
Toroidal plasma rotation in a tokamak has attracted a great interest due to its effects on
plasma stability and confinement. On EAST, toroidal rotation changes due to ICRH have beenmeasured using a high resolution X-ray imaging crystal spectrometer[16]. Core toroidal
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rotation behavior with ICRF minority heating have been examined in the plasmas with andwithout Lower Hybrid Current Drive (LHCD) on EAST[17,18]. ICRH induced rotation changes
in both the co- and counter-current direction have been observed on EAST L-mode LHCD
plasmas. Here, we show an example of two discharges with very similar parameters. As shownin Fig.5, small outer gap change(left) and small density change(right) may lead to totallydifferent rotation behavior. The reason for this is still not clear. We found that the changes in
the profiles are almost the same both before and during ICRF phase. It seems to havecorrelation with current profiles.
Acknowledgements
This work was supported by: 1) National key research and development program (grant No.
2016YFA0400600 and 2016YFA0400601); 2) National Natural Science Foundation of China under grantNo. 11675213 and 111628509; 3) National Magnetic confinement Fusion Science Programme (grant No.
2015GB101001, 2015GB103002 and 2013GB106001B).

References[1] B.N. Wan, International Collaborators, Nucl. Fusion 49 (2009) 104011.
[2] J. Li, B. Wan, for the EAST Team and International Collaborators, Nucl. Fusion 51 (2011) 094007.
[3] Y.P. Zhao, X.J. Zhang, et al Fusion Engineering and Design 89 (2014) 2642–2646[4] L.N. Liu, X.J. Zhang, et al J. Plasma Phys. 85, 101017 (2019).
[5] N. Liu, X. J. Zhang, Y. B. Zhu et al., Rev. Sci. Instrum. 90, 101063 (2019).
[6] X.J.Zhao, L.N. Liu et al Nucl. Fusion 59 (2019) 044004[7] G. Urbanczyk, X.J. Zhang, et al Nucl. Fusion 59 (2019) 066023[8] X.J. Zhang, B.N. Wan, Y.P. Zhao, L. Hu, B.J. Ding, X.Z. Gong, et al., Invited Talk, 20th Topical Conferenceon RF Power in Plasmas, Sorrento, Italy, June, 2013.
[9] X.J. Zhang, Y.P. Zhao, Y.Z. Mao, S. Yuan, D.Y. Xue, L. Wang, et al., Plasma Sci. Technol. 13 (2011) 172.
[10] X.J. Zhang, Y.P. Zhao, B.N. Wan, X.Z. Gong, Y.Z. Mao, S. Yuan, et al., Nucl. Fusion 52 (2012) 032002.
[11] X.J. Zhang, Y.P. Zhao, B.N. Wan, X.Z. Gong, J.G. Li, Y. Lin, et al., Nucl. Fusion 53 (2013) 023004.
[12] A Ekedahl et al. AIP Conference Proceedings, pages 259–262, 2003.
[13] E H Kong et al. Plasma physics and controlled fusion, 54(10), 2012.
[14] G Wallace. Behavior of Lower Hybrid Waves in the Scrape Off Layer of a Diverted Tokamak. PhD thesis,
Massachusetts Institute of Technology, December 2009.
[15] K K Kirov et al. Effects of icrf induced density modifications on lh wave coupling at jet. Plasma physics andcontrolled fusion, 51(4), 2009.
[16] Lyu, et al., Rev. Sci. Instrum. 87 (2016) 11E326[17] Pan, et al., Plasma Sci. Technol. 18 (2016) 114[18] Lyu, et al, 25th IAEA FEC, EX/P3-5[19] Rice J.E., Greenwald M.J., Hutchinson I.H. et al., ‘‘Observations of central toroidal rotation in ICRF heated
Alcator C-Mod plasmas’’, Nuclear Fusion 38 (1998) 75.

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Conceptual design of CFETR ECRH system
X.J. Wang1, F.K. Liu1, Y.Y. Tang1, L.Y. Zhang1, D.J. Wu1, H.L. Wang1, W. Wei2, J.G. Li1
1

Institute of Plasma Physics Chinese Academy of Sciences, No.350 Shushan lake Road, 230031, Hefei,
China
2

Hefei University of Technology, No.193 Tunxi Road, 230009, Hefei, China

Abstract
The China Fusion Engineering Test Reactor (CFETR) is the proposed next-generation fusion facility in
China, during its two operational phases, a large fraction of externally driven current is foreseen to meet thescientific goals of the two operating scenarios for CFETR. The design of electron cyclotron resonanceheating and current drive (ECRH/ECCD) will be carried out with the requirements of physical objectives,
as a key design activity of auxiliary heating mix for CFETR. Based on the physical requirements and thetechnical availability, the frequency is chosen to be 170GHz in order to benefit from the technologies
development for ITER, the consideration and the preliminary design of the EC system will be presented inthis paper.
1. Introduction
The China Fusion Engineering Test Reactor (CFETR) is the next device for China magnetic controlled
fusion (CN-MCF) program which aims to demonstrate fusion energy production up to 200 MW initiallyand to eventually reach DEMO relevant power level 1 GW, to manifest a high duty factor of 0.3–0.5, and torealize the tritium self-breeding with tritium breeding ratio (TBR) >1. For meeting both phase I and II
targets with achievable technical solutions, a new design has been made by choosing a larger machine withmajor radius R = 7.2 m, minor radius a = 2.2 m, the toroidal magnetic field BT at major radius of 6.5 T for
both the physics and engineering designs [1-3].
Two operating scenarios which are self-consistent steady-state and hybrid mode scenarios for CFETR
have been developed by using a multi-dimensional code suite with physics-based models [4, 5]. A large
fraction of externally driven current is foreseen for achieving the two operating scenarios of CFETR, whichcalls for optimization of both plasma CD efficiency as well as the engineering integration design of the CD
systems. The design of electron cyclotron resonance heating (ECRH) is a key activity in order to achieve
the operational scenarios by mixing with other auxiliary heating systems (neutral beam injection (NBI) andlower hybrid current drive (LHCD)).
The main tasks assigned to the EC system are assisted start-up, plasma heating, current drive and
MHD instability suppression. The consequent requirements are used for the conceptual design of the EC
system. According to the physical integrated simulation, the ECW power is about 30 MW operating at
fundamental O mode with the frequency of 250 GHz, applied to drive ~0.88 MA to maintain the magnetic
shear in the core and to control q min > 2 to avoid any low n MHD mode. In the frame of the present designphase, the conceptual design of ECRH system is undertaking in order to comply with the engineering
integration requirement of CFETR as well as to exploit the key technologies for the critical components.
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The preliminary design progress along with the considerations of ECRH system are demonstrated in thispaper.
2. Overview of ECRH system design

Fig.1 The schematic layout of EC system.
The ECH system, as shown in Fig.1, is composed of gyrotrons, transmission lines, launcher andauxiliary subsystems, such as high voltage power supply, interlock & control, water cooling. In 2015, aversion of conceptual design of 170 GHz / 20MW ECRH system was made based on the small size CFETR,
in which the total 20MW power was injected through an equatorial window [6]. The challenges that ECRH
design encounter are both in technology and physics, advanced design will be foreseen due to the increased
Bt, particularly in the development of high frequency gyrotron (~250GHz), the physical needs for eachfunction are not fully defined. From the point of view of the integration machine design, the present
approach is to perform a feasible, integrated concept design with parameters of 30 MW /170GHz, which
will contribute to the interfaces iteration and engineering integration to comply with CFETR design
specifications and assessment the impact on the Tritium Breeding Ratio (TBR), safety and so on.
In this compromise, the selection of the RF source could benefit from the state of art of the megawatt
gyrotron for ITER. The arrangement of gyrotrons should be modular to satisfy design criteria on reliability,
availability, safety and maintenance. For the target 30MW injection power, 36 gyrotrons with theparameters of 1MW output power at 170GHz are proposed based on the evaluation of appropriate
redundancy. The RF source is split into 5 +1 backup clusters delivering power to the launchers.
3. R&D activities of transmission line
The generated power will be transmitted by the evacuated corrugated waveguide transmission lines.
Transmission lines should satisfy a series of technical requirements, therefore there are different microwave
components distributed along the line, such as MOU, corrugated waveguide, miter bends and some
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auxiliary devices. For the sake of vacuum maintaining, Chemical Vapor Deposition (CVD) diamondwindow together with fast gate valve will be equipped in each line. The external site allocation of CFETR
is not yet defined, some R & D activities are implemented for microwave components design (seen in
Fig.2), based on the EAST 140GHz /1000s ECRH system [7,8]. The performances of TL components will
be evaluated on the new test-bench at the frequencies of 140GHz and 170GHz.

Fig.2 R&D of some key transimission components.
4. Launcher
For the sake of meeting the engineering integration requirement of CFETR, the launcher design is inhigh priority. The front of the launcher is exposed to the plasma with harsh neutron damage. The launcher
configuration should be compact and robust in order to realize reliable continuous operation. Movable parts
are not preferred in facing of plasma. However, for the purpose of current profile control, the capability ofbeam steering is needed. Any launcher structure in CFETR needs to be compliant with other systems, ofmost importance are the breeding blankets (BB) and remote handling procedures. In order to facilitate theiteration with other engineering tasks, the launcher is currently design based on the front steering concept,
which is extensively used in current fusion devices.

Fig.3 Schematics of launcher concepts.
With regards to the limited equatorial ports, ECW power is anticipated to be top injected. The view oflaunchers installed in three top ports is shown in Fig.3. Two of the vertical launchers are used for current
drive with the maximum power of 16MW, two bunches with 8 beams each are accommodated. Anotherlauncher that emits a bunch of 7MW is considered for NTM control with wide-angle range. To investigate
the performance and optimize schemes of power injection for the design of the launcher, ray tracing
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simulation needs to be performed based on new plasma equilibrium for optimization launching parametersand EC operational regime [9].
As is shown in Fig.3, in this case, four breeding blanket modules are affected by the required openings
for beam penetrations of each launcher. The first estimate of the decrease in Tritium Breeding Ratio(ǻ7%5)
and the induced neutron heat flux on launcher components and toroidal field coils is under calculating. The
impacts are depending on the design of the launcher and its integration in the blanket, for saving TBR, theoptical routes of beam-line are further optimized to minimize the openings area of blankets. Fig.4
demonstrates the modified transmitting routes of Gaussian beam which results in embedding the lastreflecting mirror into the blankets. The occupied area can be slightly reduced by modification in the
launched beam.

Fig.3 EC Launchers integration in breeding blankets .

Fig.4 Alternative optical routes for top launcher to reduce the opening area on BB .

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5. Summary
The conceptual design of 30MW 170GHz ECH system for CFETR is presented, mainly concernedwith the technical feasibility assessment to comply with the integration requirement of CFETR as well as topursue technical solutions for critical components. As the project is in the initial stage, efforts are focus on
the design of launcher, several beam injection options for front steering top launcher are under assessment
taking into account the guidelines for the integration with other systems. The crucial issues of multi-beamlauncher will be addressed further. Detailed design is anticipated to be conducted which could provide abaseline design with maximal achievable reliability and availability in the near future.
Acknowledgements
This work was partly supported by the Post-CUP Program. It was also supported by National Key
R&D Program of China (grant nos. 2017YFE0300500 and 2017YFE0300503).

References[1] G. Zhuang, et al., Nucl. Fusion 59 (2019) 112010.
[2] Y.X. Wan, et al., Nucl. Fusion 57 (2017) 102009.
[3] J. Li, et al., J. Fusion Energy 38 (2018) 113-24.
[4] X. Jian, et al., Nucl. Fusion 57 (2017) 046012.
[5] J. Chen et al., Plasma Phys. Control. Fusion 59 (2017) 075005.
[6] Y.Y. Tang, et al., Fusion Engineering and Design 94 (2015) 48–53.
[7] D. J. Wu, PHD dissertation, University of Science and Technology of China (2019).
[8] D. J. Wu, et al., IEEE Access 5 (2017) 6187-6191.
[9] W. Wei, et al., Plasma Sci. Technol. 21 (2019) 065101.

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Simulation of energetic particle driven geodesic acoustic mode channeling in LHD using
MEGA code
Hao WANG1, Yasushi TODO1, Masaki OSAKABE1,
Takeshi IDO1, and Yasuhiro SUZUKI1,2
1

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

2Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan

Abstract
Energetic particle driven geodesic acoustic modes (EGAMs) in the Large Helical Device (LHD) plasmas
are investigated using MEGA code. MEGA is a hybrid simulation code for energetic particles (EPs)interacting with a magnetohydrodynamic (MHD) fluid. The low frequency EGAMs are reproduced. Also,
the energy transfer is analysed and the bulk ion heating during the EGAM activity is observed. The ionsobtain energy when the EPs lose energy, and this indicates that an energy channel is established by EGAM.
The EGAM channelling is reproduced by simulation for the first time. The heating power of bulk ions is 3.4
kW/m3 which is close to the value 4 kW/m3 evaluated from the experiments. It is found that the sideband
resonance is dominant during the energy transfer from EGAM to the bulk ions, and the transit frequencies ofresonant bulk ions are half EGAM frequency.
1. Introduction
Geodesic acoustic mode (GAM) is an oscillatory zonal flow coupled with density and pressure
perturbations in toroidal plasmas[1-2]. In the past years, energetic particle driven GAM (EGAM) has been
observed in JET, DIII-D, Large Helical Device (LHD), ASDEX-Upgrade, and HL-2A[3-9]. In the DIII-D
experiment, the drops in neutron emission during the EGAM bursts suggest beam ion losses. Also, in the
LHD experiment, anomalous bulk ion heating during the EGAM activity suggests an energy channel createdby EGAM. In addition, EGAMs interact with turbulence and affect the plasma confinement. Then, theunderstanding of the EGAMs is important for magnetic confinement fusion, because the energetic particles(EPs) need to be well confined and the bulk plasma heating needs to be improved. At present, the existenceof EGAM energy channel has not been demonstrated by simulation. Also, it is quite conceivable that the bulk
ions obtain energy from EGAM via Landau damping, but it is not clear which resonant bulk ions are thedominant. The present proceeding solves the above problems and gives clear evidence in the following
sections.
2. Simulation model and parameters
The simulation of the EGAM channelling is conducted with MEGA code[10-11], and the simulation modelis the same as Ref. [12]. In the extended model, both the bulk ions and EPs are described kinetically. The
following equations in the extended model are different from that in the conventional model:

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A realistic 3-dimensional equilibrium generated by HINT code is used for the simulation[13]. Thisequilibrium data is based on the LHD shot #109031 at time t = 4.94 s. At this moment, the EGAM activity isvery strong, thus it is good to reproduce the EGAM phenomenon. The parameters for the EGAM simulation
are based on an LHD experiment[5-6]. The injected neutral beam energy is E NBI = 170 keV. For simplicity,
a slowing-down EP distribution is assumed in the proceeding. The bulk ion distribution is considered as
Maxwellian.
3. Simulation results
MEGA code was applied to the LHD experiment. Figure 1(a) shows the spatial profile of the simulatedmode. This mode peaks around r/a = 0.2, the mode width is about 0.4a. The simulated mode is identified asa global mode, which is consistent with the global structure of EGAM. The dominant mode number is m/n =0/0 for poloidal velocity as represented by the red curve in Fig. 1(a). Also, the dominant mode number is m/n= 1/0 for pressure perturbation. These mode numbers are also consistent with the nature of EGAM. Finally,
the magnetic perturbation is much weaker than the poloidal velocity perturbation, and this indicates that it isan electrostatic mode. This is also consistent with the nature of EGAM. Thus, based on the above 3 properties,
the simulated global electrostatic mode is identified as an EGAM. This the first time to reproduce an EGAM
using the extended model of MEGA code. In addition, the simulated EGAMs’ linear frequencies are between
40 kHz to 50 kHz as shown in Fig. 1(b), lower than the theoretically predicted conventional GAM frequency
54 kHz, thus these are low frequency branches of EGAM. The mode frequency decreases with the increasingof EP pressure, and this is consistent with the EGAM property.

