21st International Toki Conference (ITC-21)

Integration of Fusion Science and Technology
for Steady State Operation

Book of Abstracts

November 28, 2011 – December 1, 2011
Ceratopia Toki, Toki-City, Gifu, Japan

The International Toki Conference (ITC) is an international conference for the discussion
and presentation of research activities related to nuclear fusion and plasma. It has been
held annually by the National Institute for Fusion Science in Toki city since 1989. Despite
recent progress, steady state operation of fusion energy systems remains a major issue of
research. The ITC-21 will focus on the integration of key science and technology towards
high performance fusion energy systems with the capability for steady state operation. The
conference topics cover a wide range of research in fusion science and near-term and longterm technology oriented to steady state operation.

INTERNATIONAL ADVISORY COMMITTEE
A. Komori (Japan, Chair), H. Azechi (Japan), B. Blackwell (Australia), T. Hino (Japan), R.
Horiuchi (Japan), S. Imagawa (Japan), T. Imai (Japan), S. Ishiguro (Japan), O. Kaneko(Japan), K. Kawahata (Japan), T. Klinger (Germany), L. M. Kovrizhnykh (Russia), G. S. Lee(Korea), J. G. Li (China), O. Mitarai (Japan), T. Mito (Japan), T. Muroga (Japan), T. Mutoh(Japan), H. Ninomiya (Japan), Y. Ogawa (Japan), Y. Ohsawa (Japan), S. Prager (U.S.A.),
A. Sagara (Japan), J. Sanchez (Spain), F. Sano (Japan), M. Sasao (Japan), H. Sugama(Japan), Y. Takeiri (Japan), M. Tendler (Sweden), H. Yamada (Japan), J. W. Van Dam(U.S.A.), H. Zushi (Japan)

INTERNATIONAL PROGRAM COMMITTEE
J. Jacquinot (France, Chair), T. Muroga (Japan, Local Secretary), M. Akiba (Japan), Y.
Asakura (Japan), E. Azizov (Russia), A. Becoulet (France), A. Fukuyama (Japan), P. Garin(Japan), M. Glugla (France), T. Goto (Japan), T. Hamajima (Japan), K. Hanada (Japan), H.
Hashizume (Japan), Y. Hirooka (Japan), A. Ishizawa (Japan), S. Kado (Japan), D. Kato
(Japan), S. Kubo (Japan), M. Kwon (Korea), B. Lloyd (U.K.), T. Mito (Japan), M. Mori
(Japan), K. Nagasaki (Japan), F. Najmabadi (U.S.A.), H. Nakanishi (Japan), M. Nishiura(Japan), T. Norimatsu (Japan), N. Ohno (Japan), S. Okamura (Japan), K. Okuno (Japan),
M. Ono (U.S.A.), M. Porkolab (U.S.A.), K. Sakamoto (Japan), M. Sakamoto (Japan), R.
Sakamoto (Japan), B. Sarkar (India), M. Skoric (Japan), Y. Ueda (Japan), B. Wan (China),
B. Wolf (Germany)

LOCAL ORGANIZING COMMITTEE
A. Sagara (Chair), H. Zushi (Co-chair), T. Mito (Co-chair), S. Imagawa (Scientific Secretary),
T. Akiyama, H. Chikaraishi, T. Goto, Y. Hishinuma, K.Ida, R. Ishizaki, A. Ishizawa, M. Isobe,
A. Itoh, A. Iwamoto, D. Kato, T. Kawano, M. Kobayashi, S. Kubo, S. Masuzaki, J. Miyazawa,N. Mizuguchi, T. Muraga, H. Nakamura, T. Nagasaka, M. Nishiura, T. Obana, K. Okada, R.
Sakamoto, M. Skoric, Y. Suzuki, T. Taki, H. Tamura, M. Tanaka, T. Tanaka, M. Tokitani, Y.
Tomita, K. Tsumori, S. Yamada, M. Yamanaka, N. Yanagi, K. Yasui, M. Yoshinuma

2

Contents
ProgramMondayTuesday
WednesdayThursday28th November
29th November
30th November
1st December-------------------------------------------------------------------------------------------------------------------------------------

4
6
8
9

Poster Session 1
Poster Session 2(28th November)(29th November)-------------------------------------------------------------------

10
17

28th November
29th November
30th November
1st December-------------------------------------------------------------------------------------------------------------------------------------

27
147
269
279

Abstracts of Papers
MondayTuesday
WednesdayThursday3

Program
Monday, 28th November, 2011
Time
09:10-09:50

Title

Page

OPENINGWELCOME ADDRESSGROUP PHOTO

09:50-10:30
09:50-10:30

PLENARY SESSION 1
PL-1 T. Mutoh

10:30-11:10
10:30-11:10

PLENARY SESSION 2
PL-2 S.J. Zinkle

11:10-11:30

COFFEE BREAK

11:30-12:30
11:30-12:00

SESSION 1 Confinement/2-D Devices
M.G. Bell
Progress Towards Steady-State
I-1
Operation in NSTX Through Advancesin Plasma Control and Plasma-Wall
Interactions

12:00-12:30

I-2

12:30-13:40

LUNCH

13:40-14:20
13:40-14:20

PLENARY SESSION 3
PL-3 X. Gong

14:20-15:20
14:20-14:50

SESSION 2 Confinement/3-D Devices
T.E. Evans
Suppression of Edge Localized Modes
I-3
in the DIII-D Tokamak with Small 3D
Magnetic Perturbations

14:50-15:05

O-1

W. ChenExcitation of Beta-induced Alfvén
Eigenmode during Strong Interchange
Mode in Large Helical Device

33

15:05-15:20

O-2

S. Yamamoto

Studies of MHD stability in Heliotron J
Plasmas34

15:20-15:40

COFFEE BREAK

A. Pochelon

Physics and Technology of SteadySate Operation in LHD

27

Materials Science under Extreme
Conditions: Overview of Materials
Challenges for Realizing Practical
Fusion Energy

28

29

Recent TCV results – innovativeplasma shaping to improve plasma
properties and insight

30

Overview of Experimental Results inthe EAST Tokamak

31

4

32

Time
15:40-16:40
15:40-16:10

Title
SESSION 3 Confinement/Physics Topics
S. Shiraiwa
Progress in LHCD experiment andI-4
modelling to approach AT regime on
Alcator C-Mod

16:10-16:25

O-3

A. Melnikov

Plasma Potential in Toroidal Devices: T10, TJ-II, CHS and LHD

36

16:25-16:40

O-4

G.M.D. Hogeweij

Optimizing the current ramp-up phasefor the hybrid ITER scenario

37

16:40-18:10

POSTER SESSION 1

18:30-20:00

RECEPTION FOR FOREIGN PARTICIPANTS HOSTED BY THE
MAYOR OF TOKI CITY

5

Page
35

Tuesday, 29th November, 2011
Time
09:00-09:40
09:00-09:40

PLENARY SESSION 4
PL-4 H.K. Park

Title

Page

2-D Microwave Imaging Experimentson KSTAR

147

09:40-10:40
09:40-10:10

SESSION 4 PWI/Divertor
H. MeyerSuper-X Reality: Implementing an
I-5
Innovative Divertor

10:10-10:25

O-5

I.Y. Tolstikhina

Influence of the isotope effect on the
charge-exchange process betweenhydrogen isotopes and ions and atomsof plasma facing component materials

149

10:25-10:40

O-6

H. Takeda

Divertor Simulation study using the
GAMMA 10 end-mirror cell150

10:40-11:00

COFFEE BREAK

11:00-12:30
11:00-11:30

SESSION 5 Heating/Fuelling
G.T. Hoang
Advances in Lower Hybrid Current
I-6
Drive for Tokamak Long Pulse
Operation: Technology and Physics

11:30-12:00

I-7

H. Idei

ECW / EBW Heating and Current
Drive Experiment Results andProspects to CW Operation in QUEST

152

12:00-12:15

O-7

A. Isayama

Recent Results from the Developmentof the Electron Cyclotron Heating
System for JT-60SA toward HighPower Long-Pulse Operations

153

12:15-12:30

O-8

M. WadaSteady State Operation of Ion Sources
for Fusion Plasma Heating154

12:30-13:40

LUNCH

13:40-15:10
13:40-14:10

SESSION 6 Diagnostics/Control
T. Estrada
Microwave Reflectometry Diagnostics:
I-8
Present day systems and challenges
for future devices

14:10-14:40

I-9

I. Yonekawa

ITER instrumentation and control
system toward long pulse operation

156

14:40-14:55

O-9

M. Hirsch

Diagnostic- and other in-vessel
components in the high-power m-wave
background of ECR heated steadystate discharges

157

6

148

151

155

Time
14:55-15:10

O-10 S. Kajita

Title
Fatal damages due to breakdown on adiagnostic mirror located outside thevacuum vessel in JT-60U

15:10-15:30

COFFEE BREAK

15:30-16:30
15:30-16:00

SESSION 7 Safety/Tritium
I-10 T. Tanabe
Summary of Japanese Researches on
Tritium Science and Technology for
Fusion Reactor

16:00-16:15

O-11 Y. Someya

Estimation of decay heat in fusion
DEMO reactor

160

16:15-16:30

O-12 M. Nishikawa

Overall Burning Efficiency of Tritium
and Tritium Balance in a D-T Fusion
Reactor

161

16:40-18:10

POSTER SESSION 2

18:30-20:00

OPEN LECTURE FOR CITIZENS

7

Page
158

159

Wednesday, 30th November, 2011
Time
09:00-09:40
09:00-09:40

PLENARY SESSION 5
PL-5 G. Federici

Title

Page

Challenges on the Path to DEMO andActivities in the Power Plant Physics
and Technology Department under
EFDA

269

09:40-10:40
09:40-10:10

SESSION 8 Material/Blanket
I-11 M. Enoeda
Progress of ITER Test Blanket Module
Development in Japan

270

10:10-10:25

O-13 V. Tsisar

Effect of lithium purity on in-situ
formation of Er2O3 oxide layer on V-4Ti4Cr alloy271

10:25-10:40

O-14 A.R. Gines

Silicon Carbide Nanopillar Fabricationon Silicon Carbide Substrate by Direct
CF4 Etching Using the Gas Discharge
Ion Source272

10:40-11:00

COFFEE BREAK

11:00-12:30
11:00-11:30

SESSION 9 DEMO Concept
I-12 F. Najmabadi
Assessment and comparison of pulsedand steady-state tokamak power plants

11:30-12:00

I-13

12:00-12:15

N. Yanagi

273

Design and R&D activities for the LHDtype heliotron DEMO

274

O-15 K. Yamazaki

Critical Issues of Burning Plasma,Engineering, Economic andEnvironmental Assessments in SteadyState Fusion Reactors

275

12:15-12:30

O-16 R.C. Wolf

Assessment of the physics andtechnology requirements for a fusion
DEMO

276

12:30-13:30

LUNCH

13:30-18:30

EXCURSION & NIFS TOUR

19:00-21:00

BANQUET

8

Thursday, 1st December, 2011
Time
09:00-09:40
09:00-09:40

Name
PLENARY SESSION 6
PL-6 M. Noe

Title

Page

History and prospects of applied
superconductivity technology for fusion
magnets

279

09:40-10:40
09:40-10:10

SESSION 10 Magnet
I-14 D.P. Hampshire

What is the nature of the critical current
density in superconducting materials
under strain in high magnetic fields?

280

10:10-10:25

O-17 T. Hemmi

Neutron Diffraction Study of Internal
Strain in Nb3Sn Cable-In-Conduit
Conductors

281

10:25-10:40

O-18 S. Awaji

Internal strain states of high strength
Nb3Sn wires and cables282

10:40-11:00

COFFEE BREAK

11:00-12:30
11:00-11:25

SESSION 11 Model/Simulation
I-15 V.K. Decyk
Particle Simulations on GPU
supercomputing systems

11:25-11:50

I-16

H. Sugama

Kinetic Simulations of Neoclassical andAnomalous Transport Processes in
Helical Systems

284

11:50-12:15

I-17

C.S. Chang

High Performance Integrated
Simulation of Fusion Plasmas

285

12:15-12:30

O-19 F.L. Waelbroeck

Theory of driven magnetic islands286

12:30-13:40

LUNCH

13:40-14:10
13:40-14:10

SESSION 12 Cross Cutting
I-18 H.J. Kim
Cross-cutting Ideas for a Fusion DEMO
Plant with Current and Generation IV
Nuclear Power Plants

14:10-14:40

I-19

14:40-15:10
14:40-15:10

SESSION 13 Steady State DEMO
I-20 S. MasuzakiParticle and heat control for steadystate burning plasma in helical reactor

15:10-15:35
15:10-15:35

SUMMARY
J. Jacquinot

15:35-

CLOSING

K.A. Tanaka

Research platforms of laser and ion
beam induced ablation plasmas asinterdisciplinary reactor studies

9

283

287

288

289

Monday, 28th November, 2011-Poster Session 1
No.

Author

Title

Page

P1-1

J. Miyazawa

Assessment of Plasma Performance in a Magnetic
Configuration with Reduced Poloidal Coils for a
Helical DEMO Reactor FFHR-d1

41

P1-2

Y. Takemura

Parameter dependence of MHD mode rotation in
the Large Helical Device

42

P1-4

K.Y. Watanabe

Study of Confinement Properties on Unfavorable
LHD Configurations for Ideal MHD Instabilities
using TASK3D-code

43

P1-5

E. Narita

Impact of T e/T i on energy confinement properties
in a tokamak plasma

44

P1-6

D. Sawada

Effect of Nuclear plus Interference Scattering on
Fast Alpha-particle Orbit and Confinement in DT
Plasmas45

P1-7

V. Mykhaylenko

Non-modal renormalized gyrokinetic approach fortime-dependent plasma shear flow

46

P1-8

T. Watanabe

Reduced Size LHD-type Fusion Reactor with Dshaped Magnetic Surface

47

P1-9

A.F. Nastoyashchiy Another approach to problem of controlled nuclear
fusion48

P1-10

A.F. Nastoyashchiy Laser nuclear fusion on the basis of mini-torus
shaped target

49

P1-11

M. Shoji

Design of a vacuum pumping system for a closed
helical divertor for steady state operation in LHD

50

P1-13

G. Kawamura

1D model study on the effect of impurity radiation
cooling on LHD SOL plasma

51

P1-14

T. Iijima

Effect of divertor target geometry on detachedplasma in divertor simulator TPD-SheetIV

52

P1-15

I. Katanuma

Calculation of Ion Orbits in the Divertor/Dipole
Regions of the GAMMA10 A-divertor

53

P1-16

S. TogoEffects of Radial Losses of Particle and Energy on
the Stability of detachment Front in a Divertor
Plasma54

P1-17

M. Yoshida

Hydrogen Isotopes Retention in the Tile Gaps of
JT-60U55

P1-18

A. Adachi

Tritium retention to the first wall of JT-60U

56

10

No.