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Fig. 1 (a) Poloidal velocity profile of the EGAM. (b) The EP pressure dependence of EGAM frequency inthe linear growth phase.
A typical case is simulated to demonstrate the EGAM channelling phenomenon as shown in Fig. 2. Figure2(a) shows the frequency spectrum of simulated EGAM. The mode frequency in linear stage is 50 kHz, andthen, frequency chirps up in the nonlinear phase with time evolution. At t = 0.5 ms, the frequency has alreadyexceeded 60 kHz. Fig. 2(b) shows the time evolution of EGAM amplitude vșThe linear stage is from t = 0
to about t = 0.1 ms. At t = 0.1 ms, the mode amplitude reaches the maximum value, and then, steps into thenonlinear phase. Fig. 2(c) shows the energy transfer of various species. The product of the perpendicularcurrent of each species and the electric field is integrated in space and time. The perpendicular current consistsof grad-B and curvature drift current, and magnetization current. The bulk ion heating during the EGAM
activity is observed. The ions obtain energy when the EPs lose energy, and this indicates that an energy
channel is established by EGAM. The EGAM channelling is reproduced by simulation for the first time.
From t = 0 to t = 0.36 ms, the energy transferred from EP is 63 J. About half of this energy (51%) is transferred
to bulk ions (34%) and electrons (17%), while another half is dissipated. The heating power of bulk ionsaround t = 0.1 ms is 3.4 kW/m3 which is close to the value 4 kW/m3 evaluated from the experiments[5].

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Fig. 2 (a) The frequency spectrum of EGAM. (b) The time evolution of EGAM amplitude vș(c) Energy
transfer of various species during EGAM activity.
In order to investigate the resonant particles, the įf distribution of both EPs and bulk ions at different times
are analysed in the particle transit frequency space, as shown in Fig. 5. The particle transit frequency f tr isdefined by f tr = v||/(2ʌqR0), where v|| is particle parallel velocity, q is the safety factor value, and R0 = 3.75
m is plasma major radius. For simplicity of f tr calculation, q = 2.8 is a constant for all the particles in theabove f tr equation, although q value has a normal shear profile in the equilibrium of simulation. The EGAM
obtains energy from EPs via inverse Landau damping, and EGAM transfer energy to bulk ions via Landau
damping. These processes modify the distribution function of both EPs and bulk ions, and thus, įf values
change with mode evolution. The negative įf value forms the hole structure in phase space, and the positive
įf value forms the clump structure. Large absolute įf values indicate strong interactions between EGAM andresonant particles. Fig. 5(a) shows the resonant EPs. A hole around f tr = 50 kHz is formed. The transit
frequencies of particles in the hole increase with time evolution, and this increase of frequency is kept
consistent with the chirping up of EGAM frequency. The resonance condition between EGAM and EPs isgiven by f EGAM = f tr,EP . Fig. 5(b) shows the resonant bulk ions. Two groups of clumps around f tr = 25 kHz
and f tr = 5 kHz are formed. The transit frequencies of bulk ions in these clumps increase with time evolution,and these transit frequencies are kept at half the EGAM frequency (and one-tenth of EGAM frequency). The
resonance condition between EGAM and bulk ions is given by
f EGAM = l·f tr,bulk,(6)and dominant l values are l = 2 and l = 10. This is the first time to quantitatively reveal the resonance condition
between EGAM and bulk ions during the establishment of EGAM channelling.

Fig. 3 The įf distribution of (a) EPs and (b) bulk ions in f tr phase space at t = 0.18 ms (red), t = 0.36 ms(green), and t = 0.58 ms (blue).
In Eq. (6), l = 2 is important because l = 2 sideband resonance condition is much easier to be satisfied thanhigher l value cases. However, l = 10 corresponds to very low transit frequency ions, and these low transit
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frequency ions should be difficult to obtain energy from EGAM. In order to confirm that, the energy transfer
rate of bulk ions in f tr phase space is analysed as shown in Fig. 4. There is a peak around f tr = 25 kHz, andthis peak gradually moves rightward. Similar to Fig. 3(b), this rightward moving indicates that the bulk ionswith half mode frequency are kept resonant with the mode. The energy transfer rate is positive, and thisindicates that the bulk ions absorb energy from EGAM. In Fig. 4, the red curve at t = 0.109 ms represents atransition between linear growth phase and nonlinear frequency chirping phase, this can also be confirmed
from Fig. 2(b). From t = 0.145 ms in the fully nonlinear phase, another peak appears around f tr = 15 kHz. In
this simulation, the bulk ion temperature is 4.85 keV, and this thermal velocity corresponds to a transit
frequency 14.7 kHz. The peak around f tr = 15 kHz appears in Fig. 6 because most bulk ions’ transit
frequencies are around 15 kHz. In Fig. 3(b), there is a peak around 5 kHz, but this peak is very weak anddifficult to identify in Fig. 4, because the particles around 5 kHz do not absorb too much energy. The bulk
ions mainly absorb energy via the particles whose transit frequencies are half the EGAM frequency. In Eq.
(6), l = 2 is more important than l = 10 for the establishment of EGAM channeling.

Fig. 6. Energy transfer rate of bulk ions in f tr phase space at different times.
4.

Summary and conclusions

A global electrostatic mode is reproduced using extended model, and the mode is identified as an EGAM.
Both the mode number and the mode frequency are consistent with the theoretical prediction. The ions obtainenergy when the EPs lose energy during the EGAM activity, and this indicates that an energy channel isestablished by EGAM. The EGAM channelling is reproduced by simulation for the first time. The heating
power of bulk ions is 3.4 kW/m3 which is close to the value 4 kW/m3 evaluated from the experiments. The
įf distribution of both EPs and bulk ions at different times are investigated, and the transit frequencies ofresonant particles are analysed. Also, the energy transfer rate of bulk ions at different times are investigatedand compared with įf distribution. The resonance condition f EGAM = l·f tr,bulk is satisfied where the dominant
l values is l = 2. Another resonance with l = 10 was also found. The resonance condition between EGAM and
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bulk ions during the establishment of EGAM channelling is quantitatively revealed for the first time.
Acknowledgements
This work was partly supported by the Post-CUP program. Numerical computations were performed onthe ‘Plasma Simulator’ (FUJITSU FX100) of the National Institute for Fusion Science (NIFS) with thesupport and under the auspices of the NIFS Collaboration Research program (NIFS17KNST111,
NIFS18KNST136, NIFS18KNXN365, and NIFS19KNXN397), and the ‘K Computer’ of the RIKEN
Advanced Institute for Computational Science (Project ID: hp170260 and hp180200). This work was partly
supported by MEXT as a ‘Priority Issue on Post-K Computer’ (Accelerated Development of Innovative Clean
Energy Systems), JSPS KAKENHI Grant Nos. JP15K06652, JP18K13529, and JP18H01202. The authors
thank Professor H. Sugama, Professor K. Toi, Dr K. Shinohara, and Professor D. Zarzoso for fruitful
discussions.

References[1] DIAMOND, P. et al., Zonal flows in plasma—a review, Plasma Phys. Control. Fusion 47 (2005) R35-R161.
[2] SUGAMA, H. and WATANABE, T., Collisionless damping of zonal flows in helical systems, Phys. Plasmas, 13 (2006)
012501.
[3] BOSWELL, C. et al., Observation and explanation of the JET n = 0 chirping mode, Phys. Lett. A 358 (2006) 154-158.
[4] NAZIKIAN, R. et al., Intense Geodesic Acousticlike Modes Driven by Suprathermal Ions in a Tokamak Plasma, Phys.
Rev. Lett. 101 (2008) 185001.
[5] OSAKABE, M. et al., “Indication of bulk-ion heating by energetic particle driven Geodesic Acoustic Modes on LHD”,
Paper No. EX/10-3, IAEA Fusion Energy Conf., St. Pertersburg (2014).
[6] IDO, T. et al., Identification of the energetic-particle driven GAM in the LHD, Nucl. Fusion 55 (2015) 083024.
[7] IDO, T. et al., Strong Destabilization of Stable Modes with a Half-Frequency Associated with Chirping Geodesic Acoustic
Modes in the Large Helical Device, Phys. Rev. Lett. 116 (2017) 015002.
[8] HORVATH, L. et al., Experimental investigation of the radial structure of energetic particle driven modes, Nucl. Fusion
56 (2016) 112003.
[9] CHEN, W. et al., Observation of energetic-particle-induced GAM and nonlinear interactions between EGAM, BAEs andtearing modes on the HL-2A tokamak, Nucl. Fusion 53 (2013) 113010.
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distributions, Phys. Plasmas, 13 (2006) 082503.
[11] TODO, Y. et al., Comprehensive magnetohydrodynamic hybrid simulations of fast ion driven instabilities in a Large
Helical Device experiment, Phys. Plasmas, 24 (2017) 081203.
[12] TODO, Y., “A new magnetohydrodynamic hybrid simulation model with thermal and energetic ions”, Paper No. O9, Int.
Toki Conf. and Asia Plasma Fusion Association Conf., Toki (2017).
[13] SUZUKI, Y. et al., Development and application of HINT2 to helical system plasmas, Nucl. Fusion, 46 (2006) L19-L24.

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Evaluation method of effective ion charge (Z eff ) in LHD plasma
Y. Kawamoto1, M. Goto1,2, T. Oishi1,2, M. Osakabe1,2 and T. Morisaki1,2
1National Institute for Fusion Science, Toki 509-5292, Gifu, Japan2

Department of Fusion Science, Graduate University of Advanced Studies, SOKENDAI,
Toki 509-5292, Gifu, Japan

Abstract
The measurement of effective ion charge ܼeff in LHD is reconsidered. The plasma equilibrium database,the so-called TSMAP, recently made available is used for mapping of locations on the line-of-sight on theeffective minor radius ‫ݎ‬eff so that accuracy of the electron temperature and density profiles for evaluating
the Bremsstrahlung intensity has been improved. The influence of impurity lines is also estimated from otherdiagnostic data, and it is demonstrated that unrealistically large ܼeff values derived with the conventional
method for low density plasmas are avoidable.
1. Introduction
The effective ion charge ܼeff is a fundamental parameter for the plasma confinement study. Although
efforts have been made for providing accurate ܼeff for all the discharges in the Large Helical Device (LHD)[1, 2], it has been always a concern that magnetic surface structures are changed when the plasma pressure
becomes high which causes inconsistency in deriving ܼeff .
ܼeff is evaluated from the Bremsstrahlung intensity, which is in many cases measured by a lineintegrated observation. Evaluation of ܼeff therefore needs profiles of the electron temperature ܶe andelectron density ݊e along the line-of-sight of the Bremsstrahlung measurement. In LHD, the ܶe and ݊eprofiles are obtained by the Thomson scattering measurement along the major radial direction on thehorizontally elongated plasma cross section. For obtaining the ܶe and ݊e values on an arbitrary line-ofsight, mapping of locations on the line-of-sight to the locations on the axis of the Thomson scattering
measurement is necessary, and this mapping requires information on the plasma equilibrium state. Recently,
the TSMAP project has started to provide the plasma equilibrium data for all the discharges in LHD [3], andthey can be used for the ܼeff measurement.
Another problem in deriving ܼeff from the Bremsstrahlung intensity is contamination of impurity lines
in the signals. In the evaluation of ܼeff in LHD to date, ܼeff shows a skyrocketing behavior when plasma
density is decreasing, which is believed to be due to the influence of a collection of small impurity lines. We
have attempted to exclude such impurity lines in the evaluation of ܼeff by quantifying the influence of such
impurity lines with the help of other diagnostic data. This paper reports our efforts for such improvements inevaluation of ܼeff for LHD.
2. Experimental setup
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The measurement of Bremsstrahlung intensity is being made for all the LHD discharges. The radiation
is observed by an optical fiber having a core diameter of 300 Ɋm with a collimator lens. The line-of-sight

Fig. 1: Vertical cross section of the plasma which includes line-of-sight of the present measurement.
is shown with the plasma cross section in Fig. 1. The collected light passes through an interference filter
which limits the transmittable wavelength range in a narrow band width, and is guided to a photomultipliertube (PMT, HAMAMATSU H5784-01MOD). The central wavelength and FWHM (full width at half
maximum) of the transmittance profile of the filter are 536.5 nm and 6 nm, respectively, and the profile isshown with the black dashed line in Fig. 2. The signal from the PMT is digitized with 12 bit dynamic range
and recorded with a 10 kHz sampling rate. The measurement with this observation system is referred to asVisible spectrometerinterference filter
trancemittance









Bremsstrahlung





3

REV

B









O 3 (O)>:P QP @the "intensity measurement" in later discussions.







O>QP@



Fig. 2: Example of spectra by the spectral measurement (red dots and solid line), transmittanceprofile of the interference filter (black dashed line), and synthetic Bremsstrahlung spectrum(blue solid line).
For a supplementary purpose, we sometimes made observation of the Bremsstrahlung spectrum with thesystem composed of a UV-visible spectrometer and a CCD (charge coupled device) detector. An almost
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identical line-of-sight to that for the intensity measurement is used for this observation. The spectrometer,
Chromex 500is, has a focal length of 0.5 m and is equipped with three gratings. A grating of 100 grooves/mm
out of three is used for the present measurement. The reciprocal linear dispersion is 20.0 nm/mm. The CCD
detector (Andor DV-420) has 1024 × 255 pixels with pixel size of 26 Ɋmଶ.
The optical fiber image on the detector is extended over several pixels in the vertical direction, which isperpendicular to the wavelength dispersion. All the signals within such extended fiber image are accumulated
and recorded as an intensity signal at the corresponding wavelength position. The time resolution of thismeasurement is 200 ms. The absolute sensitivity has been calibrated for this system with an integrating sphere(USS-600C). The measurement with this system is hereafter called the "spectral measurement."
3. Theoretical Bremsstrahlung intensity
In the present study, ܼeff is assumed to be constant and has no spatial profile in the plasma. In that case,ܼeff is evaluated as a ratio of the observed and synthetic Bremsstrahlung intensities, for the latter a purehydrogen plasma is considered. The synthetic Bremsstrahlung spectrum ܲB (ߣ), where ߣ is the wavelength,
is given asܲB (ߣ) = ܼeff ‫׬‬Here, ‫׬‬

LOS

LOS

ߩB (ߣ) ݀‫ݏ‬[W mିଷ ],(1)

indicates that the integration is conducted along the line-of-sight, and the local quantity ߩB (ߣ)

is explicitly written asߩB (ߣ) =ଶల ξଶ గభΤమ ఈ௛௔బమ ோమଷξଷ௠ యΤమ ௖

ଵ

௛௖Τఒ

౛

e

ଵ

ට௞் exp ቀെ ௞் ቁ ‫݃ۃ‬ff (ܶe , ߣ)‫݊ ۄ‬eଶ ఒమ[W mିସ ],(2)

where ߙ, ݄, ܽ଴ , and ܴ are the fine structure constant, Planck constant, Bohr radius, and Rydberg constant,respectively, ݉ and ܿ are the electron mass and the light speed, respectively, and ‫݃ۃ‬ff (ܶe , ߣ)‫ ۄ‬is thethermally averaged Gaunt factor for free-free transitions.
For local ܶe and ݊e values which are necessary to evaluate Eq. (2), the data measured by the Thomson
scattering diagnosis are used. Any location along the line-of-sight can be mapped onto the effective minor
radius ‫ݎ‬eff by the TSMAP database, and ܶe and ݊e values at arbitrary ‫ݎ‬eff can be obtained by interpolation.
The values of ‫݃ۃ‬ff (ܶe , ߣ)‫ ۄ‬are calculated with the method of Ref. [4]. Numerical integration is made for Eq.
(1) with these parameters.
4. Results and discussion
Although for the intensity measurement we have chosen the wavelength range by the interference filter
in which no noticeable line radiation is found, it is known that the signal always contains influence of small
impurity lines, and their fraction becomes noticeably large when ݊e is low so that the Bremsstrahlung
intensity is relatively weak.
For a quantitative evaluation of such impurity lines fraction, we use the data obtained by the spectral
measurement. An example of the measured spectra is shown with the red dots in Fig. 2. This spectrum
contains the Bremsstrahlung and impurity lines. We try to distinguish the Bremsstrahlung part from the

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impurity lines by comparing a synthetic spectrum evaluated by Eq. (1) with the measured spectrum.
In this comparison, ܼeff in Eq. (1) is increased from unity so long as no experimental value lies below
the corresponding synthetic value, and the finally obtained maximum value is here regarded as ܼeff of thecorresponding plasma. The Bremsstrahlung spectrum evaluated with ܼeff obtained this way is drawn withthe blue solid line in Fig. 2. It is noted that ܼeff derived here should be therefore regarded as the upper limit
rather than the best estimate.
Because the spectral measurement can be made only for a limited number of discharges, the method ofܼeff evaluation described above is inappropriate to be used as a routine plasma diagnosis in LHD. Instead,
we consider to make corrections for the intensity measurement data with the help of other diagnostic data.
The black dotted line in Fig. 2 shows the transmittance profile of the interference filter ܶ(ߣ) as
mentioned above. What we obtain by the intensity measurement ‫ܫ‬obs can be understood as the measured
spectrum ܲobs (ߣ) multiplied with ܶ(ߣ) and integrated over the wavelength, i.e.,‫ܫ‬obs = න ܲobs (ߣ)ܶ(ߣ) ݀ߣ .