Author

Title

Page

P1-19

N. Shinoda

Toroidal distributions of gas retention and impurity
deposition for 13th experimental campaign in LHD

57

P1-20

K. Matsuoka

Impurity Effects of Hydrogen Isotope Retention on
Boronized Wall in LHD

58

P1-21

M. Matsuyama

Tritium Retention on Stainless Steel Surface
Exposed to Plasmas in LHD

59

P1-22

Y. Tomita

Dust Dynamics Released from Plasma-Facing
Components in HL-2A Tokamak

60

P1-23

Y. Nobuta

Deuterium retention and desorption behavior of codeposited carbon film produced by deuterium arc
discharge with carbon electrodes

61

P1-24

A. Matsumoto

Deuterium retention in graphite and its removal byinert gas glow discharge

62

P1-25

S. Saito

Hydrogen Atom Behavior with Structural Change
of Carbon Materials

63

P1-26

A.M. Ito

Time Development of Agglomeration Process on
Carbon Dusts in Hydrocarbon Plasma

64

P1-27

T. Kenmotsu

Retention of Hydrogen Isotopes in Tungsten
Exposed to ELM Plasma

65

P1-28

Y. Hatano

Deuterium Trapping by Radiation Defects in
Tungsten

66

P1-29

K. OhyaSimulation of Deuterium Retention in Tungsten
Exposed to Divertor Plasmas

67

P1-30

A. Takayama

Simulation Study of Tungsten Exposed to Helium /Neon / Argon Based on Binary Collision
Approximation

68

P1-31

H. LiuCross-Field Motion of Plasma Blob-filaments andRelated Particle Flux in an Open Magnetic Field
Line Configuration on QUEST

69

P1-32

K. Ichimura

Analysis of end-loss-ion-flux for application studiesof the plasma flow from the end-mirror exit of
GAMMA 10.

70

P1-33

T. KonoMaterial dependence on plasma shielding inducedby laser ablation

71

P1-34

T. Oishi

Fundamental study on hydrogen retention incarbon aerosol relevant to inertial fusion reactor

72

11

No.

Author

Title

Page

P1-35

T.E. Tuballa

Optical emission characteristics of an atmospheric
Ar/N2 microwave plasma jet in relation tosuperhydrophilic surface treatment of stainless
steel

73

P1-36

M. Saigusa

Proposal of New Type Diplexer for ECCD System

74

P1-37

Y. Yoshimura

ECCD Experiment Using an Upgraded ECH
System on LHD

75

P1-38

H. Takahashi

Improvement of output power and electric
efficiency for 77 GHz gyrotrons by externallycontrolled anode voltage in LHD

76

P1-39

H. Igami

Experimental results of electron Bernstein wave
heating on the LHD and consideration for a helicaltype fusion device

77

P1-40

T. Seki

ICRF heating experiment using phasing antenna in
LHD

78

P1-41

R. SekiStudy of ICH Heating Power Profile Evaluation
Method in LHD Type Plasma

79

P1-42

T. Yokoyama

Improvement of ICRF antenna loading in theminimum-B configuration on GAMMA 10

80

P1-43

M. Sakaguchi

Quasi-Optical high purity HE11-mode Exciterfor
Oversized Corrugated Waveguide Transmission

81

P1-44

H. Nakamura

Finite-Difference Time-Domain Simulation on
Millimeter-Wave Propagation in Corrugate
Waveguide

82

P1-45

S. Kamio

Study of magnetic reconnection heating in UTST
measured by Doppler spectroscopy

83

P1-46

T. Mizuuchi

Effect of Gas Fueling Control on Plasma
Performance in Heliotron J

84

P1-47

A. Murakami

Improvement in the fueling efficiency of supersonicgas puffing in LHD

85

P1-48

N. Tanaka

Production and Measurement of Negative
Hydrogen Ions in a FET based RF Ion Source

86

P1-49

S. Ohshima

Study of Edge Fluctuation Characteristics using
Multiple Langmuir Probes in Heliotron J

87

P1-50

S. Oldenburger

Configuration of flows in a cylindrical plasma
device

88

12

No.

Author

Title

Page

P1-51

T. Kobayashi

Time Evolution of Power Spectrum Density during
Spontaneous Transition in Cylindrical Magnetized
Plasma89

P1-52

S. Yamada

Evaluation of Electron Temperature Fluctuations
Using a Conditional Technique on PANTA

90

P1-53

T.G. Watanabe

Magnetic Helicity Injection Experiments for Doublenull Startup of the UTST spherical tokamak

91

P1-54

E. XueModeling of Vacuum Field in Start-up in EAST

92

P1-55

T. Yamada

FDTD simulation analysis for improving the twodimensional electron density distribution
measurement by using phase imaging method in
GAMMA1093

P1-56

S.G. Baek

Observation of spectral broadening of LH waves in
Alcator C-Mod

94

P1-57

H. Tsuchiya

Identification of MHD Mode Structure using ECE
measurement

95

P1-58

I. HongFast, nonlocal, radially coherent changes of T eduring ELM crashes observed by 2D ECE imaging
in KSTAR

96

P1-59

Y. Nagayama

Electron Density Fluctuation Measurement with 3D Microwave Imaging Diagnostics in LHD

97

P1-60

D. SeoThe development of the Infrared TV system for
KSTAR

98

P1-61

H. Iiduka

Cylindrical Surface Wave on PeriodicallyCorrugated Metal Cylinder

99

P1-62

E. Wang

Vertical profiles of EUV spectral emissions fromedge impurity ions observed at different toroidal
angles in LHD

100

P1-63

C. Dong

Extension of wavelength range in absoluteintensity calibration of space-resolved EUV
spectrometer for LHD diagnostics

101

P1-64

H.A. Sakaue

EUV Spectroscopy of Highly Charged Tungsten
Ions with Electron Beam Ion Traps

102

P1-65

X. DinCollisional-Radiative Modeling of moderate ionizedTungsten Plasma

103

13

No.

Author

Title

Page

P1-66

A. Okamoto

Steady state recombining plasma in a radiofrequency plasma device for divertor-detachment
study

104

P1-67

M.C. Lacdan

Initial Studies of Microwave-Induced Atmospheric
Plasma Jets Using Optical Emission Spectroscopy

105

P1-68

J. Daseco

Optical emission spectroscopy studies on theeffect of negatively biased copper shield to low
pressure glow discharge plasmas

106

P1-70

T. Ozaki

Improved Charge Exchange Measurements in
Large Helical Device

107

P1-71

D. Raburn

Concept for Numerical Calculation of 3D MHD
Equilibria with Flow and FLR Effects

108

P1-72

Y. Suzuki

MHD Equilibrium Analysis of SDC Plasmas in LHD

109

P1-73

Y. Asahi

MHD equilibrium analysis with anisotropic pressure
in LHD110

P1-74

A. ItoNumerical analysis of axisymmetric toroidal
equilibria with flow in single-fluid and two-fluid
MHD models

111

P1-75

K. Saito

MHD equilibrium including a static magnetic island
for the reduced MHD equations in straightheliotron configuration

112

P1-76

R. GotoSimulation of R-T and K-H hybrid instabilities in a
2D slab

113

P1-77

H. Miura

Development of a portable AMR module
applicable to various fluid/particle simulations

114

P1-78

K. Ichiguchi

Numerical MHD Analysis of LHD Plasma in
Magnetic Axis Swing Operation

115

P1-79

H. Usui

Dynamic Domain Decomposition for 3D PIC
simulation with Adaptive Mesh Refinement

116

P1-80

S. Nishimura

Nonlinear stability of externally driven magnetic
islands

117

P1-81

R. UedaEffect of resistivity on mode structure ofinterchange instability in heliotron plasma

118

P1-82

W. Takado

Development of an MHD Code Based on the CIP
Method to Consider the Advection with Alfvén
Velocity

119

14

No.

Author

Title

Page

P1-83

N. Mizuguchi

Modeling of Formation of Helical Structures inReversed-Field Pinch

120

P1-84

M. SatoStudy of resistivity effect on MHD stability beta
limit in LHD using TASK3D-code

121

P1-85

M. Yokoyama

Development and experimental application of
integrated transport code, TASK3D, for helicalplasmas, ~equipment and extension ofneoclassical transport module~

122

P1-86

S. Matsuoka

The radial electric field formation in high T eplasmas with the electron internal transport barrier
in LHD123

P1-87

R. Kanno

Dependence of thermal diffusivity on plasma
parameters in perturbed magnetic field in toroidal
plasma124

P1-88

Y. Miyoshi

Research on burn control of core plasma with thetransport code

125

P1-89

K. Nishioka

The neoclassical transport and viscosity in helicalplasma consisting of multi-species ions including
high-Z impurities

126

P1-90

F.P. Iizima

Electron fluid perturbation and related electrontransport in a field-reversed configuration

127

P1-91

T. Watanabe

Computation of Neutral Gas Flow Generation froma CT Neutralization Fuel-Injector

128

P1-92

T. Takahashi

Toroidal spin-up caused by an anomalous loss ofthe electron angular momentum of a field-reversed
configuration

129

P1-93

Y.D. Jung

Nonthermal effects on the resonant instability ofthe dust-acoustic wave in semi-bounded
Lorentzian plasmas

130

P1-94

P.K. Chan

Numerical Analysis of Quantum-Mechanical GradB Drift II

131

P1-95

E. Okubo

Quantum Mechanical Plasma Scattering II

132

P1-96

K. Seto

Theoretical study of ultra-relativistic laser electroninteraction with radiation reaction by quantum
description

133

15

No.

Author

Title

Page

P1-97

N. Singh

Laser Guiding Through an Axially Nonuniform
Magnetoplasma Channel in the Weakly Relativistic
Limit

134

P1-98

T. Kusakabe

Charge-transfer cross sections of ground state He+
ions in collisions with He atoms and simple
molecules in the energy range below 4.0 keV

135

P1-99

A. Saitoh

Application of Collocation Meshless Method to
Eigenvalue Problem

136

P1-100S. Ikuno

Numerical Simulation of Electromagnetic Wave
Propagation using Time Domain Meshless Method

137

P1-101

T. ItohImplicit Function with a Natural Behavior over the
Entire Domain

138

P1-102

S. Nozaki

A new De-noising Method of Laser-produced
Plasma Penumbral Images by Principal
Component Analysis

139

P1-103

H. Wakabayashi

Studies on Plasma Production and Acceleration
Using Rotating Electromagnetic Fields forElectrical Propulsion Engine

140

P1-104

S. OikawaExact trajectory of a charged particle in several
non-uniform magnetic fields

141

P1-105

T. Kamei

Application of the Binary Interaction Approximation
to Plasma Oscillation142

P1-106

M. GotoAccuracy Assurance in Binary Interaction
Approximation for N -body Problems

143

P1-107

Y. Takabatake

Localization method using Microsoft Kinect forindoor structures

144

16

Tuesday, 29th November, 2011-Poster Session 2
No.

Author

Title

Page

P2-1

H. Utoh

Modification of DOHEAT for optimization of
coolant conditions in DEMO blanket

165

P2-2

A. Tsuchiya

Removal of deuterium in lithium titanate by gas
exposure

166

P2-3

D. KatoIon-beam induced luminescence of Er2O3

167

P2-4

M. Kondo

Electroplating of erbium on steel surface in ErCl3
doped LiCl-KCl

168

P2-5

Y. Hishinuma

Microstructure and Mechanical Properties on
Oxide Insulator Coating before and after Thermal
Cycling Test

169

P2-6

T. Nagasaka

Fabrication of the hydrogen recovery unit in themolten salt loop Orosh2i-1:Operational Recovery
Of Separated Hydrogen and Heat Inquiry-1

170

P2-7

T. Watanabe

Hydrogen Permeation in Stainless Steel Modified
with Surface Nitride by Electrochemical Technique

171

P2-8

A. Watanabe

The evaluation of heat transfer for the spherepacked pipe by using a molten salt under highheat flux

172

P2-9

J. Guhit

Effects of Argon and Water Plasma Treatments on
Philippine Coconut Fibers

173

P2-10

J. Salamania

Anti-bacterial property of oxygen-ion treated cotton
gauze

174

P2-11

C. Romero

Sterilization of polyethylene terephthalate (PET)
using low-pressure glow discharge H2O2 and O2
plasmas

175

P2-12

K. Pabelina

Wettability and Antimicrobial Property of Plasma
Treated Polyethylene Terephthalate (PET)

176

P2-13

T. YagaiIc analysis of Nb3Sn strand cable-in-conduit
conductor under the electromagnetic force by the
structural mechanics

177

P2-14

S. Nakazawa

Analysis of contact lengths of strands with Cu
blocks in CICC joints

178

P2-15

Y. Tsuchiya

Measurement Technique of Stress-Strain
Properties for Cryogenic Materials by Neutron
Diffraction

179

17

No.

Author

Title

Page

P2-16

K. Takahata

Long-Term Monitoring of Hydraulic Characteristicsof LHD Poloidal Coils

180

P2-17

A. Iwamoto

Discussion of heat transfer to liquid helium on
surface orientation dependence

181

P2-18

S. Imagawa

Design Study of a 15 T Test Facility for Highcurrent Superconductors

182

P2-19

S. Satoh

Directory on Recent Information about Operational
Experiences on Large Superconducting System
and Cryogenic Facility at Fusion and Particle
Accelerator Research Organizations

183

P2-21

S. Hamaguchi

Thermal hydraulic analysis on supercritical heliumin cable-in-conduit conductor of helical coils for
FFHR-d1

184

P2-22

H. Chikaraishi

Conceptual design of dc power supplies for FFHR
superconducting magnet

185

P2-23

H. Tamura

Radial build design in the Helical DEMO reactor
FFHR-d1

186

P2-24

K. Natsume

Feasibility study on applying cryogenic oscillatingheat pipes to fusion magnet

187

P2-25

K. Osamura

Potential Use of HT-SC Wires in Fusion Reactor
Application

188

P2-26

Y. YamadaHTS Current Lead Units Prepared by the TFAMOD Processed YBCO Tapes

189

P2-27

A. Kamitani

Accurate and Stable Numerical Method for
Analyzing Shielding Current Density in HighTemperature Superconducting Film Containing
Cracks190

P2-28

K. Matsuda

Microstructure of V3Ga Superconducting Wire
Using Cu/V with High Ga Contents

191

P2-29

T. Kawabata

Microstructure and Superconductive Property of
MgB2/Al Based Composite Materials

192

P2-30

T. Ohinata

Fundamental evaluation of joint resistance inmechanical butt joint of a stacked GdBCO
conductor

193

P2-31

Y. Terazaki

Measurement of joint resistance of large-current
YBCO conductors

194

18

No.

Author

Title

Page

P2-32

S. ItoDiscussion of structural design issues on aremountable high-temperature superconducting
magnet195

P2-33

K. Ogawa

Beam-ion losses due to magnetic field ripple undervarious plasma parameter ranges on the Large
Helical Device

196

P2-34

T. Iwai

Evaluation of the non-axisymmetry of the central
cell plasma by using a segmented limiter on
GAMMA 10

197

P2-35

Y. Suzuki

Magnetic Diagnostics for Magnetic Islands

198

P2-36

N. MineTime-dependence of Balmer series emission
intensity during pulse plasma flow in divertor
simulator TPD-SheetIV

199

P2-37

B.J. Peterson

Calculation of geometry matrices for IRVBs forapplication to 3D tomography of radiative
phenomena in LHD

200

P2-38

S.N. Pandya

Design considerations for an Infrared Imaging
Video Bolometer for observation of 3D radiation
structures of detached LHD plasmas

201

P2-39

S.P. Pandya

Design and Development of Infrared Imaging
Video Bolometer for ADITYA Tokamak

202

P2-40

R. SanoCalibration of IR imaging bolometer foil by laser
irradiation

203

P2-41

X.D. Du

Development of an Array System of Ultra-Soft Xray Detectors with Large Sensitive Area on the
Large Helical Device

204

P2-42

A. Sanpei

Tangential and vertical Soft-X ray imaging for three- 205
dimensional structural studies in the RFP

P2-43

S. Ogasawara

Identification of spurious modes of high power 77
GHz gyrotron for collective Thomson scattering in
LHD

206

P2-44

R. Yasuhara

Coaxial multiple laser beam combiner for the LHD
Thomson scattering system

207

P2-45

J.H Lee

The Thomson Scattering Diagnostic System on
the KSTAR tokamak

208

P2-46

S. Oh

The comparison of electron temperature and its
density calculation methods in Thomson scattering
diagnostics

209

19

No.