(3)

The Bremsstrahlung intensity ‫ܫ‬B can be similarly evaluated from the synthetic Bremsstrahlung spectrum
ܲB (ߣ) as
LOS‫ܫ‬B = න ܲB (ߣ)ܶ(ߣ) ݀ߣ = ܼeff න ቆන

ߩB (ߣ) ݀‫ݏ‬ቇ ܶ(ߣ) ݀ߣ,(4)

where ܼeff value derived in the above analysis is used. The difference between ‫ܫ‬obs and ‫ܫ‬B would be thecontribution of the impurity lines ‫ܫ‬imp . In LHD, the impurity monitoring system measures several impurity
line intensities for watching impurity contaminations in the plasma. Figure 3 shows ‫ܫ‬imp against the carbon
ion line (CIII, 97.7 nm) intensity ‫ܫ‬CIII from the impurity monitoring system for various discharges used in
the present study. It is clearly seen that ‫ܫ‬imp has a strong correlation with ‫ܫ‬CIII , and ‫ܫ‬imp is well expressed+\GURJHQSODVPD-2[Wcm ]





I

imp



\ H [A  5 












I [arb.unit]





CIII

Fig. 3: Correlation between ‫ܫ‬imp and ‫ܫ‬CIII.
with ‫ܫ‬CIII as‫ܫ‬imp = 9.235 × 10ିହ (‫ܫ‬CIII )଴.ଶସହଶଽ ,
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(5)

and this relationship is expected to be valid for all the LHD discharges. Because ‫ܫ‬CIII is always measured by
the impurity monitoring system, ‫ܫ‬imp can be also evaluated from Eq. (5) without the spectral measurement.
‫ܫ‬B is then obtained as ‫ܫ‬obs െ ‫ܫ‬imp and ܼeff is evaluated with Eq. (4). Figure 4 shows an example of theresults which shows the temporal variation of ܼeff for a discharge with increasing ݊e . In this figure, theresult without correction of impurity lines is also shown, and it is confirmed that the correction by the present
method works effectively. Finally, we conclude that the evaluation method of ܼeff has been significantly
improved and the reliability is high enough for being used for the various LHD plasma analyses.



㼚㼑
=

UDZ

HII





㼓㼍㼟㻌㼜㼡㼒㼒㻔㻴㻞㻕




D[

=

HII







FRUUHFWHG




W>V@







H

=

HII





P

W75%Q > P @Fig. 4: Example of derived ܼeff for a discharge with increasing ݊e. The result without impurity
lines correction is also shown.
Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by NIFS budget code
UPLL010. The help of LHD experiment group for conducting the experiment is also acknowledged.
References[1] H. Y. Zhou et al., Jpn. J. Appl. Phys. 49 (2010) 106103.
[2] X. L. Huang et al., Plasma and Fusion Research 10 (2015) 3402036.
[3] M. Emoto et al., Fus. Eng. Des. 87 (2012) 2076.
[4] P. J. Storey and D. G. Hummer, Comp. Phys. Comm. 66 (1991) 129.

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Observation of tungsten spectra from W5+-W45+ in EAST
L. Zhang1, S. Morita2, Z. Xu3, Y. X. Cheng1,4, X. D. Yang1,4, L. Li1,4, Z. W. Wu1, Y. Yang5,
X. B. Ding6, T. Oishi2

1 Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230026, China
2
3
4National Institute for Fusion Science, Toki 509-5292,Gifu, Japan

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

Science Island Branch of Graduate School, University of Science and Technology of China, Hefei 230026,
China
5

6

Institute of Modern Physics, Fudan University, 220 Handan Road, Shanghai 200433, China

Key Laboratory of Atomic and Molecular Physics and Functional Materials of Gansu Province, Northwest
Normal University, Lanzhou 730070, China

Abstract
In the present work, the tungsten spectra at 20-150Å from W24+-W45+ions are analyzed for L-mode
plasmas (T e0 =1.4keV), typical H-mode plasmas (T e0 =2.6-3.0keV) with RF-heating and with NBI heating.
The analysis is also done for plasmas with tungsten accumulation (T e0 ~3.0keV). Recently, identificationsof EUV spectra from weakly ionized tungsten ions of W4+-W7+ are being newly attempted based on EAST
plasmas with extremely low temperature, e.g. T e0 <=0.2keV. Preliminary results are presented in theproceeding.
1. Introduction
In order to improve the capability of mitigating heat load of the divertor target, graphite divertor plates
at upper divertor in the EAST tokamak were replaced by tungsten monoblock plates [1]. Since then, in the
EAST experiment, studies on tungsten behavior and its effect on the plasma performance are crucially
important for maintaining high-performance plasmas for a long period. For the purpose fast-time-response
and space-resolved extreme ultraviolet (EUV) spectrometers [2-4], which are workable at 10-130Å,
20-500Å and 30-520Å, respectively, have been developed to observe line emissions from tungsten ions andto study the tungsten behavior in the EAST plasma.
2. Tungsten spectra at 40-140Å
Tungsten spectra at 40-140Å are observed in many different plasma scenarios in EAST. In this session,
the typical tungsten spectra at 40-140Å observed in five different cases are compared, i.e., L-mode phasewith tungsten dust dropping (#80565), NBI and LHW dominant H-mode phase (#80575 and #80093), and
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the NBI and LHW dominant H-mode phase with tungsten accumulation (#70769 and #83880). The electron
temperature in central plasma, T e0 , plasma density, n e , heating method, and plasma condition and tungsten
source for the plasma in these five discharges are listed in Table 1.
Figure 1(a)-(e) shows the tungsten spectra in wavelength range of 40-140Å measured in five different
plasma scenarios described in Table 1. As shown in Figure 1 that there are several W-UTA (Tungsten
Unresolved Transition Array) existing in this wavelength range, i.e., 45-55Å, 55-72Å, 78-82Å. Additionallyserval isolated tungsten lines at the wavelength range of 120-140Å could be observed. It could be foundthat W-UTA near 50 Å composed of W24+ to W34+ ions is always observed with strongest intensity, even inhigh T e plasmas with T e0 Ӌ4.0keV. And the highest ionization stage of tungsten ions depends on T e0 . forexample, the highly ionized tungsten ions of W37+ - W45+ are absent in low T e case of T e0 =1.4keV. In
general, in the NBI dominant H-mode discharges the observed signal noise in the spectra (see Figure 1(b))
is much larger than that in RF heated H-mode discharges (see Figure 1(c)) due to the existence of energetic
particles, which will reduce the certainty of identification of weak tungsten lines. And it is obvious thatstrong or saturated tungsten spectra are observed during tungsten accumulation phase in Figure 1(d) and
Figure 1(e).

Table 1. Comparison of plasma scenarios forobservation of the tungsten spectra

Figure 1. Comparison of W spectra in 40-140 Å for five plasma scenarios

Figure 2 presents the detailed line identification in the spectra at wavelength range of 40-140Å
observed in typical RF-heated H-mode based on the NIST database. It could be noted that the W43+-W45+
lines with strong intensity are identified from W-UTA at 55-72Å. And W40+-W45+ lines with weak intensity
is identified at longer wavelength range of 120-137 Å. There are six lines with 4p-4s transition which arevery important for quantitative analysis [4], i.e., W43+ (E i =2.210keV) at 61.334 Å and 126.29Å withtransition of 4s24p-4s24s, W44+ (E i =2.354keV) at 60.93 Å and 132.88Å with transition of 4s4p-4s2, W45+
(E i =2.414keV) at 62.336 Å and 126.998Å with transition of 4p-4s. Unfortunately, the line of 132.88 Å isblended with Fe XXIII at 132.91 Å as shown in Figure 2(c).

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Figure 2. Identification of W spectra in 40-140 Å

Figure 3. Waveform of EAST discharge #83022

3. Observation of EUV spectra from weakly ionized tungsten ionsObservation of lines from weakly ionized W ions in normal discharges is important because theintensity can suggest tungsten influx rate. Figure 3 shows the waveform of the discharge (#83022).
During this discharge minor disruption occurred at t=5.143s when LHW heating is switched to ICRF
heating, and tungsten sputtering may be enhanced at tungsten guard limiter of LHW antenna. Strong line
intensity of WVII lines (216.219 and 261.35 Å) was observed during 5.143-5.239s (20 CCD frames in total)
indicating increased tungsten influx. Plasma was cooled down quickly with large reduction of T e0 from
1.5keV to about 0.2keV. And run-away electrons were triggered at tӋ5.239s.
Figure 4 shows the time-developed EUV spectra at longer wavelength range of 195-425 Å during
t=5.143-5.239s. Several lines from W5+ and W6+ ions are identified and indicated based on previous
experiments in SSPX spheromak [5] and HL-2A tokamak [6]. Identification of more emission lines from
W7+, W6+, W5+ and W4+ ions based on NIST database is indicated with red, pink, orange and light green
arrows in Figure 5(c).

Figure 5. Identification of lines from W4+-W7+

Figure 4. Time-developed EUV spectra at 195-425 Å

4. Summary and future work
In this work, tungsten spectra at 40-140Å from W24+-W45+ ions are analyzed for L-mode plasmas,
H-mode plasmas with RF- and NBI heating, and plasmas with tungsten accumulation. Observations of
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EUV spectra from weakly ionized tungsten ions of W4+-W7+ are being newly attempted, based on EAST
plasmas with extremely low temperatures.
In the near future the second fast EUV spectrometer working at wavelength range of 20-500 Å will be
installed with observation chords passing through edge plasma and the simultaneous observation of EUV
spectra at full range of 10-420Å will be realized. And the accuracy of tungsten line identification at longer
wavelength range of 195-425 Å will be increased with the assistance of experimental results from EBIT
device.
Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by Users with
Excellence Program of Hefei Science Center CAS (No. 2019HSC-UE014), National Natural Science
Foundation of China (Grant Nos. 11575244, 11775269) and National Key Research and Development
Program of China (No. 2018YFE0311100).

References[1] B. N. Wan et al., Nucl. Fusion 55 (2015) 104015[2] L. Zhang et al., Nucl. Mater. Energy 12 (2017) 774-778[3] L. Zhang et al., Rev. Sci. Instrum. 86 (2015) 123509[4] L. Zhang et al., Nucl. Instrum. Meth. A916 (2019) 169[5]J. Clementson et al., J Phys B 43(2010)14409[6] C. F. Dong et al., Nucl. Fusion 59(2019)016020

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Tungsten Measurements at LHD using Visible M1 Lines
D. Kato1,2, S. Era2, H.A. Sakaue1, I. Murakami1,3, M. Goto1,3, T. Oishi1,3, S. Morita1,3,
N. Nakamura4, K. Fujii5, F. Koike6, X.-B. Ding7, C.-Z. Dong7
1

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

2

Dept. Advanced Energy Engineering Sciences, Kyushu University, Kasuga 816-8580, Fukuoka, Japan
3Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan
4

Inst. for Laser Science, The Univ. of Electro-Communications, Chofu 182-8585, Tokyo, Japan

5

Graduate School of Engineering, Kyoto University, Kyoto Nishikyo-ku 615-8246, Kyoto, Japan
6

Faculty of Science and Technology, Sophia University, Chiyoda-ku 102-8554, Tokyo, Japan

7Key laboratory of Atomic and Molecular Physics & Functional Materials of Gansu Province, College ofPhysics and Electronic Engineering, Northwest Normal University, Lanzhou, 730070, China

Abstract
Visible emission lines of tungsten ions have a potential usefulness for tungsten measurements at ITER
and future DEMO reactors. Near ultra-violet (UV) and visible lines of highly charged tungsten ions havebeen reported at Large Helical Device (LHD). In the last LHD campaign, we experimentally observed local
intensity change of the near-UV magnetic-dipole (M1) line emission from the ground state W27+ ionscorresponding to local electron temperature change by superimposing electron cyclotron heating (ECH) on
neutral beam injection heating. At elevated temperatures (about 3 keV) obtained by the ECH, we found a
new near-UV emission line of a higher charge state of tungsten ions.
1. Introduction
Tungsten, which is used for divertor materials in ITER, has good properties as a plasma facing
component such as high thermal conductivity, small thermal expansion, low erosion yield, and less tritium
retention [1]. However, once tungsten accumulates in the core plasma, large radiative power losses oftungsten highly charged ions will limit the plasma performance [2,3]. Tungsten concentration in the coreplasma of ITER is a key issue for sustainable operation of high-performance operation with tungsten invessel components.
Visible emission lines from the tungsten ions provide useful diagnostics means for ITER and future
DEMO reactors because the radiation shielding of detectors is not necessary by using optical fibers. However,spectroscopic observations of tungsten highly charged ions in the core plasmas of existing large devices [48] have been limited in soft X-ray, Extreme-Ultra-Violet (EUV) and Vacuum-Ultra-Violet (VUV) regions.
Since visible magnetic-dipole (M1) lines of iron highly charged ions in the ground states are observed insolar flares, its potential usefulness for fusion plasma diagnostics has been pointed out [9]. Such visible M1line emission of highly charged tungsten ions in high-temperature magnetically confined plasmas wasobserved for the first time by tungsten pellet injection experiments at Large Helical Device (LHD) [10].
However, the highest charge state definitely identified with the visible M1 lines at LHD has been W28+ which
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is abundant at electron temperatures of about 1 keV [11]. In the present work, near-UV and visible emission
lines of higher charge states of tungsten ions are investigated at elevated electron temperatures bysuperimposing 154 GHz electron cyclotron heating (ECH) on neutral beam injection (NBI) heating.
2. Tungsten pellet injection and visible spectroscopy at LHD
Tungsten is introduced into the LHD plasmas by injecting a carbon pellet containing a tungsten wire.
Size of the tungsten wire in the carbon tube is 0.1 mm in diameter and 0.7 mm in length. The pellet isaccelerated by pressurized He gas of 10 – 20 atm and injected at 10-O port with the angle of 12ÛWRa radial
axis on the equatorial plane. Photon emission of injected tungsten was measured at 7.5-L port along 44 linesof sight on a vertically elongated poloidal cross section. Time-resolved (sampling times for 41.55 ms at every
100 ms) emission spectra are obtained using a Czerny-Turner visible-UV spectrometer (grating of 1200
gr/mm and slit width of 50 ȝm) equipped with a CCD detector.
3. Effects of ECH on local M1 line intensities

Left: Fig. 1 Radial profiles of electron temperatures by Thomson scattering. Solid squares
stand for the discharge with 154 GHz ECH, and open square for without the ECH.
Right: Fig. 2 Local line intensities of the W27+ M1 line at 337.7 nm for each discharge with
(151836) and without (151832) the ECH.
Figure 1 shows evolutions in radial profiles of electron temperatures after tungsten pellet injection at
4.1 s for each discharge without and with superimposing 154 GHz ECH (about 2MW from 4.2 s until 4.7 s).
The NBI heating of about 10 MW was being applied during the both discharges. Without the ECH the central
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electron temperature decreases down to about 1 keV due to a strong radiation of tungsten, whereas
superimposing ECH recovers that up to about 3 keV. A near-UV line emission due to an M1 transition of
W27+ in the ground state was observed at 337.7 nm [12] in the both discharges. Line integrated intensities ofthe M1 line along each line of sight were measured, and local intensities of the M1 line on the poloidal crosssection were obtained by Able inversion. Figure 2 shows the local intensities as a function of normalizedminor radius, ɏ. At 4.2 s (before superimposing ECH), the both discharges show the similar local intensity
profiles for the M1 line peaking at ɏ = 0.4 െ 0.5 where the electron temperature is about 1 keV (see Fig.
1). Without the ECH, the peak position moves toward the center (ɏ = 0) following the decrease of the central
electron temperature. In contrast, the M1 line emission at 4.5 s with the ECH which gives a high central
electron temperature of about 3 keV, has a vanishing intensity at the center, and has a peak intensity at ɏ =0.5 െ 0.6.
Photon emissivity coefficients of the M1 line become almost constant at electron densities above 1018
m-3. Average electron densities of the both discharges were 3 – 4 × 1019 m-3. Therefore, the evolutions in thelocal line intensities reflect those in the ion abundance of W27+ on the poloidal cross section.
4. A near-UV line emission of a higher charge state formed by superimposing ECH

Fig. 3 Extreme-ultra-violet (EUV) and near-UV spectra of the discharge 151836. The spectra
before tungsten pellet injection at 4.1 s (black) are also shown as a reference. The ECH wasbeing superimposed for 4.2 – 4.7 s.
Figure 3 shows extreme-ultra-violet (EUV) and near-UV spectra of the discharge with the ECH
superimposed for 4.2 – 4.7 s. It is noted that each peak in the EUV spectra is formed by many emission lines.
The EUV spectrum at 4.5 s (during the ECH) consists of peaks of W28+ or higher charge states, and the EUV
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spectrum at 5.0 s (after ECH is turned off) consists of W24+ - W28+. The charge state assignment of each peakin the EUV spectra has been done by experiments with a compact electron beam ion trap (CoBIT) [13]. The
near-UV spectrum at 4.5 s has a weak emission line at about 323 nm which is absent in the spectrum at 5.0
s. This emission line must be assigned to W28+ or higher charge states according to the corresponding EUV
spectra. Further investigations are necessary for a definite identification of the emission line.
Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by NIFS budget code
NIFS18KLPF063 and was also partially supported by KAKENHI 18H01201.