Author

Title

Page

P2-47

H. Tojo

Demonstration of an in-situ relative calibration
method for a Thomson scattering diagnostics on
TST-2

210

P2-48

J. Hiratsuka

Off-axis temperature anisotropy measurement by
a double-pass Thomson scattering diagnostic
system on TST-2

211

P2-49

L.G. Eliseev

The method to measure of poloidal plasma
rotation by Heavy Ion Beam Probe

212

P2-50

N. ShiDevelopment of a multichannel far-infraredinterferometer for the Experimental Advanced
Superconducting Tokamak

213

P2-51

M. Isobe

On Plasma Diagnostics Required for Heliotrontype DEMO Reactor

214

P2-52

T. Akiyama

Conceptual design of electron density
measurement system for DEMO-relevant helicalplasmas

215

P2-53

T. Tokuzawa

Developments of Pulsed Terahertz Wave
Diagnostics for Fusion Plasma

216

P2-55

M. SatoAssessment of multi pulse laser damage thresholdof metallic mirrors for laser diagnostics in fusion
devices

217

P2-56

H. OzekiConfinement Analysis of Spherical TokamakStellarator Hybrid Configurations with Simple
Shaped Coils

218

P2-57

M. Hasegawa

Experiments of tokamak discharge and simulationof magnetic field configuration in tokamak-helicalhybrid device TOKASTAR-2

219

P2-58

H. Nakanishi

Improved Data Acquisition Methods for
Uninterruptible Signal Monitoring and Ultra Fast
Plasma Diagnostics in LHD

220

P2-59

M. Emoto

Study of Performance Improvement of Real-Time
Mapping of Thomson Scattering Data to Flux
Coordinates in LHD

221

P2-60

F. XiaThe Initial Design of Process Control System
Based on EPICS for HL-2A & HL-2M

222

P2-61

Y. Asakura

Operation planning of tritium recovery system
based on investigation results of LHD exhaustsystem

223

20

No.

Author

Title

Page

P2-62

T. UdaElectromagnetic Fields Measurement and Safety
Consideration in Magnetic Confinement Fusion
Test Facilities

224

P2-63

A. Sagara

Analysis on Tritium Management in FLiBe Blanket
for FFHR2

225

P2-64

S. Shigeharu

Hydrogen Isotopes Recovery from Liquid Li with Y
Hot Trap

226

P2-65

M. Okada

Analysis of simultaneous H and D permeation
through lithium-lead

227

P2-66

N. Ezumi

Optical emission and mass spectra observations
during hydrogen combustion processes inatmospheric pressure microwave plasma

228

P2-67

M. Nunami

Nonlinear Gyrokinetic Simulation Study on ITG
Turbulence and Zonal Flow in LHD Discharge

229

P2-68

A. Ishizawa

Effects of ion-temperature-gradient driven
turbulence on magnetic islands

230

P2-69

T.H. Watanabe

A two-scale model for zonal flow and turbulence in
non-axisymmetric systems

231

P2-70

D. Kurita

Neoclassical Tearing Mode Analysis in Tokamak
Plasmas

232

P2-71

A. Bierwage

Nonlinear Simulation of Energetic Particle Modes
in High-Beta Tokamak Plasma

233

P2-72

H. Wang

Simulation Study of Energetic Particle Driven
Geodesic Acoustic Mode in LHD Plasma

234

P2-73

H. Matsuura

Effect of Nuclear Plus Interference Scattering on
Fast-Ion Slowing-Down Distribution Functions in
Thermonuclear Plasmas

235

P2-75

H. Yamaguchi

Simulation study of NBI beam ion distributions andheat depositions in the time development plasmaof LHD236

P2-76

H. Hasegawa

Study of Plasma Blob Dynamics with Particle
Simulation

237

P2-77

K. HadaAnalysis of ECRH Preionization for Plasma Startup
in JT-60SA

238

P2-78

Y. Masaoka

Nonlinear collision effect on energetic particle
confinement in LHD plasmas

239

21

No.

Author

Title

Page

P2-79

S. IdePlasma domains and development of operation
scenarios in JT-60SA

240

P2-80

J. ManoCurrent profile control for high bootstrap current
operation in ITER

241

P2-81

R. Ishizaki

MHD simulation on pellet injection in torus
plasmas

242

P2-82

K. Hosoi

Modeling of SMBI experiments based on MonteCarlo simulation in GAMMA 10

243

P2-83

Y.S. Stadnik

Self-consistent Modeling of Plasma Production
with Radio-frequency Heating

244

P2-84

O. Meneghini

Wavelet-based analysis of Lower Hybrid full-wave
fields245

P2-85

Y. Fujikawa

Reconstruction of a toroidal flow profile of a fieldreversed configuration

246

P2-86

Y. Moriya

Dependence of the quasi linear heating term
model on the ECCD in helical plasmas

247

P2-88

I. Murakami

Effect of Radiation Power Loss due to Ne Impurity
Gas Puff for Ergodic Layer

248

P2-90

G.H. Neilson

A Pilot Plant as the Next Step toward an MFE
Demo

249

P2-91

T. OishiComparative study of cost models for tokamak
DEMO fusion reactors

250

P2-92

T. Kondo

Economic evaluation of D-T, D-3He, and catalyzed
D-D fusion reactors

251

P2-93

K. BanEconomical and Life-Cycle Energy Assessment ofMagnetic Fusion Power Reactors

252

P2-94

T. GotoDesign Window Analyses for the Helical DEMO
Reactor FFHR-d1

253

P2-95

T. Tanaka

Neutronics Investigations for Helical DEMO
Reactor FFHR-d1

254

P2-96

K. Tsumori

Conceptual Design of Heating Devices for
Heliotron-type DEMO Reactor

255

P2-97

M. Tanaka

Fuel Particle Balance Study in FFHR DEMO
Reactor

256

22

No.

Author

Title

Page

P2-98

S. Harada

Studies on Investigation of Behavior of Traveling
Wave Direct Energy Converter Using a Simulator
Installed on GAMMA 10 Tandem Mirror with One
Side Plugging

257

P2-99

Y. Munakata

Studies on Application of Cusp-Type Particle
Separation and Two-Stage Deceleration to Dense
Plasma for Direct Energy Conversion

258

P2-100S. Harjo

Internal Strain Measurement in Superconducting
Composites by Neutron Diffraction259

P2-101

T. Takayama

Numerical Investigation on Accuracy and
Resolution of Contactless Methods for Measuring
j C in High-Temperature Superconducting Film:
Inductive Method and Permanent Magnet Method

260

P2-102

T. Wajima

Adsorption Behavior of Lithium from Seawaterusing Manganses Oxide Adsorbent

261

P2-103

H. Ohtani

Application of Virtual Reality System to Fusion
Science262

P2-104

T. Umetani

Localization of Mobile Client using Wireless LAN
and Natural Features of Camera-Image
Sequences for Indoor Structure263

P2-106

M. Mizuguchi

Behavior of core plasma potential and fluctuationin end-diverter simulation experiments on thetandem mirror GAMMA10

264

P2-107

H. Matsuura

Measurement of LHD divertor heat flux responseof plasma detachment with the Hydride Directional
Langmuire Probe

265

23

Plenary / Invited / Oral
Sessions(Monday 28th November)PL-1
Physics and Technology of Steady Sate Operation in LHD
T.Mutoh, LHD Experiment Group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Steady state plasma sustainment for more than one hour is one of the main missions of the LHD
project. A fusion reactor should be inevitably a steady state operation device. LHD is a helicaltype device and well-suitable to form the steady confinement configuration and to control steadystate plasma. It is contrasted with high autonomous tokamak plasma which has to hold large
internal plasma current and also to control its radial profile. The steady state plasma operation has
been conducted from the beginning of the LHD project. LHD has the highest record of 1.6GJ(500kW/54 minute) in the incidence energy. The targeted value of the LHD mission is the incident
power of 3MW and one hour plasma sustainment. However, there are many issues to be solved tothe steady state operation. Controlling the particle balance that is a result of particle feeding, wall
recycling, wall and vacuum pumping, and also control of temperature rise of divertor, wall andheating unit are key issues.
An avoidance of impurity accumulation is also an important physics issue. Accumulation of
impurities in the plasma is observed in some limited conditions of pulsed operation in LHD. We
know the conditions to avoid the accumulation now. Usually, the exhaust of plasma particles
include impurities tends to occur in ECH heating, and more interesting character is exhaustion ofhigh Z ions observed in the high power NBI heating in pulsed operation. This confinement
character has been paid attention as a good feature for the fusion reactor.
Heat balance at divertor plate and vacuum wall, and steady driving of heating tools (ICH, ECH
and NBI) are also important technical issues. Thermal loading at the diverter plate is mitigatedusing the magnetic axis swing technique changing the poloidal coil currents which is a noble andunique technique only usable in heliotron magnetic configuration. Moreover, the booster ECH
pulse (around 0.4MW) was synchronously injected to avoid a radiation collapse against occasional
impurity flake drops and it demonstrated successfully to continue the steady state operation.
The steady state operation was mainly performed using ICRF and ECR heating, and it was
terminated by radiation collapse due to the sudden increase of plasma density caused by droppingof flakes from the divertor or vacuum wall. Recently it is clarified that the flakes are mixedmaterial of stainless steel and carbon generated by helium glow discharge for wall conditioning
and the usual plasma discharges.
The physics and technology issues relate to the steady state plasma operation connect to those of
the future fusion reactor and should give us valuable information for reactor design.

27

PL-2
Materials Science under Extreme Conditions: Overview of Materials Challenges for
Realizing Practical Fusion Energy
S.J. Zinkle
Oak Ridge National Laboratory, Oak Ridge, TN 37831-6249 USA
Recent advances in plasma physics are bringing the prospect of fusion energy close toachievement of steady state operation at fusion reactor-relevant conditions. Durable, highperformance structural and plasma-facing materials (along with numerous special-purpose
materials systems, ranging from tritium breeding and plasma heating and diagnostic systems tosuperconducting magnets) will be key for the successful development of fusion energy. The
anticipated temperatures, heat fluxes and radiation damage levels for plasma-facing and structural
materials represent an extraordinary jump compared to the conditions experienced by fuelcladding and core internal structures in existing fission power plants. Similar daunting operational
challenges face the tritium breeding and plasma heating systems, due to issues such as high heat
flux gradients, intense radiation fields, high temperature coolant compatibility requirements, andthe necessity for low absorption of transmitted plasma heating power. A science-based approach
will be crucial for the successful resolution of these extreme materials challenges. This talk will
focus on the challenges associated with neutron radiation damage, high heat flux and highoperating temperatures on materials for potential steady-state demonstration fusion reactors, andhow multiscale materials modeling and experimental validation can expediently result in thedevelopment of high-performance radiation-resistant materials.

28

I-1
Progress Towards Steady-State Operation in NSTX Through Advances in Plasma Control
and Plasma-Wall Interactions
M.G. Bell, S.P. Gerhardt and the NSTX Team
Princeton Plasma Physics Laboratory
Spherical tokamaks for fusion energy development will operate at high-beta, and require excellent
confinement and non-inductive current drive to maintain a steady state. In pursuit of these goals,
the National Spherical Torus Experiment (NSTX) has developed improved control of its strongly
shaped plasmas, implemented routine feedback control of n = 1 error fields and resistive wallmodes and utilized extensive lithium coating of its plasma facing components. The combination ofthese techniques has produced non-inductive current fractions up to 70% which were sustained forseveral current-relaxation times. These discharges have also achieved the highest stored energy in
NSTX and, in different combinations of the toroidal and poloidal magnetic fields, the highest
sustained values of toroidal and poloidal beta. In these H-mode discharges, which became
essentially free of ELMs as a result of applying lithium beforehand, the scaling of energy
confinement with current is somewhat stronger than was evident in previous experiments, whereas
the toroidal field dependence previously observed is nearly absent. Passive stability to modes withtoroidal mode number n = 1 has been facilitated by shaping the plasma cross-section andmaintaining a broad pressure profile. Reducing intrinsic n = 1 error fields and active n = 1 mode
control using a set of non-axisymmetric coils to generate controlled n = 1 radial field perturbations
further improved reliability at high beta, although n = 1 kink and tearing instabilities in the plasmacore often terminated the high-performance phase. Simulations with several MHD stability codes
show that these particular modes can be avoided by maintaining an elevated central safety factor q.
The relaxed current profile in MHD-quiescent phase of these discharges is well modeled withessentially neoclassical physics, although it is necessary to take into account the effect of large
Toroidal Alfven Eigenmode (TAE) “avalanches” which occur repetitively in redistributing thebeam-driven toroidal current and thereby elevating the central q. This effect has been simulated
heuristically with the transport analysis and modeling code TRANSP by applying a transient radial
diffusivity to the unthermalized beam-ion component coincident with each observed burst of TAE
activity. The results of the simulations then match well the measured toroidal current profile andreproduce the observed drops in the DD neutron emission which is dominated by beam-target
reactions in NSTX. Using this physics model, simulations with TRANSP and the DCON stability
code show that the higher toroidal field and off-axis beam current drive in the planned NSTXUpgrade can produce a range of ideally stable, fully non-inductive plasma scenarios, and partiallyinductively sustained scenarios with toroidal beta exceeding 25%.
Work supported by US Department of Energy Contract DE-AC02-09CH11466

29

I-2
Recent TCV results – innovative plasma shaping
to improve plasma properties and insight
A.Pochelon for the TCV-Team
Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique desPlasmas, Association Euratom-Confédération Suisse, CH-1015 Lausanne, Switzerland
The TCV tokamak experiment is used to study the influence of unconventional plasma shapes oncore and edge confinement. In low collisionality L-mode plasmas with electron cyclotron
resonance heating (ECRH) confinement increases with increasingly negative triangularity δ, The
confinement improvement correlates with a decrease of the very inner core electron heat transport,
even though the triangularity of the plasma boundary quickly decreases towards the core, which
points to the effect of global – non-local – transport properties. This confinement improvement
remains for a large range of radial ECRH power deposition profiles. TCV is also used to study
negative triangularity effects in H-mode plasmas. In contrast to the L-mode results H-mode
confinement is known to improve towards positive triangularity, due to increasing pedestal heightwith triangularity, though plagued by large edge localised modes (ELMs). An optimum
triangularity can thus be sought between steep edge barriers (δ>0) with large ELMs, and improved
core confinement (δ<0). This is significant for a reactor, since it opens the possibility of having Hmode-like confinement time within an L-mode edge, or at least with reduced ELMs. ELMy Hmodes with δup<0 have already been obtained in TCV. Another approach to reduce the effect of
ELMs is the exploration of different divertor shapes, like for instance the snowflake divertor. The
characteristics of such a hexapolar null configuration are compared to those of the conventional
quadrupolar X-point divertor. TCV experiments confirm some of the advantageous properties ofthe snowflake configuration such as the distribution of the exhaust power on more strike points
than the two that characterize conventional X-point divertor. Plasma shaping may further help
understanding the coupling process of sawteeth to NTMs, leading to confinement degradation.
TCV experiments show that sawtooth crashes can almost instantaneously (<30μs) generateharmonics resonant at q>1 flux surfaces as resolved instantly by spatial Fourier analysis ofcontinuous toroidal magnetic probe arrays, potentially giving rise to 3/2 and 2/1 modes. These
examples demonstrate the effectiveness of plasma shaping, both to improve confinement and
MHD properties in the experiment, and to give further insight in stability or gyrokinetic
modelling.
This work was supported in part by the Swiss National Science Foundation.

30

PL-3
Overview of Experimental Results in the EAST Tokamak
X. Gong, J. Li, B.N. Wan and the East teamInstitute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China
The Experimental Advanced Superconducting Tokamak (EAST) is a fully superconducting
tokamak with a flexible poloidal field system to accommodate both single null (SN) and double
null (DN) divertor configurations, and its main mission is to establish steady-state highperformance plasma and study related physics and technologies. Since the first plasma was
successfully achieved on Sept. 2006, the new capabilities including actively water-cooled PFCs,
Heating & CD system, Plasma Control System, and Diagnostics have been developed andvalidated in the past five years in the EAST Tokamak. A series of wall-conditioning technologies
in superconducting machines (RF-DC, HF_GDC and Li-Coating) are explored, which will makeunique contributions and produce results of significance for ITER in the future. Integrated
operation scenarios (plasma startup, and ramp up/down) were focused on superconducting
tokomak and reliable plasma operation with Ip=1.0MA was obtained. In the last experimental
campaign on 2010, long pulse diverted plasma operations up to 100s were realized by LHCD of
0.8MW. Stationary H-mode plasma discharges, with either LHCD only or combined auxiliary
heating of ICRF and LHCD, have been achieved and sustained up to 6.5s. The power threshold for
H-mode access in EAST follows the international tokamak scaling and the energy confinement for
H-mode in LHCD has been identified. In addition, experiments of SOL & Divertor physics, firstevidence of the role of Zonal Flows for the L-H transition, intrinsic rotation and toroidal flow
driven in LHCD plasma, full non-inductive CD, effective ICRF heating, etc. are encouraging. The
future plan of EAST in next 3-5 years also will be presented in this paper.

31

I-3
Suppression of Edge Localized Modes in the DIII-D Tokamak with Small 3D Magnetic
Perturbations
T.E.Evans
General Atomics, PO Box 85608, San Diego, California 92186-5608, USA
Edge Localized Modes (ELMs), associated with large edge pressure and current profile gradientsin high confinement (H-mode) plasmas, are a significant concern for the next generation of large
tokamaks such as ITER where the stored plasma energy is expected to exceed several hundred
megajoules. The largest transient energy bursts driven by ELMs in ITER are predicted to releaseas much as 6%-7% of the plasma energy in less than a millisecond. This is expected to cause asubstantial increase in the divertor target plate erosion rate resulting in a reduction of theoperational lifetime of these essential components. In addition, the eroded divertor target platematerial may cause an increase in the influx of impurities into the core plasma that could limit thefusion power output of the device and trigger instabilities that are known to lead to a rapid
termination of the discharge. Over the last few years, a series of experiments done in the DIII-D
tokamak have demonstrated that large ELMs can be reproducibly suppressed over a wide range of
H-mode conditions and plasma shapes including ITER similar shapes with ITER pedestal
collisionalites. In these experiments, it is found that small 3D magnetic perturbations (b3D/btor ~2x10-3), primarily localized to the edge of the plasma, can be used to suppress the first ELM
following the L-H transition and to maintain an ELM suppressed discharge for up to 24 energy
confinement times, limited only by the heating duration of DIII-D neutral beams. During this time
the line average density, pedestal density, Zeff and core radiation rate all remain constant, incontrast to a conventional ELM-free H-mode, suggesting that robust steady-state H-mode
conditions, with an H-mode factor H89Y2 of approximately unity, have been achieved. ELM
suppression has been obtained using both n=3 and n=2 resonant magnetic perturbation fields fromthe DIII-D internal non-axisymmetric coils. The current requirements for ELM suppression andplasma response to the applied perturbation fields are most sensitive to the safety factor at the 95%
flux surface and the shape of the outer flux surfaces. In this talk, examples of the plasma
characteristics observed in DIII-D ELM suppressed H-mode plasmas will be discussed along witha summary of the operational space required for suppression and an overview of the physics
mechanisms that are believed to be important for obtaining ELM suppression in DIII-D.
Implications for scaling these results to ITER will also be discussed.
This work was supported by the US Department of Energy under DE-FC02-04ER54698.

32

O-1
Excitation of Beta-induced Alfven Eigenmode during Strong Interchange Mode in Large
Helical Device
W. Chen1), M. Isobe1),2), K. Toi1)3), K. Ogawa3), X.D. Du2), M. Osakabe1), S. Ohdachi1)2)and LHD experiment group1)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi 509-5292, Japan2)Department of Fusion Science, The Graduate University for Advanced Studies3)Dept. of Energy Science and Engineering, Nagoya University
The beta-induced Alfven eigenmodes (BAEs) were observed and investigated under different
conditions in tokamak plasma, including that driven by fast ions [1], energetic electrons [2], andlarge magnetic islands [3-5]. The BAEs are very important because they can be used as energy
channels to transfer the fusion-born-alpha-particle energy to the thermonuclear plasma.
In a helical plasma, the BAEs during strong interchange mode, whose mode-numbers are m/n=2/1,
have been recently observed for the first time. The first harmonic frequencies of these oscillations
range from 30 to 70 kHz in Large Helical Device (LHD) , much lower than the toroidal-Alfveneigenmode (TAE) frequency, and are provided with the same order of the low-frequency gap
induced by finite beta effects. The magnetic fluctuation spectrogram indicates that the BAEs often
occur in pairs, and their mode-numbers are m/n=2/1 and -2/-1. The analysis reveals that the modes
propagate poloidally and toroidally in opposite directions, and form standing-wave structures ininterchange-mode rest frame. The frequencies of the pair mode are associated with the Te/Ti ratio,
and the frequency difference of the pair modes is determined by the frequency of interchange
mode. The novel experimental results indicate that this phenomenon is very similar with that ofBAE during strong tearing mode in tokamak plasma, but the BAE frequencies do not completelydepend on the intensity of interchange mode. The new finding shed light on the underlying physics
mechanism for the excitation of the low frequency Alfvenic fluctuation. Comparing with thetearing mode, although the free energy, which drives the interchange mode, is different, thesteepen pressure gradients at the vicinity of resonant surface both potentially excite the type of
BAE instability.
References[1] W. W. Heidbrink et al., Phys. Rev. Lett. 71, 855 (1993).
[2] W. Chen et al., Phys. Rev. Lett. 105, 185004 (2010).
[3] P. Buratti et al., Nucl. Fusion 45, 1446 (2005).
[4] O. Zimmermann, et.al. Proc. 32nd EPS Conf. on PPCF, P4.059. (2005).
[5] W. Chen, et al., J.Phys.Soc.Jpn 79, 044501 (2010); ibid, Nuclear Fusion, 51,063010(2011).

33

O-2
Studies of MHD stability in Heliotron J Plasmas
S. Yamamoto, S. Kobayashi, S. Ohshima1), M. Suwa2), K. Yasuda2), K. Nagasaki, Y. Nakamura2),
T. Mizuuchi, H. Okada,T.Minami, K. Hanatani, S. Konoshima, B. Blackwell3), D. Pretty3)
and F. Sano
Institute of Advanced Energy, Kyoto University, Gokasho, Uji 611-0011, Japan
1)Center for the Promotion of Interdisciplinary Education and Research, Kyoto University,
Gokasho Uji 611-0011, Japan2)Graduate School of Energy Science, Kyoto University, Gokasho, Uji 611-0011, Japan
3)Research School of Physics and Engineering, The Australian National University, Canberra
ACT 0200, Australia
To clarify the MHD instabilities is important because the MHD instabilities could lead to thedeleterious displacement of the equilibrium as well as the increment of the anomalous transport,thus resulting in the degradation of plasma performance. Therefore, the MHD stability isextensively being studied in many tokamaks and helical systems.
The main purpose of Heliotron J, which is the middle size helical device with the pole number ofhelical coil l=1 and toroidal field period Np=4, is to optimize the magnetic configuration aiming at
the simultaneous improvements in particle transport and MHD stability. With regard to MHD
stability, the magnetic configuration of Heliotron J has a low magnetic shear in combination with amagnetic well by which MHD instabilities particularly at rational surfaces with low order can be
avoided or would be stabilized for pressure driven MHD instabilities. We experimentallyinvestigate the MHD stability of Heliotron J plasmas by utilizing the scan of rotational transformand magnetic well/hill in order to demonstrate the advantage of our concept with low magnetic
shear and magnetic well.
Two kinds of MHD instabilities are observed in Heliotron J plasmas. One is the low-n pressure
driven resistive interchange mode with m/n=2/1 and m/n=5/3 in the plasmas with low orderrational surface 0.5 and 0.6, respectively. When these MHD instabilities are observed, the bulk
plasma parameters such as plasma stored energy is saturated or decreased indicating that theyaffect global energy confinement. It should be noted that we have observed no coherent mode in
the plasma without rational surface. The other is the energetic ion driven MHD instabilities
including GAEs and/or EPMs in NBI heated plasmas. They might affect the energetic ion
transport from the results that some plasma parameters are simultaneously modulated with thebursting GAEs. We will also report the characteristics including spatial profile of these
instabilities and parameter dependencies in Heliotron J plasmas.

34

I-4
Progress in LHCD experiment and modelling to approach AT regime on Alcator C-Mod
S. Shiraiwa, P. T. Bonoli, I. Faust, A. Hubbard, O. Meneghini, R. R. Parker,
G. M. Wallace and J. R. Wilson1)
MIT Plasma Science and Fusion Center, Cambridge, MA, USA
1) Princeton Plasma Physics Laboratory, Princeton, NJ, USA
Assessing the performance of heating and current drive techniques at ITER relevant parameters is
critical for designing non-inductive current drive systems on that device and for other future
steady-state experiments. On Alcator C-Mod, excellent lower hybrid (LH) current drive (CD)efficiency (η = 2.0-2.5 × 1019 AW-1m-2) has been observed during fully non-inductive, reversed
shear plasma operation at line averaged density of 0.5-0.6×1020 m-3 and magnetic field of 5.4T,
close to what is expected in ITER steady state scenarios [1]. Some of these discharges exhibit an
ITB characterized by an abrupt increase of core electron temperature (r/a < 0.2). The ITB
formation coincides with reversal of magnetic shear about 200 - 300 ms after the LH turn-on. A
key issue to approach high fBS (> 50%) advanced tokamak regimes from these plasmas is tomaintain the good CD efficiency at high density (1.0-1.5 × 1020 m-3), where a significant fractionof LHCD power was observed to be absorbed either directly or indirectly in the SOL region,
especially in diverted configurations [2]. LHCD simulations using both ray-tracing and full wave
codes have been performed to better understand this unfavorable phenomenon and parasitic edge
losses were shown to be a candidate to explain the experimental results. Although multiple
mechanisms, such as fast electron diffusion and collisional absorptions of LH waves, cancontribute to such losses, these losses tend to become more important in multipass absorption
regimes, suggesting that stronger single pass absorption could be generic methods to improve CD
efficiency at high density. Indeed, recent discharges using ICRF mode-conversion heating toincrease the electron temperature (Te0 > 4 keV) indicate that such a recovery is possible. The
extensive modelling effort and direct comparisons with experiment improve our understanding of
LHCD, giving a solid guideline for planned experiments at twice the LHCD power (>2MW) on CMod and enhancing our capability to predict the performance of LHCD for future fusion devices.
References[1] S. Shiraiwa, et. al., to be published in Nucl. Fusion[2] G. Wallace, et. al., Nucl. Fusion, 51 083032 (2011)

35

O-3
Plasma Potential in Toroidal Devices: T-10, TJ-II, CHS and LHD
A. Melnikov, C. Hidalgo1), T. Ido2), A. Shimizu2), A. Fujisawa3)
Institute of Tokamak Physics, NRC Kurchatov Institute, 123182, Moscow, Russia
1)Asociación EURATOM-CIEMAT, 28040, Madrid, Spain2)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
3)Research Institute for Applied Mechanics, Kyushu University, Kasuga, Japan
Direct measurements of the potential φ have been undertaken in the T-10 tokamak [1] (B=1.5-2.5
T, R=1.5 m, a=0.3 m, PECRH<2 MW), and stellarators TJ-II [1] (B=1 T, <R>=1.5 m, <a>=0.22 m,
PECRH< 0.6 MW, PNBI< 0.9 MW), CHS [2](B=1 T, <R>=1.0 m, <a>=0.2 m, PECRH<0.6 MW,
PNBI<1.5 MW), and LHD [3] (B=1.5-3 T, R=3.6-3.9 m, <a>=0.63 m, P ECRH <0.6 MW, PNBI<19MW) using the Heavy Ion Beam Probing. L- and H-modes with and without plasma current wereconsidered. Despite the differences in machine sizes, heating methods and magnetic
configurations, and different expected mechanisms for the electric field Er generation in tokamak
and stellarators, the observed φ shows the striking similarities: (i) Similar magnitudes of Er; (ii)
For low densities, nе <0.5×1019 m-3 (unattainable in T-10), φ is positive, and an increase in nе isassociated with the decrease of positive φ and formation of a negative Er; (iii) For higher densities,nе>(0.5-1)×1019 m-3, both φ and Er tends to be negative despite different heating methods: Ohmicand ECR heating in T-10, ECRH and/or NBI in TJ-II, CHS and LHD; (iv) Application of ECRH,
causing a rise in Tе, results in more positive values for φ and Er. Analysis shows that forstellarators the main features of φ dependences on nе and Tе agree with neoclassical predictions
within the experimental and simulation precisions. Spontaneous and biased transitions to the Hmode, observed in TJ-II and T-10, are associated with more negative Er and broadband turbulence
suppression. Similar effects were observed in JFT-2M. Potentials on the four devices, similarities
and differences in the mechanisms of Er generation are discussed.
References[1] A. Melnikov et al., Nucl. Fusion 51 (2011) 083043[2] A. Fujisawa et al., Phys.Plasmas 7 (2010)4152[3] T. Ido et al., Plasma Phys. Control. Fusion 52 (2010) 124025
The work was supported by RFBR grants 10-02-01385 and 11-02-00667.