References[1] N. Noda, V. Philipps and R. Neu, J. Nucl. Mater. 214-243 (1997) 227.
[2] R. Neu et al., Fusion Eng. Des. 65 (2003) 367.
[3] T. Pütterich et al., Nucl. Fusion 50 (2010) 025012.
[4] K. Asmussen et al., Nucl. Fusion 38 (1998) 967.
[5] T. Pütterich et al., Plasma Phys. Control. Fusion 50 (2008) 085016.
[6] T. Nakano et al., Nucl. Fusion 49 (2009) 115024.
[7] C.S. Harte et al., J. Phys. B: At. Mol. Opt. Phys. 43 (2010) 205004.
[8] S. Morita et al., AIP Conf. Proc. 1545 (2013) 143.
[9] G.A. Doschek and U. Feldman, J. Apll. Phys. 47 (1976) 3083.
[10] D. Kato et al., Phys. Scr. T156 (2013) 014081.
[11] K. Fujii, D. Kato et al., J. Phys. B: At. Mol. Opt. Phys. 50 (2017) 055004.
[12] Z. Fei et al., Phys. Rev. A 86 (2012) 062501.
[13] H.A. Sakaue, D. Kato et al., Phys. Rev. A 92 (2015) 012504.

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Wall recycling experiment in EAST
N. Ashikawa1,2, K. Nojiri3, A. Terakado3, Y.W. Yu4, Z.H. Hu4, L. Wang4, Y. Suzuki1,2, J.
Huang1, F. Ding4, H.M. Zhang4, Y.J. Chen4, L. Zhang4, L. Meng4, M. Sakamoto3, J.S.
Hu4, G.N. Luo4, S. Ohdachi1
1National Institute for Fusion Science, Toki 509-5292, Gifu, Japan
2

Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan
3

University of Tsukuba, Tsukuba 305-8577, Ibaraki, Japan

4

Institute of plasma physics, Chinese academy of science, Hefei 230031, Anfui, China

Abstract
Hydrogen isotope recycling on metal plasma facing wall is one of the most critical issues in ITER and
DEMO. EAST has tungsten upper single null configurations and lithium coatings as wall conditioning
methods are used. In this work, divertor plasma parameters at different magnetic configurations arecompared for understandings of wall recycling on n tungsten with coated lithium wall. From divertor flux
measured by Langmuir probes, differences of spatial distributions at three magnetic configurations andwetting areas are shown. In a previous work, wall recycling related to divertor pumping efficiency at
different configurations was discussed. But, it can be seen that the difference in plasma characteristics indifferent configurations is not only due to pumping efficiencies but also to spatial distributions and wettingareas. In future work, a comparison between experimental and numerical results related to magnetic
configurations is planned.
1. Introduction
Understanding of hydrogen recycling on metal plasma facing wall is one of critical issues for achieving
the steady state operation in fusion devices, toward ITER and DEMO. Hence a comparison of wall
recycling on tungsten divertor targets after lithium wall conditions is planned.
Remained lithium coating was about 1.5 microns after one-day plasma discharges at the first wall
position in EAST [1]. But divertor targets have higher erosion rate by attached plasma and facing tungsten
might be occurred after few plasma discharges, due to different density decay at morning and night on theday. In this research, wall recycling at different coated lithium on tungsten divertor is compared at different
positions of divertor attachments. In addition, diverter parameters at different positions are normalizedusing numerical calculations for divertor configurations, and compared with tungsten divertor targets. And
then wall recycling characterizations on tungsten wall are shown in EAST.
2㸬Experimental
Figure 1 shows divertor structure of upper single null configurations at plasma experiment on May 19
in EAST. We requested three kinds of different configurations and standard configuration is a position of#84817 and 84815.
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An experimental time was selected at morning session in EAST [2-3]. At just morning, Ohmic

Figure 1. Divertor configurations at 5 seconds at

Figure 2. Time evaluations of a plasma

four plasma discharges in EAST, on May 17,

current and an electron density.

2019.
discharges were operated and then tokamak discharges were started. We had done about 5 plasma
discharges at three different configurations.
Data of spectroscopes at mid-plane and Langmuir probe measurement around upper divertor regions
were obtained. Electron density of 1.8 E19 during 6 seconds is sustained as shown in Fig.2. Electron
densities were modulated that might be done by density feedback control. For different divertor
configurations, electron densities are similar to each other as shown in Fig.2.
3. Results and Discussions
Spatial profiles of divertor flux by divertor probes at upper single null as shown in Fig.3 Distances froma corner edge of the divertor to probe positions are shown at Y –axis in Fig.3. Divertor configurations at
each discharge are already shown in Fig.1 A white dashed line shows peak position of divertor profiles at#84815 and #84817 discharges, which is the same configuration.
A comparison between #84818, 384817 and #84816 shows similar peak positions, but profiles are different.
Based on these results in Fig.3, it is shown that different wet areas and configurations of divertor plasmas
in EAST. In ref.4, wall recycling related to divertor pumping efficiency at different configurations was
discussed. In Fig.3, it can be seen that the difference in plasma characteristics in different configurations is
not only due to pumping efficiencies but also to spatial distributions and wetting areas. In Fig.3, details of
the divertor configurations have not yet been shown due to limitations on detection limits, such as positionsof diverter probes.
Figure 4 shows spatial profiles of tungsten and deuterium intensities measured by spectroscopies [5],which observed upper single null configurations. Due to problems of diagnostics, data at some plasma
discharges was not obtained. At least, we did not obtain differences at different configurations as shown in
Fig.4, and we require additional plasma experiments.
Numerical analyses of magnetic field at edge plasmas for divertor configurations are useful tools and it was
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84818

84817

84815

848

Figure.3 Spatial profiles of divertor flux at upper single null. Each divertor configurations areshown in Fig.1 Distances from a corner edge of the divertor to probe positions are shown at Y –axis. A white dashed line shows peak position of divertor profiles at #84815 and #84817
discharges.

Figure 4. Spatial profiles of deuterium and tungsten intensities,
measured by spectroscopies measurement from mid-plane port.

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started to support diagnostics issues for EAST. In future works, experimental and numerical results arecompared.
4. Results
We had plasma experiments at different magnetic configurations for understandings ofwall recycling in EAST. From divertor probe measurement, different spatial distributions atthree kinds of magnetic configurations were obtained, we considered that these spatial
profiles and wetted areas on divertor targets are also important parameters forunderstanding of wall recycling in EAST. But it is also clearly understanding that it is not
sufficient to show spatial distributions, because space channel intervals are required due toinstallation location issues in actual probe measurements at fixed plasma discharges.
Based on this experiment, we plan to do scanned divertor configuration experiments at
next time, and then more details about relationships between wall recycling related andspatial distributions on divertor targets are discussed.
Acknowledgements
This work was partly supported by the Post-CUP program, and the bilateral collaboration, JSPS-CAS,“Control of wall recycling on metallic plasma facing materials in fusion reactor”. It was also supported by
NIFS budget code NIFS11UFFF028-2.
References[1] N. Ashikawa, Y.W.Yu, J. Wu, E. Bang, S. Son, K. Katayama, et al., ITC-26/APFA-11 conference,
Oral-1, December 2017, Toki, Japan.
[2] Y.W. Yu, et al., Phys. Scr T70 (2017) 014070[3] G.Z. Zuo, J.S.Hu, et al., J. Nucl. Mater. 438 (2013) S90.
[4] L. Wang et al., Nucl. Fusion 59 (2019) 086036.
[5] F. Ding, et al., J. Nucl. Mater. 455 (2014) 710.

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Laser-based wall diagnostics of LIAS and LIBS for the plasma wall interaction study
on the EAST tokamak
Z. Hu1, J Oelmann2, C. Li3, F. Ding1, J. Wu1, L. Sun3, J. Liu3, N. Ashikawa4,
S. Brezinsek2, H. Ding3, G. Luo1
1

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui, China
2Institute of Energy and Climate Research, Forschungszentrum Jülich,

AssocEURATOM-FZJ, Jülich 52425, Germany
3

Dalian University of Technology, Dalian 116024, Liaoning, China

4

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

Abstract
The PWI at the first wall in fusion reactors is critical for the life time of plasma-facing components (PFC)and tokamak safety. There is an urgent need to measure PWI processes in situ. Two laser-based diagnostics
of laser-induced breakdown/ablation spectroscopy (LIBS and LIAS) were successfully developed for theplasma wall interaction of fuel retention, material deposition and transport study on EAST tokamak. The
LIBS was applied for the daily routinely monitoring the first wall compositions. The dedicated LIAS
experiment on graphite and tungsten during limited and divertor discharge indicate that LIAS is suitable forthe dynamic measurements of first wall condition in EAST tokamak.
1. Introduction
Plasma–wall interaction (PWI) plays a key role in future fusion reactors[1-3]. The post-mortem analysis
cannot meet the requirements of in situ characterization of fuel retention and impurity deposition in the future
fusion device with the active cooled plasma-facing components (PFCs). To ensure the safety operation of thefusion device, there is an urgent need to develop in situ diagnostics to characterize the material deposition
and fuel retention on the first wall and to further understand the PWI processes for plasma performance.
Laser-based methods are promising candidates and highly important for ITER and future fusion devices[4,5],
as they are required to monitor the tritium inventory without the break of vacuum and allow the monitoringof PWI processes in real-time, compared to campaign-integrated results based on post mortem material
analysis. Here we report on the development of LIBS and LIAS diagnostics for the plasma wall interaction
study on the EAST tokamak.
2. LIBS results in EAST campaign
During the EAST experimental campaign of 2016 and spring 2017, LIBS measurements were carriedout in vacuum conditions after the plasma discharge finished on a routine daily basis[6]. Experiments werecarried out after wall conditioning (lithium coating, boron coating, etc.) finished. LIBS measurements wereconducted on the Mo substrate with lithium coating. To highlight the depth profiling capabilities of LIBS, anumber of subsequent laser shots were repeatedly focused at a fixed position on the sample surface and the
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LIBS spectra as a function of the laser shot number were recorded for each measurement position. Since
LIBS can well detect the hydrogen isotope from the co-deposition layers, it was routinely used on daily basis
for monitoring the isotope ratio of the first wall in EAST. The isotope ratio measured under different
conditions in the 2016 EAST campaign is presented. The LIBS results showed that the isotope ratio was of
64% in the boron film because of the carborane was used for the boronization. The isotope ratio after lithium
conditioning decrease to about 20% because the lithium condition would effectively remove hydrogen by
Li/H co-deposition processes. After the water leakage in the vacuum chamber, the isotope ratio increases to
70% because of the water. The first wall was saturated after long-pulse and high-performance discharge. As
shown in Fig.1, Isotope ratio change daily on the first wall measured by LIBS, when the vacuum chamber
was opened, the H/(H+D) was increased due to the wall absorbed the water vapor. The results in different
conditions indicated that LIBS could be a useful method for predicting the wall condition of EAST tokamak.

Fig.1 Isotope ratio on the first wall measured by LIBS
3. LIAS results in EAST campaign
Dedicate LIAS experiments were carried out on graphite and tungsten during the limiter and divertor
discharges to study the dynamic retention in situ, separately. A schematic illustrating the measurement
strategy is shown in Fig.2(A) and Fig.3(A).
In the dedicate experiments of the limiter shape of the plasma, due to the location which was measured
by LIAS is lower than the midplane, so the plasma was actively moved down 5cm to increase the particle
flux which is shown in Fig.2(A). Two groups of triple probes were installed near the samples. From the signalof the probe, the saturation ions of the probe gradually increased. During the discharges, multiple laser pulses
were shotted on the first wall to measure the dynamic retention. As shown in Fig.2(B), the signal of LIAS
was plotted with the plasma position. The LIAS signal on both of graphite material and tungsten material
have a good linear relationship with the ion saturation. The retention on graphite material is high than that oftungsten material because the retention on the metallic material is low.

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Fig.2 (A) Measurement strategy for dynamic retention with LIAS during limiter discharges. Shifting theplasma position downwards to increase the LIAS signal. (B)Dynamic deuterium retention on graphite andtungsten during limiter discharges
In the dedicate experiments of the divertor shape of the plasma, the plasma was actively moved out fromthe high field side from 3.5 cm to 6.5 cm to decrease the particle flux which is shown in Fig.3(A). From thesignal of the probe, the saturation ions of the probe gradually decreased. During the discharges, multiple laser
pulses were shotted on the first wall to measure the dynamic retention. As shown in Fig.3(B), the signal of
LIAS was plotted with the distance between the plasma and first wall. The LIAS signal on graphite material
has a good linear relationship with the ion saturation. For the tungsten material, there is no clear trend, moredetailed analysis is necessary.

Fig.3 (A) Measurement strategy for dynamic retention with LIAS during divertor discharges. Shifting theplasma position out of the high field side to decrease the LIAS signal. (B)Dynamic deuterium retention ongraphite and tungsten during divertor discharges

4. Conclusion
Laser-based diagnostics of LIAS/LIBS was successfully developed and applied for the plasma wall
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interaction study in EAST tokamak. The results indicate that the LIBS is a useful method to predict thewall condition for the operation and wall conditioning. The dedicated LIAS experiments performed ongraphite and tungsten indicates LIAS could be suited for dynamic retention measurements. With theupgrade of the lower tungsten divertor, the LIAS/LIBS will upgrade to measure the divertor region in
the future.

Acknowledgements
This work was supported by supported by JSPS-CAS Bilateral Joint Research Projects (No.
GJHZ201984) and National Natural Science Foundation of China (Nos. 11605238, 11575243, 11605023).