36

O-4
Optimizing the current ramp-up phase for the hybrid ITER scenario
G.M.D.Hogeweij, J.-F.Artaud1), T.A.Casper2), J.Citrin, F.Imbeaux1), F.Köchl3), X.Litaudon1),
ITM-TF ITER Scenario Modelling group
FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio
Cluster, P.O.Box 1207, Nieuwegein, The Netherlands, www.rijnhuizen.nl
1) CEA, IRFM, F-13108 Saint Paul lez Durance, France2) ITER Organization, F-13115 Saint Paul lez Durance, France
3) Association EURATOM-OAW/ATI, Atominstitut, TU Wien, 1020 Vienna, Austria
The current ramp-up phase of ITER is a critical stage: MHD instabilities have to be avoided, flux
consumption has to be minimized, and this has to be achieved within the narrow operational
window of ITER. Ramp-up for the hybrid scenario is more critical than for the standard (H-mode)
scenario, since the q profile must be shaped carefully: qmin should stay near or slightly above 1
and, for an optimized fusion performance, the q profile should have the typical hybrid shape with awide flat region [1]. This paper reports on a systematic effort within the ITER Scenarios
Modelling working group (ISM), part of the European Integrated Tokamak Modelling (ITM) Task
Force, to optimize the current ramp-up phase for the ITER hybrid scenario, and to assess thesensitivity of the results to the assumptions made.
Validation on the ramp-up phase of JET, AUG and Tore Supra [2] has shown that both empiricalscaling based models and the semi-empirical Bohm/gyro-Bohm model (L-mode version, ITB
shear function off) yield a good reproduction of this phase for considered discharges, in terms of
Te and q profile and li. Therefore these models have been used in the reported work, which was
carried out with the CRONOS integrated suite of codes.
Following assumptions from the ITER team were adopted: (i) An expanding ITER shape is usedwith early X-point formation at 3.5 MA; (ii) A flat Zeff profile is assumed, decreasing in time, as in[3]; (iii) A rather low density of ne = 0.25 neGw is taken; (iv) The ne profile is assumed parabolicwith a moderate peaking factor ne(0)/<ne> = 1.3.
The simulations start 1.5 s after breakdown, when Ip = 0.5 MA. The heating systems are usedwithin their designed limitations. The total input power is taken to be below the L-H threshold
during the whole ramp-up phase. Other assumptions (Te,i(edge), initial Te,i and li) are based on
experimental evidence. Regarding sensitivity of the results to the assumptions, following
parameters were varied: Te,i(edge) (by 40%); ne (by 60%), profile shape (parabolic vs. flat) andZeff.
The simulations show that the available heating systems allow the attainment of a hybrid q profileat the end of the current ramp-up. This is reached by using off-axis NBI, ECCD (from UPL) and
LHCD. A heating scheme with only NBI and ECCD is slightly less effective.
The optimum heating scheme depends on the chosen transport model. Moreover, the simulations
show that modified assumptions on ne peaking, edge Te,i and Zeff can be easily accounted for by ashift in time of the heating scheme. By post processing the simulation results with the freeboundary equilibrium code FREEBIE, run in Poynting mode, it has been checked that thereference case, both with and without additional heating, is safely within the boundaries put by
ITER coils.
References[1] J.Citrin et al, Nucl. Fusion 50 (2010) 115007[2] G.M.D.Hogeweij et al, Proc. 37th Eur. Conf., 2010, CD-ROM file P1.1041[3] V.Lukash et al, Plasma Devices and Oper. 15 (2007) 283

37

Poster Session 1(Monday 28th November)P1-1
Assessment of Plasma Performance in a Magnetic Configuration with Reduced Poloidal
Coils for a Helical DEMO Reactor FFHR-d1
J. Miyazawa, T. Goto, R. Sakamoto, G. Motojima, C. Suzuki, T. Morisaki, S. Masuzaki,
I. Yamada, H. Yamada, A. Sagara, and the FFHR Design Group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
The performance of high-density plasmas heated by neutral beam (NB) injection in a verticallyelongated configuration using 4 poloidal-coils has been investigated in LHD and shown to bebetter than that in the normal configuration using 6 poloidal-coils. Reduction of the poloidal-coil isfavorable in designing an LHD-type helical reactor FFHR-d1 [1], from both points of view of cost
reduction and realization of large maintenance ports. In the normal condition in LHD experiment,
the toroidally averaged plasma cross-section is kept circular by cancelling the quadrupole
components of the magnetic field for 100% with 6 poloidal-coils. The toroidally averaged plasma
cross-section becomes vertically elongated when the 2 of 6 poloidal-coils are not used, while theacceptable size of maintenance port is maximized. A better energy confinement was observed in
the circular configuration for low-density plasmas heated by electron cyclotron heating [2]. On theother hand, vertical elongation is effective to mitigate the Shafranov shift in high-beta plasmas [3].
In this study, the plasma performances in the vertically elongated and circular configurations havebeen compared for high-density NB heated plasmas, using the direct profile extrapolation (DPE)method, which has been developed to predict the radial profiles in fusion reactors from the profile
data obtained in the experiment [4]. Although no clear difference between the two configurations
is recognized in the relation between the central density and the central pressure, a factor Cexp usedin the DPE method, which is proportional to the reactor size, Rreactor, at a given reactor magnetic
field Breactor, as Rreactor ∝ Cexp Breactor-4/3, is smaller in the vertically elongated configuration. The
difference in Cexp between the two configurations is enhanced to > 10 % as the plasma beta
increases. This might be due to the mitigated Shafranov shift in the vertically elongatedconfiguration [3], which results in a larger plasma volume at a given plasma beta.
References[1] A. Sagara et al., ISFNT-10, 11 - 16 September 2011, Portland, US. 0138.
[2] H. Yamada et al., Plasma Fusion Res. 3 (2008) S1032[3] J. Miyazawa et al., Plasma Fusion Res. 3 (2008) S1047[4] J. Miyazawa et al., Fusion Eng. Des. (to appear)
This study has been financially supported by NIFS10ULFF006.

41

P1-2
Parameter dependence of MHD mode rotation in the Large Helical Device
Y.Takemura, S.Sakakibara1), K.Y.Watanabe1), K.Ida1), S.Ohdachi1), Y.Suzuki1), Y.Narushima1),
I.Yamada1), K.Narihara1), K.Tanaka1), T.Tokuzawa1), H.Yamada1)
SOKENDAI, Toki, Gifu 509-5292, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, JapanThe characteristics of MHD mode rotation have been investigated on Large Helical Device(LHD). It is widely known in tokamaks that MHD mode rotation is a key for maintaining the“stable” plasma state. In the case of tearing mode, reduction of the plasma flow induces the mode
locking, leading the disruption finally. Also, since the resistive wall mode, which limits theachieved beta, is destabilized by insufficient plasma flow. On the other hand, in helical plasmas,
the rotation mechanism of the interchange mode and the effect of the plasma flow on the stability
is unclear so far. In the heliotron configuration without net current, an interchange mode is one of
key instabilities because of magnetic hill formation. Especially, the modes excited in peripheral
region of plasma are dominantly observed when the beta increases and/or L/H transition occurs.
Several modes are excited in the plasma edge with the stochastic layer and they are key
instabilities for the high beta plasma production and/or the formation of the edge transport barrier.
In the previous experiments, the onset and parameter dependence of the modes have been
investigated in various configurations [1]. Here, these rotations were compared with the plasma
flow in order to understand the characteristics of mode rotation. The density scan experiment was
performed so as to change the poloidal plasma flow because the toroidal flow is almost negligible
in the peripheral region. The electron density was changed to 2-5 × 1019 m-3 and then the volume
averaged beta value, <β>, was maintained at about 1.5 %. The m/n=1/1, 3/4 and 2/3 modes where
the resonant surfaces are located at ρ > 0.9 were dominantly observed and rotate in the electron
diamagnetic direction in the laboratory frame. The mode rotation is compared with the electron
and ion flows taking into account the diamagnetic drift frequency. The experimental results showthat each mode rotation is consistent with the electron flow within the measurement error, whichsuggests that the observed mode freeze the electron fluid as well as the tearing mode [2]. The
plasma was heated by co-NB in the experiments, whereas it is confirmed that this tendency isalmost the same as the counter-NB case. In high beta discharges with <β> > 2 %, MHD modes
such as the m/n = 1/2 mode in the stochastic region (outside the last closed flux surface) are
excited typically. On discharges where the 1/2 mode dominantly observed, this rotation frequency
is close to the ion flow rather than electron one. As a result, this mode seems to rotate in the ion
diamagnetic direction in the plasma frame.
References[1] S.Sakakibara et al., Fusion Sci. Technol., 58 (2010) pp.176.
[2] O.Kluber et al., Nucl. Fusion, 31 (1991) pp.907.

42

P1-4
Study of Confinement Properties on Unfavorable LHD Configurations for Ideal MHD
Instabilities using TASK3D-code
K.Y.Watanabe, M.Sato, R.Seki, S.Sakakibara, Y.Narushima, M.Yokoyama, A.Fukuyama1),
N.Nakajima
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1) Graduate School of Engineering Kyoto University, Kyoto 606-8501, JapanThe Large Helical Device (LHD) has various magnetic field configurations, by changing thevertical field coils and the pitch parameters of the helical coils, which include the unfavorable ones
for the interchange type of ideal MHD instabilities. In deed, in the LHD experiments with the
MHD unfavorable configurations like the magnetic axis torus outwardly shifted configurations andthe high aspect configuration with either high magnetic hill and the low magnetic shear, the beta
collapses and the flat electron temperature profiles are observed. On the above discharges, thethermal transport analyses are done by the subset of the 3D integrated-transport simulation code,
TASK3D [1] in addition to a liner MHD instability analysis. The effects of the global MHD
instability on the confinement will be discussed from the comparative study between the transport
and the stability analyses.
Moreover, the above results are compared with the prediction of the pressure profiles calculated by
the MHD instability effects evaluation module on the confinement of the TASK3D,TASK3D/EQ+TR+MSSH [2], which leads to the validity check of the 'TASK3D/EQ+TR+MSSH 'module.
References[1] R. Seki et al., Plasma Fusion Res. 6, 2402081 (2011).
[2] M. Sato et al., Contrib. Plasma Phys. 50, 665 (2010).

43

P1-5
Impact of Te/Ti on energy confinement properties in a tokamak plasma

E. Narita1), T. Takizuka2), N. Hayashi2), T. Fujita2), S. Ide2), M. Honda2), Y. Tanaka1), T. Fukuda1)
1)Osaka University, Suita, Osaka 565-0871, Japan
2)Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan
Scaling expressions for thermal energy confinement time of tokamak H-mode plasmas have a very
important role in predicting the performance of future fusion devices. The IPB98(y,2) scaling law[1]
has been widely used. In the compiled database for IPB98(y,2), the ratio Te/Ti is mainly smaller
than unity. In future burning plasmas with strong alpha heating, on the contrary, it is expected that
Te/Ti is larger than unity.
The objective of the present study is to understand the impact of Te/Ti on confinement properties.
We have studied the updated DB3v10 database[1, 2], paying attention to the Te/Ti effect. The 67
data from AUG and 68 from C-Mod have been extracted from the database. For these data, thedensity is lower than 60% of the Greenwald density limit nGW. As a consequence, we haveobtained following scaling expressions in two density profile regimes.
τsc1 = 0.0777Ip0.91Bt0.11ne0.28PL-0.59R1.73M0.27ε0.54κ0.41 : 1.0<ne0/ne<1.1 (flat density case)τsc2 = 0.0594Ip0.90Bt0.08ne0.34PL-0.63R1.78M0.16ε0.41κ0.72 : 1.1<ne0/ne<1.6 (peaked density case)(τsc: s, Ip: MA, Bt: T, ne: 1019m-3, PL: MW, R: m, M: ion mass number, ε: inverse aspect ratio, κ:ellipticity, ne0/ne: ratio of central density to line averaged density) Examining the impact of Te/Ti
on these two regimes separately, we have found that it is weak for flat density case while it
becomes apparent for peaked density case. For the latter, we have further developed a scaling
expression including Te/Ti by adding 795 JET data. We incorporated 3 nondimensional variables[3]
1.25, namely BtR , ne/nGW*, and safety factor q in addition to Te/Ti. The dependences on ne/nGW*
and q were weak, and the following new expression was obtained:τsc3 = τsc2{1-0.157(Te/Ti-1)(BtR1.25)0.59}.
The standard deviation was reduced remarkably. Here, we find that the confinement timedecreases with increasing of Te/Ti[4].
The impact of Te/Ti on the plasma transport is investigated using the code TOPICS[5] with the
GLF23 transport module. The GLF model suggests that the transport is enhanced for large Te/Ti
domain[6]. Transport simulations for JT-60SA plasmas are carried out in a wide range of plasma
conditions, especially focusing on the ratio between heating power for electrons and that for ions.
Simulation results will be checked against above scaling analyses.
References[1]Progress in the ITER Physics Basis, Chapter 2: Plasma confinement and transport, Nucl.
Fusion 47 S18 (2007)[2]The international global H-mode confinement database
http://efdasql.ipp.mpg.de/hmodepublic/[3]T. Takizuka JAERI-Research 95-075 (1995)[4]E. Narita 2011 38th EPS Conf. on Plasma Physics (Strasbourg) P2.121
http://ocs.ciemat.es/EPS2011PAP/pdf/p2.121.pdf[5]N. Hayashi, et al., J. Plasma Fusion Res. 80, 605-613 (2004)[6]R. E. Waltz, et al., Phys. Plasmas 4, 2482 (1997)

44

P1-6
Effect of Nuclear plus Interference Scattering on Fast Alpha-particle Orbit and Confinementin DT Plasmas
D.Sawada, H.Matsuura, Y.Nakao
Kyushu University, Fukuoka, Fukuoka 819-0374, JapanTo realize the fusion reactor system it is important to understand the fast alpha-particle
confinement and slowing-down process in magnetic field, and hence many numerical simulations
have been made in various reactor systems [1-3]. In the previous alpha-particle orbit analyses,
however, the effect of nuclear plus interference (NI) scattering [4] on the alpha-particle behaviors
has not been included. The NI scattering is a non-Coulombic, large-angle scattering process, and alarge fraction of the fast-ion energy is transferred in a single event. In actual DT plasmas, some of
the alpha-particle orbits would be changed significantly due to the NI scattering. In such a case,confinement of the alpha-particle and plasma heating process may be influenced.
So far we have investigated the effect of NI scattering on plasma burning characteristics andplasma diagnostics scenarios on the basis of Boltzmann-Fokker-Planck model[5]. In this paper wenewly incorporate the NI scattering effect into the alpha-particle orbit simulation code system [6].
The NI scattering effect on the alpha-particle orbit in the magnetic confinement device and theinfluences on the alpha-particle confinement are examined. It is shown that due to the NI
scattering effect, fraction of lost energetic alpha-particle to total generation decreases to someextent. The detailed mechanism and effect of plasma parameters are discussed.
References[1] K.Tobita, et al., Nucl. Fusion 35 (1995) 1585.
[2] M.H.Redi. et al., Phys. Plasmas 4 (1997) 4001.
[3] V.Yavorskij, et al., Nucl. Fusion 43 (2003) 1077.
[4] J.J.Devaney and M.L.Stein ,Nucle. Sci. Eng. 46 (1971) 323.
[5] H.Matsuura, et al., Plasma Phys. Contr. Fusion 53 (2011) 035023.
[6] R.B.White, et al., Phys. Fluids 27 (1984) 2455.

45

P1-7
Non-modal renormalized gyrokinetic approach for time-dependent plasma shear flow
V.S.Mikhailenko1), V.V.Mikhailenko1), K.N.Stepanov2)
1) V.N.Karazin Kharkov National University, 61108, Kharkov, Ukraine2)National Science Center ”Kharkov Institute of Physics and Technology”, 61108, Kharkov,
UkraineThe investigation of the long-time mode of behavior of the regimes of the improved confinementof plasma, where ExB shear stabilization mechanisms play a key role, is the theme of major
importance in the physics of the continuous operation of the fusion systems. In spite of the great
progress in gyrokinetic treatment of tokamak drift turbulence, there still remain key issues inanalytical investigations of the long-time evolution of the turbulence in plasma shear flows.
New approach is developed and applied to analyze a temporal evolution of drift ITG turbulence of
inhomogeneous plasma flow across the magnetic field with time-dependent velocity shear in planegeometry. In contrast to typical gyrokinetic treatment, we use a method of shearing modes (alsonamed as ‘nonmodal’ approach) which consists in transforming of Vlasov-Poison (V-P) system tosheared (in space and velocity) coordinates convected with shear flow and accounted for the effectof the waves stretching by shear flows. This transformation is followed by transformation torenormalized guiding center and Larmor orbit coordinates, which accounted for the effect of
turbulent scattering of particles by ensemble of shearing modes with time dependent wave
numbers. This renormalization procedure precedes the standard gyroaveraging and retains effect
the gyration angle turbulent scattering in gyroaveraged V-P system. That approach allows strong
time-dependent velocity shear rate to be treated, while retaining the effect of turbulence.
Developed theory reveals characteristic time scales and corresponding evolutionary processes,
which display different statistic properties. It was obtained, that main nonlinear effect, which isabsent in conventional gyrokinetic theory, but is responsible for extremely fast suppression of theinstability, is effect of the gyration angle scattering by sheared perturbations.
The renormalized non-modal quasilinear equation, which accounted for the effect of ionsscattering by ensemble of the sheared drift waves on the ion equilibrium distribution function, isobtained. It is applied to the investigations of the multistage processes of the temporal evolution of
the anomalous transport of ions density and energy in plasmas with time dependent velocity shear.
The time dependent diffusion and thermal conductivity coefficients, which display fast decay withtime due to velocity shear, are obtained.

46

P1-8
Reduced Size LHD-type Fusion Reactor with D-shaped Magnetic Surface
T.Watanabe
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan

LHD-type magnetic configuration (Heliotron configuration) is produced by continuous helical andvertical coil systems. LHD experiments have achieved an average beta value of 5 % without beta
collapse. Hence, LHD-type fusion reactor is indicated the possibility of compact, high-output andhigh-performance, in steady state.
LHD-type reactors require a large major radius to attain the self-ignition condition with asufficient space for blankets. To reduce the major radius size, small helical pitch parameter (γ)
configuration is considered[1]. Small γ configuration requires relatively high current density forhelical coils. The volume of the last closed magnetic flux surface(Vlcfs) of vacuum field becomes
small, because the magnetic surface of the present LHD is located close to the inboard-side of
vacuum vessel wall, with elliptic shape.
In the present paper, we propose a new winding law for the helical coils: helical coils winding
along the geodesic curve of a torus. It become possible that the magnetic axis of the magnetic
surface with excellent characteristics coincide with the center of the helical coils. The magnetic
surface in the vertically elongated cross section becomes D-shaped, so that wide blanket spacebecome possible. This indicates that the LHD-type fusion reactor can be achieved in reduced size.
References[1]S.Imagawa, A.Sagara and Y.Kozaki, Plasma Fusion Res. 3 (2008) S1050

47

P1-9
Another approach to problem of controlled nuclear fusion
A.F.Nastoyashchiy
RF SRC Troitsk Institute for Innovation and Fusion Research (TRINITI), 142190 Troitsk Moscow
Reg., Russia
Currently, in research on controlled nuclear fusion tokamaks play the dominating role. Here, theessence of the approach to solving problem is to “detach” the plasma from the walls (ββ≪1), tocreate a “vacuum” gap between the plasma and the wall, thereby preventing the destruction of thewalls. However, we know from experience that at ββ≪1 the first wall is rapidly destroyed. The
wall damage and disruption instability are those vulnerable aspects which do not permit, withcomplete confidence, considering tokamaks as a foundation for future fusion power setups. Hence,
there naturally, arises a question: if the chosen way to develop tokamaks, i.e. permanent increaseof their sizes (and cost) the only one correct? Or are there other, more simple solutions? We offer
an alternative approach to the problem, which is appropriate neo-magnetic confinement, whenunlike the conventional magnetic confinement the magnetic field is mainly used for thermal
insulation of the plasma, while the pressure of plasma itself is confined by a high-pressure gas.
This will make it possible to reduce significantly the magnitude of the magnetic field and simplify
the complex. Besides, the formation of a powerful plasma-gas layer near walls [1] should
contribute largely to better confinement and more reliable protection of the wall. Such
confinements regimes are possible due to the fact that during fusion burning in a magnetic field,
the “naturally boundary” of the plasma, where the plasma temperature formally vanishes, is at a
finite distance from the region with a maximum temperature. As will be show in the paper, underconditions of neo-magnetic confinement we deal with interesting and practically importantphenomenon of a self-sustaining magnetic field predicted by author [2]. This phenomenon can
provide effective thermal insulation of the plasma. In the framework of the proposed concept, wehave formulated a simplified physical model that allows an approximate analytical study of the
problem of a self-sustaining fusion reaction. We consider the possibility of solving a particularimportant task – the creation of a deuterium reactor. By the example of the toroidal geometry of
the plasma, we consider the physical conditions of the plasma, and present the estimates of thepossible parameters of a fusion power reactor, using DT-, DD- and D3He- reactions.
References[1] A.F.Nastoyashchiy. J. Plasma & Fusion Res. SERIES 7 (2006) 179.
[2] A.F.Nastoyashchiy. Physical concept of neo-magnetic confinement of plasma. Preprint
142A. TRINITI, Troitsk, Russia, 2010.

48

P1-10Laser nuclear fusion on the basis of mini-torus shaped target
A.F.Nastoyashchiy
RF SRC Troitsk Institute for Innovation and Fusion Research (TRINITI), 142190 Troitsk Moscow
Region, Russia
Successful solutions of the problem of harnessing nuclear fusion energy based on inertial
confinement depend, first and foremost, on the choice of the target design. Early studies in thisapproach [1] proposed several simple target variants. There were considered target where a heavy
shell was used to confine hot high-dense plasma spread [2] and proposed to amplify the reactionenergy yield by initiating a thermonuclear combustion wave [3]. Also, there was considered theeffect of magnetic field [2, 4] on the thermal insulation of plasma in the case of a cylindricalgeometry of the target. According to estimates [2] a self-sustaining thermonuclear reaction in thetarget with a shell required a laser pulse energy of ~ 1 MJ if the magnetic field B ~ 1 MG wasused. At the same time, for the cylindrical target, the technical difficulties due to the need to“close” the butt ends of the target arise. After Teller’s group [5] published the concept ofprogrammed laser compression of the spherical target, practically all researches switched to thisline. In view of their preliminary estimates, only ~ 10 kJ was required to ignite some ideal target,
and this made such models very attractive. More careful subsequent investigations showed that, infact, a much higher energy of ~ 4-5 MJ was required to ignite the target, if to take into account thereal physical conditions. Herewith, no less than 100 laser beams are required to provide thesymmetrical compression of the target. Thus, the assumed reactor facility becomes toocomplicated.
In this paper a simple target in the form of a mini-torus is considered, in which the heavy shell isused to confine the plasma spread. A significant decrease of heat losses is achieved using an
external magnetic field and/ or as result of a self-sustaining magnetic field that is generated in theplasma [5]. A mini-torus has exceptional advantages in comparison with other geometrical shapes(sphere, cylinder etc.) because the magnetic field can be most efficiently used here. In fact, themini-torus is an ideal theoretical model, since this target has not butt ends and, thus, the heat lossesof the plasma itself in the axial direction are excluded as compared with, e.g., a cylinder target. On
the basis of an approximated physical model, the conditions for “ignition” are discussed. It isshown that the ignition energy can be significantly reduced up to 50-100 kJ if the mini-torus
shaped target with magnetized plasma is used. It is especially important that, in contrast to targets
with programmed compression, the considered mini-torus target does not require a complicatedmulti-beam system of laser pulses.
References[1] B.B.Kadomtsev, N.G.Kovalsky, A.F.Nastoyashchiy. “Lasers and thermonuclear fusion”,
in: Lasers and thermonuclear problem, ed. B.B.Kadomtsev. M., Atomizdat (1973).
[2] A.F.Nastoyashchiy. At. Energy 32 (1972) 43.
[3] A.F.Nastoyashchiy, L.P.Shevchenko. At. Energy. 32 (1972) 451 (1972).
[4] J.M.Dawson. Production of plasmas by Long-Wavelength Lasers. Preprint MATT-786.
Princeton University. Princeton, New Jersey (1970).
[5] A.F.Nastoyashchiy. J. Russian Laser Research 32 (2011) 212.

49

P1-11
Design of a vacuum pumping system for a closed helical divertor for steady state operationin LHD
M.Shoji, S.Masuzaki, T.Morisaki, M.Kobayashi, M.Tokitani, Y.Takeiri
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Requirements for a vacuum pumping system in future fusion reactors are helium ash removal,
impurity reduction in the core plasma, neutral particle pumping for active control of the peripheral
plasma density profile. Efficient particle pumping is also required for steady state operation infusion reactors.
Two test modules of a closed helical divertor (CHD) were installed in the inboard side of the torusof the Large Helical Device (LHD) in the last experiment campaign (2010y). The CHD consists ofthree components: slanted divertor plates, a roof shaped dome and target plates. It successfullyenhanced the neutral particle density behind the dome by more than one-order of magnitude
compared to that in the original open divertor [1]. In the next experimental campaign, a vacuum
pumping system will be introduced behind the dome along the space between two helical coils in
the inboard side. Because the minimum distance between the divertor plates to the inlet of thevacuum pumping system is very short (~0.1m), heat loads by radiation from the divertor plates andby thermal conduction due to neutral particles have to be carefully considered for designing thepumping system.
The particle pumping efficiency and the heat loads on the pumping system were investigated using
finite element method based software for multi-physics analysis (ANSYS) and a fully threedimensional neutral particle transport simulation code (EIRENE) [2]. It proposed a possible
candidate which is composed of water cooled (WC) blinds, liquid nitrogen (LN2) cooled chevrons
and gas/liquid helium (LHe) cooled panel. The investigation indicates that the heat load onto the
LHe cooled panel due to radiation is not dominant. It also shows that buffer plates on thewater/LN2 cooled components and on the inner vacuum vessel significantly reduce the heat load
due to the conduction of neutral particles without serious degradation of the pumping efficiency[3]. The detailed analysis using the neutral particle transport simulation code predicts that amodified structure of the WC components enhances the pumping efficiency. An enlarged WC
blind inlet and a slit at the bottom of the WC component are effective in improving the pumping
efficiency. This paper reports calculations of the pumping efficiency for modified WC components
and propose an improved design of the pumping system for steady state operation in LHD.
References[1]S. Masuzaki et al., Plasma Fusion Res., 6 (2011) 1202007[2]D. Reiter et al., Fusion Sci. Technol., 47 (2005) 172[3]M. Shoji et al., Plasma Fusion Res., 6 (2011) 2401035

50

P1-13
1D model study on the effect of impurity radiation cooling on LHD SOL plasma
G. Kawamura, I. Murakami, Y. Tomita, S. Masuzaki
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Recent progress in fusion plasma performance gives us perspectives on future fusion devices like
DEMO and helical-type reactor, FFHR[1]. The increasing heat load to the plasma-facing wall inthese devices, however, can exceed the engineering limit if the same operation is carried out. The
material of divertor plates in reactor-class devices, i.e. tungsten, reduces carbon impurity in theplasma. Although it is beneficial for the energy confinement, the heat load has to be reduced byother means. One of possible candidates is gas puffing of impurity such as neon and nitrogen. The
cooling effect of the gas puffing with impurity radiation has been demonstrated in the 14th
experimental campaign of LHD in 2010 to 2011. It is found that the temperature and particle flux
onto divertor plates is reduced significantly. The physical understanding of the plasma response isdesirable to apply the gas puffing to the future operation. The cooling effect of neon on the SOL
plasma is studied here by means of 1D two-fluid model based on our previous model[2]. Since
sustaining radiation is observed after an instantaneous gas puffing in the experiment, the model
treats a steady state plasma. The radiation cooling is modeled as a source term in the balance
equation of energy flux. Another role of the impurity as a plasma source is ignored under theassumption of small amount of impurity, e.g. a few percent. Numerical solution for various
connection length from meters to kilometers in the LHD ergodic layer was obtained. A significant
reducing effect on particle flux and temperature was recovered. The physical mechanism of the
reduction is discussed in the paper.
References[1] A. Sagara et al., Fus. Eng. Des. 83 (2008) 1690
[2] G. Kawamura et al., Plasma Fus. Res., 5 (2010) S1020

51

P1-14
Effect of divertor target geometry on detached plasma in divertor simulator TPD-SheetIV
T.Iijima1), A.Tonegawa1), H.Komukai1), N.Mine1), K.Kawamura1) and K.Sato2)
1) Department of Physics, School of Science, Tokai University, 411 Kitakaname, Hiratsuka,
Kanagawa, 259-1292, Japan
2) Chubu Electric Power Co.Inc. 20-1, Kitasekiyama, Ohdaka-cyo,459-8522, JAPANIn order to achieve the high performance plasma for high power and long pulse operation, thedivertor design for stable detached plasma formation should be optimized to handle such high heatand particle fluxes. In JT-60SA or ITER, the bottom part of the divertor chamber forms a distinct
corner (V-shaped) with the target was proposed for high gas conductance between the divertor
legs. Recently, the closed Helical Divertor (HD) in LHD is planned to accomplish an active
neutral particles control to improve plasma confinement and to sustain high performance longpulse discharges. Therefore, the divertor target geometry to be compatible with the highperformance plasma is one of key significant issues on detached plasma.
In this study, we present the experimental simulation of the deivertor target geometry via detachedplasma formation of hydrogen plasma in a linear divertor plasma simulator TPD-SheetIV[1].
Three types of target geometry (V-shaped(long), V-shaped(short) and oblique targets) have been
investigated with steady state or pulse flow plasma. Measurements of the electron density, ne, theelectron temperature, Te, and Balmer series emission intensities were carried out in hydrogendetached plasma with hydrogen gas puff. The ionization and recombination events are discussed
by Collisional-Radiative model. In the V-shaped target, detached condition with high radiation
loss is produced easily in steady state and pulse flow plasma. Also, Te and Q rapidly decrease withincreasing gas pressure. The V-shaped target enhances the recycling and detachment plasma isattained there effectively.
References[1] A. Tonegawa et al., J. Nucl. Mater., 313-316, 1045(2003).