References[1] C. Linsmeier, et al., 442 (2013) J. Nucl. Mater. S834
[2] A. Möslang, et al., 45 (2005) Nucl. Fusion 649[3] H. Zohm, et al., 53 (2013) Nucl. Fusion 073019[4] A. Huber, et al., 86 (2011) Fusion Eng. Des. 1336–1340[5] V. Philipps, et al., 53 (2013) Nucl. Fusion 093002[6] Z. Hu, et al., 19 (2017) Plasma Science and Technology 025502

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The deuterium permeation behavior in W studied by gas driven permeation andplasma driven permeation methods
1
Mingzhong Zhao , Shota Yamazaki2, Moeko Nakata2, Fei Sun3, Takuro Wada2,
Ayaka Koike2, Yasuhisa Oya2
1

Graduate School of Science and Technology, Shizuoka University, Shizuoka, 422-8529 Japan
Graduate School of Integrated Science and Technology, Shizuoka University, Shizuoka, 4228529 Japan
3
Faculty of Science, Shizuoka University, Shizuoka, 422-8529 Japan

2

Abstract
Well understanding of tritium (T) permeation and retention behavior in tungsten (W) requires
the knowledge of hydrogen isotopes transport parameters. In the present study, the deuterium (D)permeation behavior in tungsten (W) are studied by gas driven permeation (GDP) and plasma driven
permeation (PDP) methods. The D-GDP process in W agrees with the Sieverts’s law that the steadystate permeation flux under different temperature are proportional to the square root of pressure.
According to the Siverts’ law and Fick’s first law, the D permeability and effective diffusivity in W
are obtained from the GDP results. To study the hydrogen isotopes plasma driven permeation (PDP)
behavior in plasma facing materials, a linear Radio Frequency (RF) plasma device has been
constructed in the radiation controlled area at Shizuoka University. In the PDP experiment, thesteady state permeation flux is approximate proportional to the square root of the ion flux indicatingthat D-PDP process takes place in the RD regime. The D recombination coefficient on upstream
surface of W is obtained by the PDP results.
1. Introduction
Tungsten (W), is the most attractive plasma facing material (PFMs) for fusion device attributeto its favorable properties, such as high melting point, low physical sputtering yield and low T
retention [1]. As the plasma facing material, W will be exposed to deuterium and tritium plasma
under elevated temperature leading to the concern on the tritium (T) retention and permeation in W.
To realize the T self-sustaining and safety operation, the evaluation of T retention and permeation
behavior in W is important. Well understanding of T retention and permeation behavior in W
requires the knowledge of hydrogen isotopes transport parameters.
The hydrogen isotopes permeability and effective diffusivity in W material obtained by thermal
desorption method [2, 3] and gas driven permeation (GDP) method [4-10] have been reported. The
hydrogen isotopes recombination coefficient in W are given in [11, 12]. However, the available dataon hydrogen isotopes permeation parameters in W are very limited at present. In addition, there isdiscrepancy in the extrapolation of the data. Hence, it is necessary to enrich the basic hydrogen
permeation parameters in W.
In the present study, the D permeation behavior in W are studied by GDP and PDP methods. A
linear Radio Frequency (RF) plasma device for PDP study has been constructed in the radiation
controlled area of Shizuoka University.
2. Experimental
W wafers cut from W rod, which is supplied by A.L.M.T. Corp. Ltd., are used in the present
study. The sample disc size for GDP experiment is 6 mm in diameter and 0.5 mm in thickness. For
1

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PDP experiment, the sample size is 10 mm in diameter and 0.5 mm in thickness. Both side of W
wafers used in the experiment are mirror finished with the surface roughness of ~50 nm. To release
the internal stress, annealing treatment in high vacuum is performed under the temperature of 1173
K for 30 min.
The D permeation behavior in W material are studied by GDP and PDP methods where thegrain elongated orientation is parallel to the D diffusion direction. In the present work, the GDP
experiments are performed at temperature from 627 to 965 K. The PDP experiments are conductedin the temperature range from 749 K to 813 K. The permeation signals, mass 4 (D2) and mass 3(HD), are measured by a quadrupole mass spectrometer (QMS). The mass 4 signal measured by
QMS is calibrated by two standard D 2 leak bottles with different leak rate. The sensitivity of HD istaken as an average of the H 2 and D 2 sensitivities. Both of HD and D 2 signal are considered in the
GDP and PDP experiments.
3. The deuterium gas driven permeation behavior in W
The D permeation flux as a function of gas loading time is given in Fig. 1. The D 2 gas loading
was stared at time 0 s. In the present study, the permeation curves were obtained under the D 2
pressure of 83 kPa. The permeation parameters, effect diffusivity (D eff ) and permeability (ĭ), canbe hardly effected by the hydrogen isotopes gas loading pressure, unless the pressure is pretty high[13]. The effective diffusivity is calculated by time lag method [14].
L2

Deff = 6t(1), where, t is the lag time, which is obtained by the time-integrated method. L is the thickness of thesample. The temperature dependence of effective diffusivity from the present experiments can begiven by:Deff =3.0×10-6 exp ቀ-

0.64 (eV)kT

ቁ , m2 s-1(2)

The measured steady state permeation flux under various gas loading pressure are given in Fig. 2(a).
As shown in Fig. 2(b), the steady state permeation flux has a linear relationship with the square rootof gas loading pressure, meaning that the gas driven permeation process agrees with Siverts’ law.
According to the Siverts’ law and Fick’s first law, the steady state permeation flux (J  ) can be
expressed as:
J =

‫׎‬ξP

(3)

L

, where P is the upstream pressure, L is the thickness of sample, ĭ is permeability.
In the present study, the D permeability in W is given as follows,‫=׎‬3.2×10-8 exp ቀ-

0.81 (eV)kT

ቁ , mol m-1 s-1 Pa-଴.ହ

(4)

2

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098

Fig. 1. The D permeation flux as a function of gas loading time under various temperature.

Fig. 2. (a) pressure dependence of steady state permeation flux, (b) steady state permeation flux asa function of square root of pressure.
4. The deuterium plasma driven permeation behavior in W
Plasma experiments are performed with two different input powers of 500 W and 700 W. In
the case of 500 W, D 2 pressure in the upstream chamber is maintained to be 0.16 Pa. For 700 W, D 2
pressure of 0.05 Pa is used. The plasma parameters at the center of D plasma are measured by a
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099

Langmuir probe. The plasma parameters, namely electron temperature (Te), electron density (n e )and ion flux, under different discharge conditions are summarized in Table 1., where the ion flux is
calculated by applying the Bohm criterion [15].
Table 1. Plasma parameters under various discharge condition.
Te(eV)n e (m-3)

Ion flux (m-2 s1)

500 W

12.4

4.4×1014

1.1×1019

700 W

7.5

1.2×1016

2.3×1020

The permeation curves under different discharge conditions are shown in Fig. 3. As shown in Fig.
3(b), the influence of temperature on the D permeation behavior in W is studied with the discharge
power 700 W. Under the temperature of 803 K, the steady state permeation flux for the discharge
power 700 W and 500 W are 1.5×1018 D m-2 s-1 and 2.6×1017 D m-2 s-1, respectively. As shown in
Table 1., the ion flux of discharge power 700 W is about 20 times as high as that of 500 W.

Fig. 3. The D permeation curves in W under different discharge conditions, (a) 500 W, (b) 700 W.
There are three regimes for the steady state plasma driven permeation process [14]: (1) the
release of hydrogen isotopes from the upstream and downstream surfaces are limited by diffusion
process, named as DD-regime, (2) release of hydrogen isotopes from the upstream surface anddownstream surface are limited by the recombination of hydrogen isotopes and diffusion ofhydrogen isotopes, named as RD-regime, (3) the release of hydrogen isotopes from both surfaces
are limited by the recombination process, named as RR-regime. The relationship of steady state
permeation flux (J p ) and the implantation flux (J i ) in different regime can be expressed as:ோ‫ܬ‬௣ = ௅ ‫ܬ‬௜஽௃‫ܬ‬௉ = ට ೔
௅ ௄

ೆ

௄

ೆ
‫ܬ‬௉ = ௄ ା௄
‫ܬ‬௜
ೆ

ವ

(DD-regime)(5)(RD-regime)(6)(RR-regime)(7), where, R, L and D represent the ion projected range, sample thickness and the bulk diffusion
coefficient. K U and K D are the recombination coefficient on upstream surface and downstream
surface. In the present experiment, the steady state permeation flux is approximate proportional to
4

100
100

the square root of the ion flux indicating that permeation takes place in the RD regime.
For the RD regime, the D recombination coefficient (K U ) at the upstream surface can beobtained by the equation (6). The diffusivity measured by Frauenfelder [2] is used as the bulk
diffusion coefficient. Because the diffusivity in Ref. [2] is measured at elevated temperature andless likely to be effected by defects. The calculated K U as a function of inverse temperature is shownin Fig. 4 in which the literature data from I. Takagi [11] and R. A. Anderl [12] are also plotted. The
available hydrogen isotopes recombination coefficient in W material is real scarce at present. The
possible reasons for the difference in recombination coefficient are that (1) the recombination
coefficient is obtained by different experiment method, (2) the recombination coefficient is sensitive
to the sample surface condition. In Ref. [11], the recombination coefficient is obtained according to
the D concentration which is measured by nuclear reaction analysis (NRA). In Ref. [12], therecombination coefficient is determined by the IDP experiment with 3 keV D 3 + ions.

Fig. 4. The recombination coefficient of D for W sample.
5. Conclusion
The deuterium (D) permeation behavior in W are studied by gas driven permeation (GDP) andplasma driven permeation (PDP) methods. The D permeability and effective diffusivity in W are
obtained by GDP method. A linear Radio Frequency (RF) plasma device for the PDP study isconstructed in the radiation controlled area at Shizuoka University. The PDP results under different
discharge power show that the D permeation process in W agrees with the RD regime. The
recombination coefficient, which is an important parameter for the estimation of the hydrogenisotopes permeation behavior in fusion reactor, is obtained by the RD regime.

Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by NIFS
budget code 19KOBF042.

References[1] X. D. Du., et al., Phys. Rev. Lett. 114 (2015) 155003, Nucl. Fusion 56 (2016)
016002.
[2] T. Bando, S. Ohdachi, et. al., Nucl. Fusion 58 (2018) 082025.
5

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[3] X. D. Du., et al., Phys. Rev. Lett. 118 (2017) 125001.
[1] V. Philipps, J. Nucl. Mater., 415 (2011) S2-S9.
[2] R. Frauenfelder, J. Vac. Sci. Technol., 6 (1969) 388-397.
[3] G. Esteban, A. Perujo, L. Sedano, et. al., J. Nucl. Mater., 295 (2001) 49-56.
[4] R. Frauenfelder, J. Chem. Phys., 48 (1968) 3955-3965.
[5] A. Zakharov, V. Sharapov, E. Evko, Materials Science, 9 (1975) 149-153.
[6] F. Liu, H. Zhou, X.-C. Li, et al., J. Nucl. Mater., 455 (2014) 248-252.
[7] F. Liu, H. Zhou, Y. Xu, et al., Fusion Eng. Des., 113 (2016) 216-220.
[8] Y. Uemura, S. Sakurada, H. Fujita, et al., J. Nucl. Mater., 490 (2017) 242-246.
[9] G. Benamati, E. Serra, C. Wu, J. Nucl. Mater., 283 (2000) 1033-1037.
[10] D.A. Buchenauer, R.A. Karnesky, Z.Z. Fang, et al., Fusion Eng. Des., 109 (2016)
104-108.
[11] I. Takagi, R. Imade, Y. Ikegami, et al., J. Nucl. Mater., 417 (2011) 564-567.
[12] R. Anderl, D. Holland, G. Longhurst, et al., Fusion Technol., 21 (1992) 745-752.
[13] T. Oda, Fusion Eng. Des., 112 (2016) 102-116.
[14] H. Zhou, Y. Hirooka, N. Ashikawa, et al., J. Nucl. Mater., 455 (2014) 470-474.
[15] H. Amemiya, J. Phys. Soc. Jpn, 66 (1997) 1335-1338.

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Current status of NIFS-SWJTU joint project for CFQS
Akihiro Shimizu1, Haifeng Liu2, Shigeyoshi Kinoshita1, Mitsutaka Isobe1,3, Shoichi Okamura1,
Kunihiro Ogawa1,3, Motoki Nakata1,3, Yasuo Yoshimura1, Takanori Murase1, Sho Nakagawa1,
Hiroyuki Tanoue1, Yuhong Xu2, Guozhen Xiong2, Hai Liu2, Junren Shao2, Yangbo Li2,
Xianqu Wang2, Jie Huang2, Xin Zhang2, Changjian Tang2,4,
Dapeng Yin5, Yi Wan5 and CFQS team1,2,5
1National Institute for Fusion Science, National Institutes of Natural Sciences, Toki 509-5292, Japan

2Instituteof Fusion Science, School of Physical Science and Technology, Southwest Jiaotong

University, Chengdu 610031, People’s Republic of China
3SOKENDAI (The Graduate University for Advanced Studies), Toki 509-5292, Japan
4School of Physical Science and Technology Sichuan University, Chengdu 610041, People’s Republicof China
5Hefei Keye Electro Physical Equipment Manufacturing Co. Ltd, Hefei 230000, People’s Republic of

China
Abstract
The Chinese First Quasi-axisymmetric Stellarator (CFQS) is a new quasi-axisymmetric experimental
device planned for construction at Southwest Jiaotong University (SWJTU), China. This is a joint
project of the National Institute for Fusion Science (NIFS) and the SWJTU. The present paper
discusses the equilibrium configuration of the CFQS with a major radius of 1.0 m, a toroidal magnetic
field strength of 1.0 T, and an aspect ratio of 4.0. As the CFQS is a quasi-axisymmetric stellarator, atokamak-like substantial bootstrap current is expected. The magnitude of the bootstrap current wasestimated by BOOTSJ code. Next, the effects of the bootstrap current on the quasi-axisymmetricproperty and the neoclassical diffusion coefficient were estimated. Effect of the bootstrap current on
the quasi-axisymmetric property is not significant, and a good neoclassical transport property is
maintained.
1. Introduction
Helical devices are advantageous over tokamak in terms of steady-state operation capability as a futurenuclear fusion reactor because their magnetic configurations are produced by external magnetic fieldcoils, thus, no inductive plasma current is intrinsically required. Recently, plasma experiments conductedby a large-sized helical device, e. g. the large helical device (LHD) [1] and Wendelstein 7-X (W7-X) [2],presented remarkable results, by which the possibilities of a future helical reactor can be anticipated. The
neoclassical transport of a helical device was considered to be not good basically in the collisionless


103
103

regime, so called in the 1/Q (Q is the collision frequency) regime. In recent decades, various optimized
stellarator configurations, which have improved the neoclassical transport property, have been proposedand experimental devices have been constructed. A quasi-axisymmetric stellarator (QAS) [3] is one ofthose stellarators, of which magnetic configuration is axisymmetric in magnetic coordinates, i.e., theBoozer coordinates [4,5]. The neoclassical transport property of QAS is similar to that of tokamak;
however, the current drive is not required and the advantage of the steady-state operation capability isretained. The Chinese first quasi-axisymmetric stellarator (CFQS) is the first experimental device of the
QAS, which will be constructed in China, under the international joint project of National Institute for
Fusion Science (NIFS) in Japan and Southwest Jiaotong University (SWJTU) in China [6,7,8]. The
physical design for the CFQS has been almost completed, and the engineering is underway. The proposedparameters of the device are as follows: toroidal magnetic field B t = 1.0 T, major radius R = 1.0 m, andaspect ratio A p = 4.0. This paper presents the equilibrium configuration of the CFQS, then examines theneoclassical bootstrap current and its effect on the quasi-axisymmetric configuration. After that, we
describe about construction of mockup coil briefly.
2. Equilibrium and modular coil system of the CFQS

The present equilibrium configuration of CFQS is based on the CHS-qa, 2b32 configuration [9] having
A p of 3.2 and the toroidal periodic number N p of 2, which was previously considered for a post-Compact
Helical System (CHS) project. Based on this configuration, the geometry of the outermost magnetic
surface of the CFQS was determined, therefore, the N p of the CFQS was chosen as 2. Fabricating themodular coil system of a compact machine (e.g., R = 1.0 m) with a low aspect ratio is challenging issue,
therefore, the A p of the CFQS was enlarged to 4.0. The free-boundary VMEC [10] equilibrium results


Z (m)



M=0

M = 45

M = 90R (m)

Fig.1 Poloidal cross sections of the CFQS equilibrium calculated by the VMEC code in free-boundary
condition at three toroidal angles.


104
104

for the 0 % volume-averaged beta, <E!, at different toroidal angles (0°, 45°, and 90°) are shown in Fig.
1. Fig. 2 shows the radial profiles of the rotational transform and magnetic well depth. The typicalrotational transform was around 0.4, and it has a weak shear profile as obtained in other optimized
stellarators. The magnetic well property was achieved throughout all radial region.
As in other optimized stellarators, the modular coil system was optimized to realize the target geometryof the outermost magnetic surface, which
0.5



was

Iota

0.4

determined

beforedesigningthemagnetic coil system. The modular coilsystem produces an average normalized

L



0.2


Well depth

0.1
0Well depth

0.3

normal component of the magnetic field,
B·n/|B|, on the target outermost magnetic
surface. The coil optimization process


0

0.2

0.4

0.6

U

0.8

reduces this component to zero, while
simultaneously engineering the properties of1Fig. 2 Radial profiles of the rotational transformand well depth of the CFQS at E = 0.0%.

the filament coils, such as the minimum
distance between the coils and the minimum
curvature radius of the coils are also
evaluated and optimized. The modular coilsystem was optimized by the NESCOIL code [11,12]. In total, the CFQS consists of 16 modular coils,
which is shown in Fig.3. The minimum coil-to-coil distance and radius of curvature of the coils are 18.5
cm and 21.5 cm, respectively. Note that these values are estimated for filament structure coils [6]. This
coil system is used to obtain the free-boundary VMEC equilibrium results as shown in Fig.1 and Fig.2.






Fig. 3 Modular coil system of CFQS.