52

P1-15
Calculation of Ion Orbits in the Divertor/Dipole Regions of the GAMMA10 A-divertor
I.Katanuma, S.Masaki, S.Sato, K.Sekiya, M.Ichimura, and T.Imai
Plasma Research Center, University of Tsukuba
The design of the divertor mirror cell, which is to be installed in the present GAMMA10 tandem
mirror, (called tentatively the GAMMA10 A-divertor) is in progress[1,2].
The divertor mirror cell plays a role of exhausting the ions diffused radially in the central cell tothe dipole region, where the divertor plate is installed.
That is, the GAMMA10 A-divertor is planning to perform the simulation experiments such as thedivertor of a big torus.
The calculation of ion orbits, therefore, is important to know how ions move into divertor mirror
cell and then go into dipole region through x-point.
The numerical code for calculation of ion orbits was developed for the above purpose.
In order to determine the magnetic field B at the local ion position, numerical meshes wereintroduced.
The numerical meshes correspond with the flux coordinates (ψ, θ, χ), where one axis χ is taken
along a magnetic field line and B = ∇ ψ × ∇ θ = ∇χ.
The great merit of these meshes is to be able to introduce the electrostatic potential easily as afunction of ψ, θ, χ.
The energy and angular momentum of ions in the code are conserved with good enough accuracy.
The 10,000 ions at the mirror throat initially were followed in the GAMMA10 A-divertor with adivertor plate of the dipole region.
Almost all ions enter the dipole region due to the non-conservation of magnetic moment around xpoint.
The ions stay near x-point for a long time and penetrate into the dipole region irregularly and thenhit the divertor plate finally.
The mean loss time of ions to divertor plate is a few msec depending on the size of the divertor
plate.
As a summary the code was developed in order to investigate the behavior ions in thedivertor/dipole regions of the GAMMA10 A-divertor and the behavior of ions is being made more
and more clear.
References[1] I.Katanuma, K.Yagi, Y.Nakashima, M.Ichimura and T.Imai, Phys. Plasmas Vol.17,
032303 (2010).
[2] I.katanuma, K.Yagi, Y.Haraguchi, N.Ichioka, S.Masaki, M.Ichimura, T.Imai, Phys.
Plasmas Vol.17, 112506 (2010).

53

P1-16
Effects of Radial Losses of Particle and Energy on the Stability of detachment Front in a
Divertor Plasma
S. Togo, M. Nakamura1), Y. Ogawa, K. Shimizu2), T. Takizuka2), K. Hoshino1)
Graduate School of Frontier Sciences, 5-1-5 Kashiwanoha, Kashiwa, Chiba 277-8568, Japan
1)Japan Atomic Energy Agency, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho, Kamikita,
Aomori 039-3212, Japan
2)Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193, Japan
Operation with partially detached divertor plasmas [1] is considered to be a promising way to
reduce divertor heat loads. Although many theories and modelings of PDD in tokamak devices
have been developed so far, physics of formation and stability of the detachment front has notbeen fully understood yet. Recently the cross-field transport of plasma was found to be an
important factor for the formation of PDD [2] and for the stability of the detachment front [3].
In the present paper we further study the impact of cross-field transport on PDD plasmas. We
extend the previous works in [3] where a “two-layer” one-dimensional model for PDD plasmas
was proposed. Particularly, we analyze effects of the radial loss of particle and energy toward thefirst wall from a SOL-divertor region, which were not taken into account in the previous work.
Radial diffusive fluxes of plasma particle and heat in a SOL-divertor region towards a first wall,Γloss and qloss, can be described in terms of the cross-field particle and heat diffusion coefficients,
D and χ, and the radial decay lengths of the plasma density and temperature, λn and λT. In the onedimensional model framework in [3], such plasma particle and energy losses can be described assink terms in the plasma-fluid equations. As for the neutral particles, their spatial profile in the
direction parallel to the magnetic field was given by a simple exponential decay model [3]. In thepresent work we take into account the radial transport of neutral particles. We describe this neutral
particle transport with a characteristic radial transport length λN0 which is an input parameter in thepresent work. Comparison of the numerical results with the experimental results in ASDEX [4] isalso presented.
References[1] G. F. Matthews, J. Nucl. Mater. 220-222, 104 (1995).
[2] M. Wischmeier et al., J. Nucl. Mater. 390-391, 250 (2009).
[3] M. Nakamura et al., Plasma Fusion Res. 6, 2403098 (2011).
[4] A. Kallenbach, et al., Nucl. Fusion 48, 085008 (2008).

54

P1-17
Hydrogen Isotopes Retention in the Tile Gaps of JT-60U
M. Yoshida1), T. Tanabe1),T. Hayashi2), T. Nakano2), J. Yagyu2), Y. Miyo2), K. Masaki2),
K. Itami2)1)Kyushu University, Kasuga, Fukuoka 816-8580, Japan
2)Japan Atomic Energy Agency, Mukouyama 801-1, Naka-shi, Ibaraki-ken 311-0193, Japan
Evaluation of fuel retentions in the tile gaps, or side surfaces of plasma facing tiles, is urgently
needed for modeling a fuel retention build-up in fusion devises. In this study, retention ofhydrogen isotopes (Hydrogen (H) and Deuterium (D)) on the side surfaces of plasma facing
carbon tiles used for first wall in JT-60U were measured by TDS.
The tiles were used in experimental campaigns from 1992 to 2004, and their plasma facing
surfaces were predominantly eroded. Discharges in JT-60U were mostly with deuterium (D)plasma (DD discharge). At the end of the each series of the DD discharges, discharges withhydrogen (HH discharge) were performed to remove tritium (T) retained on the plasma facing
surface. For the measurement of the H and D retentions by TDS analysis, two samples were takenfrom each side surface of the tiles. One was from the entrance side of the gap (front side) and theother from the bottom side of the tile (bottom side).
TDS spectra obtained for the samples both the front and bottom sides showed broad desorbed
peaks of HH, HD and DD with each maximum at around 900 K. Integrating the total desorbed
amount for the present TDS spectra, hydrogen isotopes (H+D) retentions in the samples of thefront and bottom sides were 1~3 × 1022 H+D atoms/m2 and 0.8~1.9 × 1022 H+D atoms/m2,
respectively. Comparing with previously obtained TDS spectra for plasma facing surface of thetile, the H+D retentions on the side surfaces were appreciably smaller. Interestingly, D/H ratios in
the retention on both the side surfaces and the plasma facing surfaces were nearly the same with avalue of ~1. In previous works [1], it has been shown that retained D on the plasma facing surface
is partly replaced by H during the HH discharges and the value of D/H ratio is simply determinedby the tile temperature. Furthermore, the H+D retentions (1~3 × 1022 H+D atoms/m2) on the sidesurfaces were similar and/or smaller than those on the plasma facing surface (~3 × 1022 H+D
atoms/m2). All these results suggest that the temperature rise of the first wall was volumetric andthe temperatures of the plasma facing surface and side surfaces were nearly the same andaccordingly the retention mechanism for both would be similar, though the former were eroded
while the latter re-deposited. Although hydrogen isotopes (H and D) retention on the all sidesurfaces of each tile was smaller than those on the plasma facing surface, the integrated retention
on the side surfaces over whole first wall tiles would be ~1/10 of the total fuel retention of JT60U.
References[1] Y. Hirohata, et al. J.Nucl.Mater. 363-365 (2007) 854-861

55

P1-18Tritium retention to the first wall of JT-60U
A. Adachi1), M.Yoshida1), T.Takeishi1), T.Tanabe1), T.Hayashi2), T.Nakano2), M.Fukumoto2),
J.Yagyuu2), Y.Miyo2), K.Masaki2), K.Itami2)1)Kyushu University, Fukuoka, Fukuoka 812-0053, Japan
2)Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193, JapanFor economic and safety reasons, tritium accumulation on plasma-facing wall of nuclear fusion
reactor is strictly limited. Since the carbon tile is one of the main candidates for the use in vacuum
vessel components of fusion reactor, it is important to get a clear understanding of how tritium istrapped by carbon.
In this study, the amount of tritium retained in the graphite tiles used as the first wall of JT-60U
was measured by the full combustion method. All the samples used here were the surface tiles
located at the equatorial position of the first wall. Each examined tile was exposed to a different
DD discharge period respectively in operation years of 1991 to 2003. At that time, T atoms
produced by DD reactions were ranging from ∼1018 to ∼1020 T atoms. For experimental
preparations, samples were cut into 10×10×1 mm3 size pieces including plasma facing surface ortaken from bulk. The samples were heated up to 1000°C for the complete combustion and purged
by humid argon gas. Released tritium(HT) from each sample was transferred to a copper oxide bed
at 350°C to convert HT to the tritiated water (HTO) and collected to the dual pure water bubblers.
It was found that T was only retained near plasma facing surfaces sides of the carbon tiles and theamount of retained T increased from ∼1011 to ∼1013 T atoms/cm2 with increasing the exposed
discharge period of the tiles. Integrating the T retention with the total surface area of first walltiles, T retention rate in the first wall was determined to be 13 % of the total tritium production
rate calculated by neutron production rate. This value agrees with the previous estimation by atritium imaging plate technique [1] All these confirms that the part of energetic tritium producedby DD reaction were implanted within a few μm from the eroded plasma facing surface andretained there without being replaced by H during HH discharges performed after each DD
discharge campaign.
References[1] M. Yoshida, et al. Fusion Science and Technology, in press56

P1-19Toroidal distributions of gas retention and impurity deposition for 13th experimental
campaign in LHD
N. Shinoda, N. Ashikawa1), Y. Nobuta, Y. Yamauchi, T. Hino, S. Masuzaki1), A. Sagara1),
LHD experimental group1)
Hokkaido University, Kita-13 Nishi-8, Kita-ku, Sapporo 060-8628, Japan
1)National Institute for Fusion Science, 322-6 Oroshi-sho, Toki 509-5292, Japan
Plasma wall interactions (PWIs) such as impurity deposition and fuel gas retention are key issues
for the improvements of confinement and safety operation in fusion devices. Material probe
analysis is very useful for evaluation of the PWIs in the fusion devices [1]. In the present study,
PWIs during 13th experimental campaign in the Large Helical Device (LHD) were investigated by
the material probe analysis. The toroidal distributions of gas retention and impurity deposition
were discussed based upon the plasma discharges and the positions of plasma heating.
The material probes made of silicon and 316L SS were placed on the wall at every toroidal sector.
These were exposed to the plasmas in the 13th experimental campaign. The numbers of hydrogenand helium main discharges were 3200 and 1200, respectively. Hydrogen and helium glow
discharge cleaning (GDC) was conducted for 133h and 36h, respectively. After the campaign, thematerial probes were taken out and then the change of surface properties was analyzed. The
toroidal distributions of gas retention, impurity deposition and change in surface morphologies
were investigated by using thermal desorption spectroscopy, Auger electron spectroscopy andscanning electron microscope, respectively.
The boron deposition was large at near the anodes, and the deposition of carbon was large in thevicinity of tangential NBI.
The hydrogen desorption behaviors largely depended on the impurity deposition on the material
probe. Namely, most of material probes with a low impurity deposition, showed a sharp desorption
at around 950K. The material probes with thick carbon and boron, showed a broad peak at thelower temperature regime.
The amount of retained hydrogen was large near the anodes for GDC. This result indicates thehydrogen is implanted mainly during the GDCs. On the other hand, the amount of retained helium
was small near the anodes, perhaps due to the small discharge times.
References[1] T. Hino et al, Nuclear Fusion, 44, 496(2004).

57

P1-20
Impurity Effects of Hydrogen Isotope Retention on Boronized Wall in LHD
K. Matsuoka1), M. Kobayashi1), K. Kawasaki1), T. Fujishima1), Y. Miyahara1),N. Ashikawa2),
K. Nishimura2), A. Sagara2), Y. Oya1) and K. Okuno1)
1) Radioscience Research Laboratory, Shizuoka University, Shizuoka, 422-8529, Japan
2) National Institute for Fusion Science, Toki, 509-5292, Japan
Boronization is thought to be one of useful wall conditioning techniques in the Large Helical
Device (LHD) of the National Institute for Fusion Science (NIFS) for steady state plasma
operations. It is considered that boron is deposited on the first wall with the impurities andenergetic hydrogen isotopes escaped from plasma will implant into the boron film. Therefore, theretention behavior of hydrogen isotopes in impurities containing boron film should be elucidated
for the evaluation of hydrogen isotope recycling in fusion devices. In our previous studies, it was
found that the retention of hydrogen isotope trapped by impurities was enhanced by hydrogenplasma exposure due to existence of impurities from plasma into the boron film, which would
strongly affect on hydrogen isotope retention. In this study, the chemical states of the boron films
boronized in LHD were investigated and the correlations between the chemical states of the boron
films and the hydrogen isotopes retention behaviors were discussed.
The LHD boron films were prepared by boronization in LHD and preheated to remove residual
impurities and hydrogen. Thereafter, the samples were exposed to hydrogen plasma in LHD or
implanted to D2+ at Shizuoka University. The chemical states of boron films before and afterhydrogen plasma exposure or D2+ implantation were measured by X-ray photoelectron
spectroscopy (XPS) and the hydrogen retention in the boron films were investigated by Thermal
Desorption Spectroscopy (TDS).
The XPS measurement showed that the impurity concentrations of oxygen and carbon were 7 and15%, respectively. The deuterium was bound to boron as B-D-B and B-D bonds for pure boron
film, whose desorption temperatures were around 500 K and 700 K, respectively. In the case of
LHD boron film, the additional deuterium desorption stages were observed around 800-1100 K,
assigning to desorption of deuterium trapped by carbon as B-C-H and oxygen as B-O-H bonds.
70% of total hydrogen retention for the hydrogen plasma exposed boron film was retained by
impurities. After exposure to hydrogen plasma, the atomic concentrations of impurities wereincreased, resulting in the increase of hydrogen retention trapped by carbon and/or oxygen. It wasconsidered that hydrogen isotope retention would increase with increasing the operation period of
D-D plasma, which leads the increase of tritium inventory in LHD.

58

P1-21
Tritium Retention on Stainless Steel Surface Exposed to Plasmas in LHD
M. Matsuyama1), J. Saikawa1), K. Nishimura2), N. Ashikawa2), Y. Oya3), K. Okuno3),
T. Hino4) and A. Sagara2)
1) Hydrogen Isotope Research Center, Univ. of Toyama, Gofuku 3190, Toyama 930-8555, Japan
2) National Institute for Fusion Science, Toki 509-5292, Japan
3) Radiochemistry Research Laboratory, Shizuoka Univ., Shizuoka 422-8529, Japan
4) Laboratory of Plasma Physics and Engineering, Hokkaido Univ., Sapporo 060-8628, Japan
From viewpoints of controlling particle balance in the reactor core and the amount of tritium
retention in the PFM’s, it is of great important issue to clarify the retention behavior of tritium in
PFM’s exposed to plasmas. In this study, we have studied retention and release behavior of tritiumon the surface of stainless steel samples, which were specially fixed on the plasma-facing surface
in the Large Helical Device (LHD) and were exposed to plasmas for one cycle.
Small sample plates used in this study were fixed at four different locations on inner walls of LHD
before start in each experimental cycle (12th, 13th and 14th cycle): namely, the locations are 1.5U,
5.5U, 6.5L, and 9.5L. The samples were analyzed by X-ray photoelectron spectroscopy and laser
Raman spectroscopy. Behavior of tritium retention and release was examined using &beta-ray
induced X-ray spectrometry.
Results of surface analyses showed that surfaces of all samples were covered with the deposition
layers, and those of 1.5U and 5.5U samples were thinner than that of 6.5L and 9.5L. Constituent
elements (Fe, Cr and Ni) of the stainless steel were observed in the all samples. Large difference inthe order of tritium retention was observed in the location of sample and experimental cycle. The
order was as follows: 1.5U<6.5L~5.5U<<9.5L for the 12th cycle, 1.5U<5.5U<9.5L<<6.5L for the
13th cycle, and 1.5U<5.5U<9.5L<<6.5L for the 14th cycle. Maximum tritium retention in the 6.5L
and 9.5L samples was ten times and more in comparison with other samples.