105
105

3. Bootstrap current and its effect on the configuration
The neoclassical properties of the QAS resemble those of tokamaks. Therefore, similarly to the CHSqa [13], a large bootstrap current is expected in the QAS. The magnitude of the neoclassical bootstrap
current was estimated by the BOOTSJ code, which uses a semi-analytic formula in the collision-lesslimit [14]. The temperature and density profiles were assumed as T = T 0 (1 - U2) and n e = n e0 (1 - 0.8U2+ 1.3U4 - 1.5U6), respectively, and T i0 was set to 0.75 T e0 . The radial profile of the bootstrap current
density at E = 1.19% is shown in Fig. 4(a). The equilibrium was obtained by VMEC free-boundary
calculation, assuming the modular coil system described in Section 2. The dependency of the total
bootstrap current on <E> is shown in Fig.4(b). The total bootstrap current in this case with <E> =1.19% was evaluated as 26.1 kA.
The effect of finite beta including bootstrap current on the neoclassical transport was investigated by
the NEO code [15], which can estimate the radial profile of the effective helical ripple, H eff . The
neoclassical diffusion coefficient in 1/Q regime is expressed as, D neo = v d 2 H eff 3/2/Q. Here, Q and v d are
the collision frequency and the drift velocity, respectively. The radial profiles of estimated H eff 3/2 isshown in Fig. 5. This figure shows the radial profiles for the free boundary VMEC equilibrium invarious <E>. The neoclassical bootstrap current effect on the equilibrium is included in this result. The

H eff 3/2 profile of the fixed boundary VMEC equilibrium for CHS (R ax = 92.1cm, E = 0.0%) [16] iscompared with that of the CFQS. The H eff 3/2 at U ~ 1 of the CFQS is about three orders less than thatof CHS. Moreover, a good confinement property of the neoclassical transport is maintained although

E is increased.

 D  E 
0.05

40

E = 1.19 %ne0 = 1.0 x 1019 m

35

0.04

Ti=0.75 Te

I (kA)

dI/ds (a.u.)

30

0.03
0.02

-3

25
20
15
10

0.01

5

0



0

0.2

0.4

U

0.6

0.8

0

1



0

0.5

1

E (%)

1.5

2



Fig. 4 (a) Radial profile of bootstrap current density of the CFQS at the <E! of 1.19 %. (b) The
bootstrap current as a function of the volume averaged E in the CFQS.



106
106

CHS VMEC
CFQS E=0.00%
CFQS E=0.35%
CFQS E=0.71%
CFQS E=1.19%
CFQS E=1.50%

100

10-2

eff

H 3/2

10-1

-3

10

-4

10

-5

10

0

0.2

0.4

U

0.6

0.8

1



Fig. 5 Radial profiles of the effective helical ripple of the CFQS for various E with the bootstrap
current.
4. Mockup coilEngineering design of CFQS modular coils has been developed. In order to check feasibility andaccuracy of manufacturing modular coil, mockup coil is now being constructed. One of modular coils,
which is the most complicated shape, is chosen to construct mockup coil. The winding moulds consistof inner mould and bottom mould. The inner mould contains 12 types blocks. The 8.5 mm u 8.5 mm
hollow copper conductor wrapped with fiber glass and Kapton tape was already wound on this mould.
The picture of the mockup coil is shown in Fig. 6. The mould for vacuum pressure impregnation (VPI)
is now being manufactured.


Fig. 6 Picture of mockup coil just after winding process of copper conductor, which was wrappedwith fiber glass and Kapton tape, was completed.


107
107

5. Summary
The present CFQS equilibrium with R = 1.0 m, B t =1.0 T, and A p = 4.0 is shown. The neoclassicalbootstrap current is estimated by the BOOTJ code in the equilibrium calculated by the VMEC code infree boundary condition. The calculation result indicates that the neoclassical bootstrap current reaches
35 kA at <E> = 1.5 %. The effect of the bootstrap current on the quaxi-axisymmetric property and theneoclassical transport is investigated. This effect is not very large and good neoclassical transport
properties are maintained up to <E> of 1.5%, which was confirmed by the NEO code. The detailed
studies of MHD stability are future issues. Mockup coil is being constructed now to check feasibility
and accuracy in manufacturing.
Acknowledgements
The authors are thankful for the strong support and encouragement to the NIFS-SWJTU joint project(NSJP) from Director General Prof. Y. Takeiri of NIFS, former Vice President of SWJTU Prof. W. G.
Zhang, and Vice President of SWJTU Prof. Z. Zhou. This research is supported by the NIFS generalcollaboration project, budget number NIFS17KBAP034, NIFS18KBAP041, NIFS budget ofpromotion of magnetic confinement research using helical devices in Asia, URSX 401, and theinternational collaboration with SWJTU, UFEX105.
Reference[1] M. Osakabe, Y. Takeiri, T. Morisaki et al., Fusion Sci. Technol. 72, 199 (2017).
[2] T. Klinger, A. Alonso, S. Bozhenkov et al., Plasma Phys. Control. Fusion 59, 014018 (2017).
[3] J. Nührenberg et al., in Theory of Fusion Plasmas (Proc. Joint Varenna-Lausanne Int. Workshop
Varenna, 1994), Editrice Compositori, Bologna 3 (1994).
[4] A. H. Boozer, Phys. Fluids 23, 904 (1980).
[5] A. H. Boozer, Phys. Fluids 24, 1999 (1981).
[6] H. Liu et al., Plasma Fusion Res. 13, 3405067 (2018).
[7] A. Shimizu et al., Plasma Fusion Res. 13, 3403123 (2018).
[8] M. Isobe et al., Plasma Fusion Res. 14, 3402074 (2019).
[9] S. Okamura, K. Matsuoka, S. Nishimura, M. Isobe et al., Nucl. Fusion 41, 1865 (2001).
[10] S. P. Hirshman, J. C. Whitson, Phys. Fluids 26, 3553 (1983).
[11] M. Dravlak, Fusion Technol 33, 106 (1998).
[12] P. Merkel, Nucl. Fusion 27, 867 (1987).
[13] M. Isobe, S. Okamura, N. Nakajima et al., J. Plasma Fusion Res. SERIES 5, 360 (2002).
[14] K. C. Shaing et al., Phys. Fluids B 1, 148 (1989).
[15] V. V. Nemov, S. V. Kasilov et al., Phys. of Plasmas 6, 4622 (1999).
[16] K. Matsuoka et al., 12th Int. Conf. Plasma Phys. and Control. Nucl. Fusion Research, Nice, 1988
(IAEA, Vienna) Vol.2, p411 (1988).


108
108

Current status of engineering aspect for CFQS
Shigeyoshi Kinoshita1, Akihiro Shimizu1, The CQFS team1,2,3,4
1
23National Institute for Fusion Science (NIFS), Toki 509-5292, JapanThe Graduate University for Advanced Studies (SOKENDAI), Toki 509-5292, Japan

Institute of Fusion Science, School of Physical Science and Technology, Southwest Jiaotong University(SWJTU), Chengdu 610031, China
4

Hefei Keye Electro Physical Equipment Manufacturing Co., Ltd, Hefei 230000, China

Abstract
The CFQS will be constructed on the inter-national joint project with NIFS in Japan and SWJTU in
China. We are currently evaluating the technical feasibility and we have just started prototyping the mainparts in China. It is being designed as a low-aspect-ratio quasi-axisymmetric stellarator with a modular coilsystem in a CHS-qa configuration. We are reporting the current status of the CFQS from the viewpoint ofengineering design.
1.

Introduction
The CFQS project began in July 2017 as a joint international nuclear fusion research project with NIFS

and SWJTU[1]. The CFQS is a low-aspect-ratio quasi-axisymmetric stellarator, based on a CHS-qa
configuration, which aims to improve neoclassical transport and explore high- configurations by combining
the best features of advanced tokamaks and optimized stellarators [2]. It is expected to be stable in magnetohydrodynamic (MHD) without a conducting structure, require no current drive at highgive good orbit

confinement, and provide neoclassical confinement equivalent to that of a tokamak. In addition, achieving a
low-aspect-ratio in a quasi-axisymmetric configuration is an attractive way to minimize the costs of nearterm experiments and the capital costs of possible future power plants [3]. In designing the magnetic
configuration, the first priority is placed on the core confinement region. However, it is well known that theconcept with peripheral magnetic configuration called divertor configuration will be very important for thefusion reactor. The CFQS project also aims to confirm the effectiveness of the island bundle diverter [4].
At first, we had studied the design without coil case. In the study, it was found that too many supports
were required, assembly work became very complicated and it was difficult to provide large and manymeasurement ports. We have proposed the improved design with the coil case and we have carried out moredetailed evaluation.
2.

Overall design
Table 1 gives physical data in the several stellarators. The CFQS’s main parameters are as follows:

toroidal periodic number m = 2, aspect ratio A p = 4, magnetic field strength B t = 1 T, and major radius R 0 =

1

109
109

1 m. The aspect ratio of the CFQS is the smallest and the modular coil shape is very complicated. They are
the disadvantages that make engineering design is relatively hard. For example, the work area in the central
region is very narrow and special structural idea is necessary for assembly and maintenance. Diameter of thecentral work area can be used for assembly is only 0.4m, that is one of the causes that make engineering
design difficult, because it is not easy for workers to enter the central area. . The electromagnetic force isvery large although the magnetic field in the plasma is only 1 tesla. Maximum magnetic field at coil surface
is about twice that of the plasma center.
Table 1 Comparison of major physical specifications

No
1
2
3
4
5
6
7

Parameters
Major Radius㻌 (m)㻌Minor radius (m)Aspect ratio
Magnetic Field㻌 (T)㻌 㻌
Toroidal Periodic
NumberRotational transform㻌
Magnetic Well Depth

CFQS
1.0
0.25
4
1
2

W7-X
5.5
0.5
11
3
5

NCSX
1.4
0.33
4.2
2.1
3

HSX
1.2
0.15
8
1
4

0.35~0.38
0~0.0250.85~1

0.4~0.65

1~1.1

Fig. 1 shows the experimental building for the CFQS which is built near Mt. Emei in China. Size of thebuilding is about 48.7 m x 35.5m. Fig. 2 shows the major equipment in the torus hall. The CFQS body, coil
power supplies, diagnostic system, heating system and so on will be installed in the torus hall. A generator
and a cooling water system also required, but may be outside. The plasma heating apparatus is intended toprovide for tangential neutral beam injection (NBI) of 40 kV and 1MW and electron cyclotron resonanceheating of 54.5GHz and 450 kW. A Thomson scattering system, a microwave interferometer, a charge
exchange system, a heavy ion beam probe, and magnetic measurements, among others will be used for plasma
diagnostics,.

Fig. 1 CFQS experimental building

2

110
110

Fig. 2 Layout in the torus hall

We have designed the CFQS body according to the design targets as shown in Table 2. Fig. 3 shows aschematic of the CFQS, which is approximately 4900 mm high and has an outer diameter of 4300 mm. The
electromagnetic force applied to the modular coils is supported by a cage-type support structure that was also
applied for the CHS in NIFS. The top and bottom frames, and the outer pillars will be used for fixing thepoloidal field coils, crossover conductors, power cables, cooling pipes and so on. Of course, they are useful
to set up the measuring device.
Table 2 Design targets for the CFQSNo Parts
Assembly
1

2

Coil

3

Vacuum vessel

Design targets
Completing the vessel before installing the modular coils because it isdifficult to repair defects of the vessel with coils
Connecting four sections of the vacuum vessel by welding from theinside because there is no working space outside the vacuum vessel
with coils
Three types of coil (MC/TF/PF) and nine power supplies will be
prepared for various experiment with different configuration and torealize the island bundle
Making the cross section as large as possible to increase flexibility ofthe experiment and to install measurement equipment in the vessel.
Preparing large ports with an inner diameter of 300 mm or more We
want to have at least three ports for ECH, tangential NBI and Thomson
scattering.
Having no one turn break to simplify the structure. The ECH will be
used as a main plasma source instead of joule heating.
Having no baking heater on the vessel to simplify the structure.
Inductive high-frequency heating will be applied for ႏbaking

The schematic of the coil system is illustrated in Fig. 4. A basic quasi-axisymmetric configuration can

3

111
111

be produced by four types of modular coils (for a total of 16). However, in order to allow for flexible
configuration, we have also installed four poloidal field coils (to move the magnetic axis horizontally) andtwelve toroidal field coils (to change the rotational transform). The coil currents are driven by four power
supplies for the modular coils, two for the poloidal field coils, and three for the toroidal field coils.

Fig. 3 Schematic of the CFQS

Fig. 4 Schematic of the coil system

3.

Modular coil design

4

112
112

Shapes of the modular coils and their cross section are shown in Fig. 5. Since the current density is very
high, the conductor temperature rises by about 20° per second of operation. In order to absorb this heat before
the next shot, the coils are cooled by pure water. Very large magnetic forces are applied to the modular coils.
The forces’ magnitude and direction may be very complicated, and their characteristics are not easy todescribe. They are similar to those of a tokamak’s toroidal field coil, but the modular coils are more difficult
to analyze and operated under more severe conditions because even the components that are negligible fortokamaks cannot be ignored. The strength of the coil conductor and insulation alone is not expected to be
sufficient for large forces, it must be reinforced with a SUS case like the tokamak’s toroidal field coil. In
order to reduce costs, the c-shape cases are selected. The c-shape case is planned to consist of two parts, L
shape plate and lid as shown in Fig. 6 and the cross section view of Fig. 5.
A number of the modular coils is 16 in 4 types as shown in Fig. 4. One modular coil is composed of 72
copper conductors with current of 4.43kA. Its current density 74 A/mm2 is very large. The size of therectangular cross section is 132 x 69 mm2.

Fig. 5 Modular coil cross section and appearance

The L shape plate may interfere with the case when it is put in as shown in Fig. 6. A similar problem mayoccur in the process of the coil removing from winding frame. In order to solve this assembly concern, themodular coil shape is devised. Its design method is shown below.

5

113
113

Lid㻌

Coil with earth
insulation

There is a barrier on the outside.㻌
L shape case㻌The winding form has a barrier on theinside so the interference problem is

Cutting a hole through which the lead conductors
passFig. 6 Coil case and coil with insulation.

The center trajectory of the coil is calculated with the NESCOIL system. The cross-section can be
determined from the engineering viewpoint. The size of coil is defined in consideration of the limit of thecurrent density. The rotation angle for the current carrying surface (CCS) in Fig. 7 can be selected so that
production or assemble is easy. We had assumed first a design that a longer axis was perpendicular to the
CCS, and understood a problem in the manufacturing. We found one method to solve the problem, but it was
not optimal. Fig. 8 shows the relation between the L shape winding form and the coil in the middle of theremoving process. In the conventional design, two elements interfere at two places. In the improved design,
the problem is certainly improved but a judgement whether or not we really removed it is not clear. We arelooking for the optimal solution by some optimization technique with the angle ߠ adjustment.

Fig. 7 The rotation angle for the current carrying surface

6

114
114

Fig. 8 Influence of the rotation angle

4.

Vacuum vessel design
Fig. 9 shows a schematic of the vacuum vessel, which will be manufactured by welding together four

sections in the toroidal direction. The flanges used for welding are shown. Since the electromagnetic forceson the vacuum vessel are expected to be small, it will be fabricated from SUS316L with relatively thin walls(6 mm). As a result, the one-turn resistance is roughly 0.3 m

which we believe to be sufficiently large

while not producing significant Joule heating. The vacuum vessel is fixed to the cage-type support structure
shown in Fig. 3 by eight leaf-type legs, added to allow for the vacuum vessel’s thermal expansion during
baking. Winding frames have also been included on the vessel in order to wind the TFCs. More than 40 large
ports, with a diameter of at least 114 mm, will be provided for heating and measurement.
Fig. 10 shows a cross section of the vacuum vessel. The inner surface cross section (green) is much larger
than that of the 2b32 reference surface defined for the island bundle diverter experiment , with the minimum
gap between the 2b32 plane and the vacuum vessel inner surface being 30 mm. The width and height of theinner surface are both at least 450 mm, which is sufficient for people to work inside. Using such a large
vacuum vessel also helps to make the experiment more flexible. Two large rectangular ports are also included,
both for workers to enter and for use in NBI and Thomson scattering. Fig. 11 shows the real scale model
made by 3D printer. We confirmed that a worker can put inside easily from the large rectangular port.

7

115
115

Four sections will be joined together bywelding. The flange to prevent welding
12 winding cores for

deformation are shown.

the TFC㻌
Tangential port for
Horizontal port
for Thomson
scattering (440 x

Horizontal port680)㻌IRUȝ-wave 㻌interferometer(ICF114)Large horizontal port for a

8 leaf spring type

Horizontal port for ECH

worker to go in and out (440Fig. 9 Schematic of the vacuum vesse

Fig. 10 Cross section of inner surface of the vacuum vessel

8

116
116

Fig. 11 Real scale vavuum vessel model made by 3D printer5.

Summary
NIFS and SWJTU are collaborating to construct the CFQS producing quasi-axisymmetric stellarator

configuration in China. The engineering design processes have been completed for the modular coils (andtheir support structures) and the vacuum vessel. Now we are discussing the project’s feasibility from anengineering viewpoint.
A design concept and a layout of torus hall was also proposed for the CFQS. In addition, One of the valid
overall structure was proposed. The vacuum vessel design considering the assembly procedure andexperiments with varying configuration of plasma was completed and its effectiveness was confirmed
through the real scale model. The preliminary engineering design for the modular coils and its support
structure was completed. It will be modified for manufacturing improvement, cost reduction and thereliability.
Acknowledgements

This work was partly supported by the Post-CUP program for Japan-China collaboration. It was also
supported by NIFS budget code KEJC001.
References[1] Haifeng Liu, A. Shimizu, M. Isobe et al., Plasma Fusion. Res. 13 (2018) 3405067.
[2] S. Okamura, K. Matsuoka et al., Nuclear Fusion 41 (2001) 1865
[3] Yuhong Xu, H. Liu et al., 27th IAEA Fusion Energy Conference (FEC 2018), Ahmedabad, India, (2018) EX/P523 (Poster)[4] S. Okamura, H. Liu, et al., 45th EPS Conference on Plasma Physics, (2018), Prague, Czech Republic,Vol. 42A, P5.1034 (Poster)

9

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117

Design and analysis of CFQS supporting structure
Guozhen Xiong1), Yuhong Xu1), Akihiro Shimizu2), Shigeyoshi Kinoshita2), Haifeng Liu1),
Mitsutaka Isobe2,3), Shoichi Okamura2), Takanori Murse2), Sho Nakagawa2), Dapeng Yin4), Yi Wan4),
Changjian Tang1,5), and CFQS team1,2,3)
1)

Institute of Fusion Science, School of Physical Science and Technology, Southwest Jiaotong University,
Chengdu 610031, China

2)

National Institute for Fusion Science, National Institutes of Natural Sciences, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advanced Studies), Toki 509-5292, Japan3)

Hefei Keye Electro Physical Equipment Manufacturing Co., Ltd, Hefei 230000, China

4)

College of Physical Science and Technology, Sichuan University, Chengdu 610041, China

Abstract
The Chinese First Quasi-axisymmetric Stellarator (CFQS) is now under design and construction.
It will be the first quasi-axisymmetric configuration device in the world. The main parameters of
CFQS are as follow: the toroidal period number N p = 2, major radius R 0 =1.0 m, aspect ratio A p =4.0
and magnetic field strength B t =1.0 T. The low aspect ratio makes it quite difficult to design asupporting structure because of the limited space availability and strong electromagnetic (EM) load.
In this paper, a cage-like supporting structure is proposed for CFQS to support the EM force and theweight of entire device. A finite element analysis is carried out for ensuring the reliability of the
supporting structure. The results show that the stress on the most area of the device is in allowable
range.
1. Introduction
The Chinese First Quasi-axisymmetric Stellarator (CFQS) is a quasi-axisymmetric stellarator
which is designed as a joint project of Southwest Jiaotong University in China and National Institutefor Fusion Science in Japan. The CFQS is a medium size device, having major radius of 1 m, averageplasma minor radius of 0.25 m, and aspect ratio (A p ) of 4.0. The magnetic field strength (B t ) can beincreased up to 1.0 T. The CFQS operation will be divided into the following two phases. In theinitial phase, experiment will be performed at B t of 0.09 T in order to confirm quality of the magnetic
configuration. In the second phase, B t will be increased to 1 T to perform confinement study ofquasi-axisymmetric stellarator plasma.
One of main components in CFQS is its magnetic coil system, especially the 16 modular coilswith four different types as depicted in Figure 1. In order to protect the device from heavy
electromagnetic (EM) load, a strong support structure is required. Because it is the first time for us
to design a support structure for a device with modular coils and low A p , we need an adequate
understanding of several critical issues and must carry out accurate numerical analysis for achievinga reliable design.

118
118

Figure 1. Top view of CFQS modular coils.
2. CFQS coil system and electromagnetic force analyses
2.1 CFQS coil system
The CFQS has three coil components, including 16 modular coils, 4 poloidal field coils (PFCs),
and 12 auxiliary toroidal field coils (TFCs). These three components are used for generating helicalmagnetic field, changing the plasma position horizontally and/or plasma shape, and adjusting therotational transform, respectively. The detail parameters of three coil components are listed in Table
1. The modular coils play a key role in CFQS operation. Their complicate shape can produce quasiaxisymmetric (QA) configuration.
Figure 2 shows the contour plot of magnetic field strength on each coil in the case of B t =1 T
operation. In this figure, it can be seen that the magnetic field strength on modular coils and TFCs
are larger than that on PFCs.
Table 1. Different coil components in CFQS.
Number of

Coil type

AmountModular coil

16 in four groups

72 turns in series

PFC

4 in two groups

32 turns in series

Auxiliary TFC

12 in three groups

8 turns in series

conductors

119
119

Coil current
One turn

Total
312.5 kA

4.34 kA

138.8 kA
34.7 kA

Primary role
Generate QA
configuration
Move plasma
horizontally
Change rotationaltransform

Figure 2. Distribution of magnetic field strength on the CFQS coil system.
2.2 Electromagnetic force analyses on CFQS coil system
For designing a support structure, the most important thing is the distribution of electromagnetic(EM) force on the coil system because EM force is the main load applied on our device. In thispaper, EM force distributions on modular coils are evaluated by a finite element analysis software
ANSYS/MaxwellTM. In this calculation, the coil currents for 1 T operation state are considered anda magnetostatics solution type is set for simply getting the force distribution in discharge process.
The EM force distributions on different type of modular coils are shown in Figure 3. Note that here
the EM force distribution of auxiliary toroidal coils is not shown because the EM forces on them
are small so they can be easily supported by winding frame on vacuum vessel. The results indicatethat the EM force concentrates on high-field side of coils as expected and total forces on modular
coils is larger than those on PFCs.

Figure 3. EM force distributions on different modular coils.
Different force components on modular coils and PFCs are summarized in Table 2. In thisevaluation, EM forces on coils are roughly divided into several parts in the cylindrical coordinates
to make it easier for us to predict behavior of coils. It can be seen that the centripetal component on
M1, M2, and M3 coils are large, the vertical component is also large on M2 and M3 coils. Therefore,
some strong supports should be set on the concerned area. Also, corresponding supports must be
adopted for supporting toroidal component and overturning component.
Table 2. Different force components on modular coils and PFCs.

120
120

Force components on modular Coils
Coil name
M1_1
M1_2
M1_3
M1_4
M2_1
M2_2
M2_3
M2_4
M3_1
M3_2
M3_3
M3_4
M4_1
M4_2
M4_3
M4_4
Coil Name
OPFC_U
IPFC_U
OPFC_L
IPFC_LCentripetal force(N)-167050-167050-167050-167030-142390-142390-142390-142370-94524-94517-94495-94490-39344-39327-39333-39361

Verticalforce (N)Toroidalforce (N)

Total
force

-19396-13066
19393
13068-19408-13070
19401
13046-60467-3765160463
37657-60469-3768760470
37674-92567-3791092568
37944-92588-3791592559
37925-38238-4591638242
45912-38232-4589338243
45900
Force components on PFCs
Total force
14609
12006
14502
11950

168680168680168680168660159210159210159230159200137620137630137620137600
71542
71532
71518
71544

3. Design of CFQS support structure
There are several problems we need to pay attention to in our design work. First, the center area
is fairly limited but robust supports against strong EM force are required. Second, the minimal gaps
between the two neighboring coils, and coil and vacuum vessel are quite small. Because of this, wehave to employ limited size of supports and keep space for future assembly process. Third, working
space for plasma diagnostics and heating devices must be large enough. As a reference, the crosssection and three-dimensional (3-D) computer-aided design (CAD) drawing of modular coils andvacuum vessel in CFQS are shown in Figure 4.

121
121

(a) CFQS vacuum vessel cross section(b) 3-D CAD drawing of CFQS modular coils and vacuum vessel
Figure 4. CFQS modular coils and vacuum vessel.
Considering about issues mentioned above, a cage-like support structure has been designed withtwo main components, i.e., the supporting frame and coil supports. The overview of support
structure of CFQS device is shown in Figure 5.

122
122

Figure 5. Overview of support structure of CFQS.
3.1 Supporting frame
The supporting frame consists of twelve base legs, top and bottom plates, eight outer pillars, anda center support. The ‘cage’ is formed by eight outer pillars and top/bottom plates. It is used to resist
the vertical and horizontal displacement of coils and vacuum vessel. The twelve base legs will be
fixed on the ground of experiment hall to support the ‘cage’ and leave space for vacuum pumping
system and plasma diagnostics. The center support is the main part for supporting modular coils. It
can also support top and bottom plates from the inside. The detail of support structure is shown in
Figure 6.
In this structure, large EM load on coils will be reduced by different parts. The vertical EM force
components and the weight of vacuum vessel are supported by top and bottom plates which arereinforced by 14 radial beams with cover plates. Large centripetal components will be supported by
two strong columns with 210 mm in diameter and 32 mm in thickness. To partly cancel the large
resultant force applied on those two columns and strengthen them, two H-shape stainless steel beams
are set between the columns. Two additional link bars are set between M4 coils for cancelingcentripetal components because the shape of M4 coils is too complicate to be fully supported bycenter column.

123
123

Figure 6. Center area of supporting frame.
3.2 Coil supports
Due to the complicate shape of coils and limited space in this device, we have to use coil cases
and additional link bar to support the coils from deformation. A ‘U’ shape coil case is designed forremaining space of assembly process and simplified manufacturing process. The design of ‘U’ shapecoil case is shown in Figure 7. Sixteen coil cases are going to be fabricated by welding one ‘L’ shape
shell and one lid together, furthermore, some fiber reinforced plastic (FRP) spacer will be inserted
into the case in assembly process to fill the gap between coils and coil cases.

Figure 7. Design of ‘U’ shape coil case.
There are several additional parts located on coil cases as shown in Figure 8. One of them is thecoil leg. The irregular shapes of coils require us to design legs for connecting them on top andbottom plates, and some holes are cut on coil legs for setting the cables of diagnostics. Those

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124

protuberance located on the inboard side of coil case are used for connecting coil cases to the center
support. The other protuberance located on the outboard side of coil cases are used for connecting
link bars between different coils which can support the toroidal EM force components.

Figure 8. Coil supports for four coils, coil legs (white and khaki), link bars (purple), andprotuberance (blue and yellow).
4. Finite element analysis on supporting structure
The numerical simulation is required as a reference in order to confirm the quality of support
structure design and help us to make the optimization. A finite element analysis is carried out by
ANSYS/Mechanical APDLTM, through the evaluation result of deformation, von-Mise stress, andelastic strain we can have a clear understanding of the support structure property. The allowablestress for typical material is shown in Table 3.
Table 3. Design guideline of typical materials.
Material
Young’s
Modulus
Design Stress

SUS316

C1020

C1020

-O

-1/2H

Resin

Composite(Coil)

FRP

(GPa)

E

197

100

100

1̚3

100

100

(MPa)

Sm

137

50

90

-

90

50

Design Strain

İ

˘0.1ˁ

Tensile Strength

(MPa)

ıu

520

230

270

10̚100

270

2000

Yield Strength

(MPa)

ıy

205

80

250

-

250

500

Elongation

(%)

40

50

25

1̚2

1̚2

5

125
125

Poisson’s ratio(kg/m3)

Density
Thermalexpansion
Comments

Ȟ

0.3

0.34

0.34

-

0.34

0.3

ȡ

8000

8960

8960

1100

8960

1500

Į

1.73x10-5

1.65x10-5

1.65x10-5

5x10-51.65x10-52.0x10-5

300k

OFC

OFC

after

OFC +

curingisolation

Varies•6P PLQ ı y ı u ) -- ASME III criterion•

The Sm is changed according to operating temperature.

•

If possible, it is desirable that VWUDLQUDWLRİLVOHVVWKDQRUDWZRUVWIRUWKHFRLO

Figure 9 shows the calculation model. In this case, vacuum vessel is not included in calculationand cross section of the coil is simplified for reducing required compute resource and simulation
time. In Figure. 10, the contour plot of deformation distribution on device indicates that the
maximum deformation is 0.95 mm. This is an acceptable value because the misalignment caused bythis deformation will not affect magnetic configuration significantly.
In consideration of large EM load applied on coils, we should check the von-Mise stress andelastic strain on modular coils for estimating the quality of support structure. In real design, coils
are formed by two parts, i.e., oxygen free copper and resin insulation, but in this calculation, coils
are replaced by one composite model. Therefore, both design value for von-Mise stress and elasticstrain should be satisfied. We restrict that the stress on coils should be less than 90 MPa to ensure
the stability of copper conductors, and restrict the strain on coils should be less than 0.1% to confirm
the insulation will not be broken by EM load. The distribution of von-Mise stress and elastic stresson coils is shown in Figure 11. The results indicate that both stress and strain are in allowable range.

Figure 9. Simplified support structure and coil model.

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Figure 10. Deformation distribution on supports and coils.

Figure 11. Stress and strain distribution on modular coils.
As for other supports in stainless steel, such as supporting frame and coil case, design stress isthe only limit for ensuring stability. Figure. 12 shows the stress on supporting frame and coil cases.
It is obviously that the stress on most area of the structure is less than 137 MPa, which is the design
stress for stainless steel. But there still some local area on coil supports that the stress is larger thandesign stress and close to the yield stress of stainless steel. This situation means that the structure isnot strong enough on relative place and we need a further optimization on this design.

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Figure 12. Stress distribution on supporting frame and coil cases.
5. Summary
In this paper, the distribution of EM force on modular coils are evaluated by finite element
analysis to predict coils behavior roughly. Based on the different EM force components, a cage-like
support structure design is carried out for supporting coil system and vacuum vessel. The support
structure has two main parts as supporting frame and coil supports, and they are set up to prevent
deformation and decrease large stress on modular coils caused by different EM force components.
A finite element analysis on support structure is carried out for ensuring the strength of this design.
The result indicates that in this design the coil support is strong enough to protect the modular coils,
but some structures in supporting frame should be modified and decrease large stress concentrateon them.
References[1] Mitsutaka ISOBE, Akihiro SHIMIZU, Haifeng LIU et al., Current Status of NIFS-SWJTU Joint Project for
Quasi-Axisymmetric Stellarator CFQS, Plasma and Fusion Research, Volume 14 3402074, 2019.
[2] Shigeyoshi KINOSHITA, Akihiro SHIMIZU et al. Engineering Design of the Chinese First Quasi-Axisymmetric
Stellarator (CFQS), Plasma and Fusion Research, Volume 14 3405097, 2019.
[3] Akihiro Shimizu, Haifeng Liu et al. Configuration Property of the Chinese First Quasi-Axisymmetric Stellarator,
Plasma and Fusion Research, Volume 13 3403123, 2018.
[4] Haifeng Liu, Akihiro Shimizu et al. Magnetic Configuration and Modular Coil Design for the Chinese First
Quasi-Axisymmetric Stellarator, Plasma and Fusion Research, Volume 13 3405067, 2018.

Acknowledgements
The authors would like to express their deepest gratitude to Director General Professor Yasuhiko
Takeiri of National Institute for Fusion Science, former Vice President of Southwest Jiaotong
University Professor Wengui Zhang for their strong support and great encouragement of the joint
project for the CFQS. This work is performed with the support and under the auspices of the NIFS
Collaboration Research Program (NIFS17KBAP034). Also, this work is partly supported byprograms of international collaborations with overseas laboratories (UFEX105), promotion ofmagnetic confinement research using helical devices in Asia (URSX401), and Post-CUP program

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for Japan-China collaboration.

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Engineering Analysis for Vacuum Vessel of CFQS
Sho Nakagawa1, Takanori Murase1, Shigeyoshi Kinoshita2,
and the CQFS team1,2,3,4
1

Department of Engineering and Technical Service, National Institute for Fusion Science23Large Helical Device Project, National Institute for Fusion Science

Institute of Fusion Science, School of Physical Science and Technology, Southwest Jiaotong
University (SWJTU), Chengdu 610031, China
4 Hefei Keye Electro Physical Equipment Manufacturing Co., Ltd, Hefei 230000, China

Abstract
The world’s first quasi-axisymmetric stellarator CFQS is constructed as the joint project of
National Institute for Fusion Science in Japan and Southwest Jiaotong University in China [1]. Physics
design of CFQS plasma was completed[2,3], and numerous efforts are now being made to finalize
engineering design of CFQS.
The CFQS vacuum vessel (VV) needs to be able to withstand several loads such as an
atmospheric pressure, thermal stress during baking, and gravitational force and is fairly complicatedin shape compared to tokamak fusion devices. Therefore, the structural reliability of vacuum vessel isof our great concern.
In addition, an eddy current on the VV will be induced by current changes of toroidal field coiland/or poloidal field coil. The effect of eddy current on magnetic confinement is also necessary to beinvestigated.
In this report, analysis have been performed to confirm the structural reliability and evaluate theinfluence of eddy current using finite element method software ANSYS/Mechanical™ and
ANSYS/MaxwellTM. The result shows that the stress on the VV is within an acceptable range and eddy
current have little effect on magnetic confinement.
1.

Introduction
The design of the CFQS vacuum vessel (VV) is proposed as shown in Fig.1(a). It will be madeof a thin plate of stainless steel having thickness of 6 mm and has fairly complicated shape because it
is designed so as to follow the convoluted shape of plasma[4].
The VV will be exposed to several loads such as atmospheric pressure, thermal stress during
baking, and gravitational force. A finite element method software ANSYS/Mechanical™ is employed

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to confirm structural reliability of design. Here we check a deformation and stress of the VV in orderto prevent interference with modular coils and breakage of the device.
Fig.1(b) shows poloidal field coil (PFC) and toroidal field coil (TFC) of CFQS. The estimationof a time constant is crucial for evaluating the electrical property of the CFQS VV. We conducted theanalysis of an eddy current induced by PFC / TFC current changes, and estimated a time constant ofVV.

Fig. 1

Schematic view of (a) CFQS vacuum vessel and (b) poloidal and toroidal field coils.

2.

Structural analysis of vacuum vessel

2.1.

Case 1: Structural analysis of atmospheric pressure
In a plasma operation, compressive force due to atmospheric pressure is applied on the VV while

keeping the inside in a vacuum. To clarify the effect of this force, any other load is not considered inthis case. Fig. 2 shows deformation and stress, respectively. The maximum deformation is about 1 mmon the area above the rectangular port. Main component of the deformation is in vertical direction.
The most stressed region is between port and VV in which the maximum stress reaches 87 MPa. The
stress in similar area is tend to be high as shown in Fig. 3.

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Fig. 2

Results of case 1 (a) deformation [mm], (b) stress [MPa].

Fig. 3
2.2.

High stress area in case 1.

Case 2: Structural analysis of thermal stress during baking
In case 2, only temperature rise due to baking is considered. The VV will be heated up to ႏ

in baking operation, which leads to thermal expansion of VV. The primary problem is interferencewith modular coils that are closely installed around the VV. As shown in Fig. 4(a), the maximum
deformation is about 3 mm which appears in the outside of VV. It is tolerable value because minimum
distance between modular coils and VV is about 10 mm.
Fig. 4 (b) shows stress distribution. Thermal stress is small in most region of VV, and concentrates

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only in the legs because these leaf-spring type legs have flexibility in radial direction. The maximum
stress is 128 MPa which appears in the area between the leg and VV.

Fig. 4

2.3.

Results of case 2 (a) deformation [mm], (b) stress [MPa].

Case 3 : Structural analysis considering three loads
Here we consider gravitational force in addition to above-mentioned loads. These three loads areapplied at the same time on the VV. Fig. 5 shows the deformation and stress. The maximum
deformation appears on the outside of VV, which is mainly caused by thermal expansion. The
maximum stress is 126 MPa in area between legs and VV, which is below stress limit.

Fig. 5Results of case 3 (a) deformation[mm], (b) stress[MPa].

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3.

Simulation of an eddy current induced by PFC/TFC current change
In this analysis, we calculate the time evolution of the eddy current while PFC / TFC current

changing linearly. From the time evolution of eddy current, we can estimate the time constant asfollows௧‫ܫ‬௩௩ (‫ܫ = )ݐ‬଴ (1 െ ݁ ିఛ )‫ܫ ׵‬௩௩ (߬) = ‫ܫ‬଴ ൬1 െ1
൰ = 0.6321 ή ‫ܫ‬଴ ,2.71828

where I vv is eddy current on VV, W is time constant.
If the time constant Ĳ is sufficiently shorter than typical discharge duration of 100 ms for the
CFQS, the effect of eddy current on the vacuum vessel can be neglected.
Fig. 6 and Fig. 7 show time evolution of the eddy current due to PFC / TFC current change. The
eddy current is calculated by integrating the current density over across section of VV. As can be seenin Figs. 6 and 7, the time constants are estimated to be 4.4 ms for PFC and 2.4 ms for TFC.

Fig. 6Time evolution of a PFC current and an eddy current on vacuum vessel.

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Fig. 7

Time evolution of a TFC current and an eddy current on vacuum vessel.

Summary
We evaluated structural reliability of CFQS VV and the influence of eddy current on magnetic
confinement using ANSYS/Mechanical™ and ANSYS/MaxwellTM.
In structural analysis, we chose three load conditions and clarified the effect of these loadconditions. When we considered all three loads simultaneously, the maximum deformation is 3 mmon the outside of VV, which is lower than the minimum gap between modular coils and VV. Moreover,the maximum stress is 126 MPa which is within an allowable range.
In analysis of eddy current, the time constant of the eddy current induced by PFCs and TFCs are
estimated to be 4.4 ms and 2.4 ms, respectively. These are sufficiently shorter than typical discharge
duration of several hundred of ms for the CFQS. Therefore, the influence of eddy current on magnetic
field configuration will not be significant.
References:[1] M. Isobe et al., Plasma Fusion Res. 14 (2019) 3402074.
[2] H. Liu et al., Plasma Fusion Res. 13 (2018) 3405067.
[3] A. Shimizu et al., Plasma Fusion Res. 13 (2018) 3403123.
[4] S. Kinoshita et al., Plasma Fusion Res. 14 (2019) 3405097.

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Micro-instabilities and turbulent transport in CFQS
M. Nakata1,2, M. Nunami1,2, J. Huang3, A. Shimizu1, M. Isobe1,2, S. Okamura1, H. Liu3,
Y. Xu3, and CFQS team1,3
1National Institute for Fusion Science, National Institutes for Natural Sciences, Toki 509-5292, Gifu, Japan
2

SOKENDAI(The Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan

3Institute of Fusion Science, School of Physical Science and Technology, Southwest Jiaotong University,

Chengdu 610-031, People’s Republic of China
Abstract
Based on the quasi-axisymmetric stellarator(QAS) equilibrium in CFQS, the linear micro-instability,
turbulent transport, and zonal flow generation are investigated by means of the electromagnetic gyrokinetic
Vlasov simulations. It is elucidated that QAS has more unstable ion temperature gradient(ITG) driven modes
with higher growth rate in a wider range of the wavenumber space, compared with that in the case ofaxisymmetric limit. However, the turbulent transport level is comparable or less in the nonlinear phase. Then,
relatively stronger zonal flow generation is identified in QAS for a wide range of the ion temperature gradient,which is a key driving parameter of ITG driven turbulence.

1. Introduction
The quasi-axisymmetric stellarator(QAS) concept[1] possesses combined features of tokamak andstellarator plasmas, where the vacuum magnetic field is approximately axisymmetric in the finite rotationaltransform of the field lines produced only from the external helical coils. Exploring the fundamental physicalproperties in QAS is anticipated in the Chinese First Quasi-axisymmetric Stellarator(CFQS), which is underconstruction in a joint project between Japan and People’s Republic of China [2,3].
The neoclassical transport and the linear MHD stability are mainly optimized in QAS, but theelectrostatic/electromagnetic micro-instabilities, turbulent transport, zonal flow characteristics have not fully
been clarified yet. In several earlier works, electrostatic and electromagnetic ITG modes in NCSX(QAS withthe field period N=3) have been investigated by the linear gyrokinetic analysis[4,5]. However, theunderstanding of turbulent transport and zonal flow dynamics are still limited. In this study, the linear andnonlinear simulation studies by means of electromagnetic gyrokinetic code GKV[6, 7] are carried out for the
ITG turbulence and zonal flows in the CFQS equilibrium with N=2, where the comparison with theaxisymmetric limit is of particular interest here.
The rest of the paper is organized as follows. The geometric features of the CFQS vacuum equilibrium
and the comparison with the axisymmetric limit are presented in Sec. 2. Then, the results of gyrokinetic
micro-instability analysis for the ITG modes are shown in Sec. 3. In Sec. 4, the nonlinear ITG simulations
results on the turbulent heat flux and zonal flow generation, including the ion temperature gradient parameter
scan, are discussed.

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136

2. Geometric properties
In this section, the geometric properties in CFQS is briefly described. In the following analysis, midminor radius of U=0.5 in the vacuum equilibrium is considered. The profiles of the magnetic field intensity
and the squared perpendicular wavenumber along the fluxtube coordinate z are shown in Fig. 1(a) and 1(b),where the axisymmetric limit (labeled by “Equiv. Tokamak”) with the same rotational transform profile isalso plotted for comparisons. One finds a significant difference on the perpendicular wavenumber profile,rather than the field intensity, and this can lead the different stabilization of ITG modes through the finite
gyro radius effects. Also, the dependence on the magnetic field line label is shown in Fig. 1(c) and 1(d),where each field-line label is represented by the reference toroidal angle Iof the fluxtube considered here.
The field-line-label dependence of the field intensity is relatively weaker than that of the perpendicular
wavenumber. Note that the profiles are symmetric for z=0 due to the stellarator symmetry.
3. Linear ITG instability
Linear ITG instability is examined by means of GKV, where kinetic equations for the hydrogen ions andreal-mass electrons including their finite collisions are solved numerically. More details of the numericalmodel of GKV is given in, e.g., Ref. 7. Figure 2(a) shows a comparison of ITG growth rate, which is afunction of the poloidal wavenumber, between CFQS and the axisymmetric limit, where the several caseswith different ion temperature gradients R ax /L Ti are plotted. It is found that, in CFQS cases, the ITG modes
are destabilized for wider range of the poloidal wavenumber, and higher growth rate is confirmed incomparison to the axisymmetric limit. The geometric difference of the perpendicular wavenumber (see Fig.
1) is responsible for the different ITG instability properties.

Fig. 1: Profiles of (a)the field intensity and (b)the perpendicular wavenumber in CFQS and theaxisymmetric limit, where the field-line-dependence in CFQS is plotted in (c) and (d).

Fig. 2: ITG growth rate in comparisons with (a)axisymmetric limit and (b)field line labels.
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137

In Fig. 2(b), a significant field-line dependence of the ITG growth rate is observed. One finds that thecase with I=0rindicates the highest growth rate.
4. Nonlinear ITG turbulence simulations
Nonlinear gyrokinetic turbulence simulations with GKV are presented in this section, where the adiabatic
electron approximation is used in the nonlinear simulations. Figures 3(a) and 3(b) show the time evolutionof the turbulent ion heat flux(Qi) and the zonal flow generation defined by the ratio of the zonal flow energy
and the total energy, where the axisymmetric limit is also plotted. It is observed that the nonlinear saturationlevels of the ion heat flux are comparable between CFQS and the axisymmetric limit, despite more unstable
ITG property in CFQS. Relatively stronger zonal flow generation is well correlated with the reduced transport
in CFQS. This implies a significant impact of the non-axisymmetric field structures on the zonal flow
dynamics.
As shown in Figs. 4, robustness of the reduced turbulent transport with stronger zonal flow generationin CFQS is confirmed from the nonlinear simulation scan for the ion temperature gradient parameter. For thewide range of R ax /L Ti , efficient suppression of turbulent transport by the zonal flow is observed in CFQS.

Fig. 3: (a)Turbulent ion heat flux(Qi) and (b)zonal flow generation(W ZF /W total ) calculated from thenonlinear ITG turbulence simulation with the adiabatic electron approximation.

Fig. 4: Numerical scan with the ion temperature gradient parameter for (a)linear instability, (b)turbulent
heat flux, and (c)zonal flow generation in CFQS and the axisymmetric limit.

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In summary, linear ITG instability and nonlinear turbulent transport is investigated by means ofgyrokinetic simulations with GKV, and reduced transport nature is found in CFQS. It should be pointed outthat one still needs further investigations of the detailed turbulence spectrum in order to clarify the criticalmechanism of zonal flow enhancement. In particular, the impact of the field-line-label dependence, which isinherent properties in stellarator plasmas, is still open. Also, the electromagnetic turbulence in QAS is anotherimportant issue, which will be addressed in future works.

Acknowledgements
This work was partly supported by the Post-CUP program. It was also supported by NIFS budget code
NIFS19KUHL096, NIFS19KNST148, and was also partially supported by the Ministry of Education,
Science, Sports and Culture Grant-in-Aid for Scientific Research 17K14899.

References[1] P. R. Garabedian, Physics of Plasmas 3, 2483 (1996)[2] Haifeng Liu et al., Plasma and Fusion Research 13 3405067 (2018)[3] M. Isobe et al., Plasma and Fusion Research 14 3402074 (2019)[4] J. A. Baumgaertel et al., Physics of Plasmas 18, 122301 (2011)[5] J. A. Baumgaertel et al., Physics of Plasmas 19, 122306 (2012)[6] T. -H. Watanabe et al., Nuclear Fusion 46, 24 (2006)[7] M. Nakata et al., Plasma Physics and Controlled Fusion 58, 074008 (2016)

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Publication List of NIFS-PROC. Series
NIFS-PROC-101

NIFS-PROC-102

NIFS-PROC-103

NIFS-PROC-104

NIFS-PROC-105

NIFS-PROC-106

NIFS-PROC-107

NIFS-PROC-108

NIFS-PROC-109

NIFS-PROC-110

NIFS-PROC-111

NIFS-PROC-112

NIFS-PROC-113

NIFS-PROC-114

NIFS-PROC-115

NIFS-PROC-116

Edited by Shigeru MORITA, Liqun HU and Yeong-Kook OH
Proceeding of A3 Foresight Program Seminar on Critical Physics Issues Specific to Steady State Sustainment of HighPerformance Plasmas 1-4 December, 2015, Gotemba, Japan
Sep. 09. 2016
Edited by Liqun HU, Shigeru MORITA and Yeong-Kook OH
Proceeding of A3 Foresight Program Seminar on Critical Physics Issues Specific to Steady State Sustainment ofHigh-Performance Plasmas May 17-20, 2016, Yinchuan, China
Jan. 23. 2017
Edited by Daiji Kato, Zhengying Cui and Chenzhong Dong
JSPS-NRF-NSFC A3 Foresight Program Seminar Proceedings of Japan-China-Korea Joint Seminar on Atomic andMolecular Processes in Plasma Jul. 26 - 28, 2016, Chengdu, China
Mar. 17, 2017
Atsuo Iiyoshi
Fusion for Sustainable World Development
Mar. 27, 2017
Edited by Weihua Jiang
Frontiers of Applied Pulse Power Technology
Aug. 28. 2017
Edited by Keiichi Kamada and Tetsuo Ozaki
Evolution of Pulse Power and its Peripheral Technology
Nov. 01, 2017
Edited by Hiroaki Ito and Tetsuo Ozaki
Recent Progress of Pulsed Power Technology and its Application to High Energy Density Plasma
Nov. 27, 2017
Edited by Yeong-Kook OH, Shigeru MORITA and Liqun HU
Proceeding of A3 Foresight Program Seminar on Critical Physics Issues Specific to Steady State Sustainment ofHigh-Performance Plasmas November 22-25, 2016, Jeju, Korea
Jan. 11,. 2018
Edited by Shigeru MORITA, Liqun HU and Yeong-Kook OH
Proceeding of A3 Foresight Program Seminar on Critical Physics Issues Specific to Steady State Sustainment of HighPerformance Plasmas 11-14 July, 2017, Sapporo, Japan
Jan. 12, 2018
Edited by Jun Hasegawa and Tetsuo Ozaki
Resent Developments of Pulsed Power Technology and Plasma Application ResearchJan. 12, 2018
Edited by Liqun HU, Shigeru MORITA and Yeong-Kook OH
Proceeding of A3 Foresight Program Seminar on Critical Physics Issues Specific to Steady State Sustainment of HighPerformance Plasmas 12 - 15 December, 2017, Chongqing, China
Jun. 11, 2018
Edited by E. Kikutani (KEK) and S. Kubo (NIFS)
Proceedings of the meeting on Archives in Fields of Natural Sciences in FY 2017
Oct. 18, 2018
Pulsed Power and High-Density Plasma and its Applications
Edited by Koichi Takaki and Tetsuo Ozaki
Feb. 6, 2019
The 7th Japan-China-Korea Joint Seminar on Atomic and Molecular Processes in Plasma (AMPP2018)
Edited by Daiji Kato, Ling Zhang, and Xiaobin Ding
May 10, 2019
CFQS TEAM
NIFS-SWJTU JOINT PROJECT FOR CFQS -PHYSICS AND ENGINEERING DESIGNVER. 2.1 2019. SEP.
Nov. 8, 2019
Satoshi Ohdachi, Editor of the Post-CUP Workshop proceeding
Collected papers at the 2019 Post-CUP Workshop & JSPS-CAS Bilateral Joint Research Projects Workshop,

24th-26th July, 2019, Nagoya, Japan
Feb. 21, 2020