59

P1-22
Dust Dynamics Released from Plasma-Facing Components in HL-2A Tokamak
Z.H. Huang, Y. Tomita1), G. Kawamura1), N. Ashikawa1), Y.D. Pan, L.W. Yan
Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan, China
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Dynamics of a dust particle with a spherical shape is studied in the SOL/divertor plasma of the HL-2A tokamak with a lower single-null closed divertor configuration [1]. The dust particle inrelatively low temperature plasmas at SOL/divertor region is negatively charged due to the highmobility of plasma electrons. The charging time is as fast as the order of nanoseconds, which ismuch faster than the dynamics time of the order of milliseconds. This allows us to apply theequilibrium charge of the dust particle as the charge state [2]. In this study the forces by ion dragforces due to Coulomb scattering and due to ion absorption, gravity, Lorentz force are taken into
account. In order to calculate the dynamics of the dust particle the background plasma andmagnetic field configuration parameters are given by the B2-EIRENE code [3], where the powertransferred from the core plasma is 200 kW and the corresponding density is 7.5 × 1018 m-3. The
plasma density and temperature around the mid-plane of the SOL/diverter are the order of 1018 m-3
and 10 eV, respectively, where parallel flow speed is around 10 km/sec.
So far, mechanisms and positions of dust generation are not clarified. Therefore, in this study it isinvestigated the dynamics of dust particles released from the strike points of the separatrix on thedivertor plates made of copper. The dust particles are ejected from the strike point to the poloidally
isotropic directions. Results indicate that the dust dynamics released from the inner divertor plate
and the outer one is quite similar. For example, the 90 % of released dust particles of the radius 3
μm with higher speed than 10 m/sec are redeposited on the first wall. The rest are redeposited on
the dome region, where the release direction is almost parallel to the divertor plate. On the other
hand the 20 % of the slower dust particles with 1 m/sec are returned to the divertor plates due tothe friction force of the poloidal plasma flow toward the divertor plate, which is the order of a few
km/sec.
The effects of the dust particles released from the other plasma-facing components such as theentrance of the closed divertor to the core plasma are discussed. These results will be compared tothe experimental observations in HL-2A, which might lead to the modelling of the dust dynamics
to be modified.
References[1] Y. Liu, J. C. Yan, C. P. Zhou, et al., Nuclear Fusion, 44 (2004) 372.
[2] Y. Tomita, G. Kawamura, Y.D. Pan and L.W. Yan, Plasma Science and Technology, 13(2011) 11.
[3] Y. D. Pan and R. Schneider, J. Nucl. Mater. 363-365 (2007) 407.
This work is partly supported by the Nuclear Fusion, a Grant-in-Aid for scientific research from Ministry of
Education, Science and Culture of Japan (No. 19055005) and the National Magnetic Confinement Fusion
Science Program of China (No. 2009GB106006).

60

P1-23
Deuterium retention and desorption behavior of co-deposited carbon film produced bydeuterium arc discharge with carbon electrodes
Y. Nobuta, J. Kanazawa, Y. Yamauchi, T. Hino, K. Yokoyama1), S. Suzuki1), K. Ezato1),
M. Enoeda1), M. Akiba1)
Hokkaido University, Sapporo 060-8628, Japan
1)Japan Atomic Energy Agency, Ibaraki 311-0193, Japan
Evaluation of tritium retention in the vacuum vessel is an urgent issue for ITER and next step
fusion devices, since it greatly effects on safety and limits the operational schedule. Graphite
material, such as carbon fiber composite (CFC), is candidate for divertor target. The carbon
material easily erodes by fuel hydrogen and the eroded carbon particle co-deposits on the wallswith fuel hydrogen. It is known that the fuel hydrogen retention of the co-deposited carbon film issignificantly high. Thus, it is important to investigate the fuel hydrogen retention of the codeposited carbon film, in order to evaluate the in-vessel tritium inventory.
In fusion devices, the co-deposited carbon film is produced not only on plasma-facing surface but
also at the gaps between armor tiles, and the removal of the fuel hydrogen in the gaps is not easy.
In the present study, the co-deposited carbon film was prepared on the gaps between the tiles, inthe deuterium arc discharge with carbon electrodes [1]. The deuterium retention and desorption
behavior of the co-deposited carbon film was examined. The deuterium concentration was
obtained as functions of discharge pressure and gap position. The relation between the deuterium
desorption behavior and microstructure of the co-deposited film is also discussed.
The deuterium arc discharge was conducted with discharge current of 50A and voltage of 12-16V.
The sample holder with gap width of 8-50 mm and gap depth of 100 mm was placed in the
discharge chamber. The discharge pressure varied from 0.8 to 36 Pa. During the discharge, boththe carbon atom and the deuterium deposited on Mo substrates placed on the wall of the gap. The
substrate temperature measured by using thermocouples during the discharge was up to 330 K.
After the discharge, deuterium concentration and desorption behavior of the co-deposited carbon
films were investigated with thermal desorption spectroscopy (TDS). The crystal structure was
examined by Raman spectroscopy.
The deuterium concentration in atomic ratio of D/C increased with the discharge pressure. In thecase of 36 Pa, the D/C reached 0.9. The deuterium retained in the film desorbed in the forms of
D2, HD, CD4 and C2D4. The desorption behavior of D2 depended on the degree of deuterium
concentration. In the film with a low deuterium concentration (D/C = 0.1), a single desorption
peak appeared at around 1000 K, while in the film with a high deuterium concentration (D/C =0.9), three desorption peaks appeared at around 670, 850 and 1000 K. The crystal structure with ahigh deuterium concentration was amorphous. The film with D/C of 0.9 partly contained polymerlike structure. The gas species desorbed also depended on the deuterium concentration. In the filmwith a low deuterium concentration, major gas species were HD and D2 while in the film with ahigh deuterium concentration, the desorption amount of CD4 and C2D4 increased.
The film thickness decreased exponentially with increase of the gap depth. In the case of narrowgap width, the film thickness rapidly decreased with increase of the gap depth. The clear relation
between the deuterium concentration and the gap depth was not observed in the present codeposited carbon films.
The results obtained in this study offers important data for the evaluation of in-vessel tritium
inventory.
References[1] T. Hino et al, Nuclear Fusion, 45 (2005)894.

61

P1-24
Deuterium retention in graphite and its removal by inert gas glow discharge
A. Matsumoto, Y. Yamauchi, T. Hino, K. Nishimura1), Y. Ueda2)
Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo, 060-8628, Japan
1)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
2)Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871, Japan
Glow discharge cleaning using inert gas has been widely used in order to reduce the fuel hydrogen
retention so far. Carbon fiber composite is used for divertor tiles in ITER because of its low
atomic number and its high fracture toughness to high heat load. However, the removal amount offuel hydrogen due to the glow discharge cleanings in graphite has not been investigated in detail.
In the present study, the amounts of retained or desorbed deuterium for graphite during the glow
discharge [1] were measured and the effect of the inert gas glow discharge on the reduction of thedeuterium retention was investigated.
The graphite sheets of IG-430U were used for a liner in the glow discharge apparatus. In order toimplant the deuterium into the liner, the deuterium glow discharge was conducted between theliner cathode and copper anode. After that, the helium, neon or argon glow discharge wasconducted in order to reduce the deuterium retention. The gas pressure before the discharge was
kept 8 Pa using a mass flow controller. The liner temperature was room temperature. The amountsof retained or desorbed deuterium were measured by residual gas analyzer. The changes in partialpressure of gases containing deuterium were measured by a quadruple mass spectrometer. The
obtained results were compared with the cases of other material liners, 316L stainless steel andtungsten.
The amount of retained deuterium in the graphite by deuterium glow discharge was very large,
compared with that of 316L stainless steel or tungsten. This is owing to high capability of carbon
to trap the deuterium. It was found that the reduction of the deuterium retention by helium glow
discharge was most effective among the inert gas species. The implantation depth of helium ion ishighest among these inert gas ions, so that the impact desorption of helium ion might have welldesorbed the deuterium.
In addition to the reduction of deuterium retention in graphite by inert gas glow discharge, theinert gas retention is also presented.
References[1] T. Hino et al, Fusion Engineering and Design, 85(2010)974-97862

P1-25
Hydrogen Atom Behavior with Structural Change of Carbon Materials

S. Saito1), A. Takayama2), A. M. Ito2), H. Nakamura1, 2)
1) Nagoya University, Nagoya, Aichi 464-8603, Japan
2) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
BCA-based Monte Carlo simulation codes, for example, EDDY, TRIM, and ACAT have been
applied to studies for plasma-surface-interaction on physical sputtering, erosion, and ion
implantation. These codes employ the Monte Carlo method, which implies that the target material
is assumed to have an amorphous structure of randomly placed atoms.
In our previous research [1], we extended BCA-based simulation to any structure of target
materials, including monocrystals, polycrystals, crystals with defects, and amorphous crystals. In
that code, however, atoms are injected into the same initial material for each injection. The
structural change of target materials is cleared prior to the next injection.
Under bombardment of plasmas whose thermal energies are higher than several eV, it is expectedthat the structures of target materials dynamically change. For example, when single crystalline
graphite material is bombarded by hydrogen plasmas, the surface of the graphite becomeshydrogenated amorphous carbon by the chemical and physical reaction of hydrogen atoms. This
structural change affects the processes of retention, reflection and sputtering.
In this study, BCA-based simulation is further extended to handle cumulative structural changes ofmaterials. This extension reveals the following issues:
・Time evolution of the structural change of target materials;
・The effect of structural change to retention, reflection and sputtering yields;
・The effect of retention of atoms in target materials to retention, reflection and sputtering yields.
We perform the extended BCA-based simulation and investigate hydrogen behavior withstructural change of carbon materials.
References[1]A. Takayama, S. Saito, A. M. Ito, T. Kenmotsu, and H. Nakamura, Jpn. J. Appl. Phys, 50(2011) 01AB03.
This work was supported by Grant-in-Aid for JSPS Fellows (No. 23-340).

63

P1-26Time Development of Agglomeration Process on Carbon Dusts in Hydrocarbon Plasma
A. M. Ito, M. Nishiura, T. Koizumi1), H. Nakamura, N, Ashikawa, T. Kenmotsu1), M. Wada1)
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)Doshisha University, Kyotanabe, Kyoto 610-0394, Japan

Carbon dusts in a fusion device are expected to absorb a considerable amount of tritium atoms andincrease tritium inventory in the device. The source of these dusts is carbon made first walls of thefusion device. Carbon walls emit carbon clusters and hydrocarbons by hydrogen plasma-carbon
wall interaction, which diffuse into plasma and agglomerate to form dusts in low temperature edge
plasma. The time development of carbon dust formation in edge plasma should govern thequantity and characteristics of formed carbon clusters, and affect the amount of overall tritium
retention. Thus, the fundamental processes governing the dust growth is important for future
fusion operation. Information on the size and density of carbon dusts in the edge plasma, as well as
the density distributions carbon hydride molecules are indispensable to clarify the dust growth
process in edge plasma. As one promising way to quantify densities of carbon hydrides theinfrared absorption spectroscopy system can be installed to observe edge plasma. A small
experimental setup has been designed and built to investigate that this scheme can be effective to
study dust formation by hydrogen plasma-carbon wall interactions.
The designed experimental system confines hydrogen plasma inside of a graphite container. The
discharge region is 80 mm diameter and 300 mm long. A carbon hollow cathode is installed insideof the all graphite wall container that serves as the anode of the discharge. The graphite madeplasma container, together with graphite made hollow cathode can be taken out of vacuum system
which is a 120 mm diameter and 350 mm long cylindrical glass chamber. Carbon dusts have been
confirmed formed on the graphite wall by running Ar discharge in the present experimental set up.
An optical path is opened on both sides of the discharge container, so that infrared absorption
spectra can be measured. The observed infrared spectra will be compared with vibration energy
levels of large hydrocarbon molecules calculated with an ab initio molecular orbital code Gaussian
03. The particle size distribution of the dust formed in our experimental setup will also becompared with molecular dynamic simulation.
This work has been supported by Collaborative Research Program of National Institute for Fusion Science and
NINS Program for Cross-Disciplinary Study.

64

P1-27
Retention of Hydrogen Isotopes in Tungsten Exposed to ELM Plasma
T. Kenmotsu, T. Ono1), M. Wada
Doshisha University, Kyotanabe, Kyoto 610-0394, Japan
1)Okayama University of Science, Ridai-cho 1-1, Okayama 700-0005, Japan
Tungsten has been selected as a candidate material of the ITER divertor plate because of its highmelting temperature, good thermal conductivity and highly resistive nature against erosion due tohigh threshold energy for sputtering [1]. Divertor plates in a fusion device are exposed to highintensity heat fluxes of energetic particles. Many experiments [2,3] have indicated that tungsten
retains tritium and deuterium under their bombardment of hydrogen isotope plasmas. The resultant
retentions of the isotopes lead to the safety and economic problems, and should be minimized for
future fusion reactor operations. Thus, efforts are being made to find out which factors is the mostinfluential upon hydrogen isotopes retention in tungsten.
We have been studying hydrogen isotope retentions in tungsten utilizing a dynamic Monte Carlosimulation code ACAT-DIFFUSE [4]. So far, a uniform temperature model is assumed to evaluate
the retentions. However, large temperature gradient near the surface of plasma facing component
will largely alter the transport of hydrogen isotopes in tungsten. To realize more accurate
simulation calculation under high heat flux exposures as observed in edge localized mode (ELM)
conditions, the code has been modified to take into account of the diffusion of hydrogen isotopes
by including subroutines of heat transfer calculation based upon the finite element method. Someinitial results of the code will be introduced.
References[1] ITER EDA Documentation Series, No. 7, IAEA (1996).
[2] A.A. Haasz, J.W. Davis, M. Poon, et al., J. Nucl. Mater. 258-263 (1998) 889.
[3] V.Kh. Alimov, K. Ertl, J. Roth, J. Nucl. Mater. 290-293 (2001) 383.
[4] Y. Yamamura, Nucl. Instr. Meth. B 28 (1987) 17.

This work was partly supported by Grand-in-Aid for Scientific Research, MEXT, Japan, Priority area 467,
‘Fusion Tritium’.

65

P1-28Deuterium Trapping by Radiation Defects in Tungsten
Y. Hatano, V. Kh. Alimov
University of Toayam, Toyama 930-8555, Japan
Tungsten is recognized as a primary candidate of plasma-facing material. Radiation defects suchas vacancies and voids can act as trapping sites against hydrogen isotopes, and hence tritium
accumulation in neutron-irradiated tungsten has to be carefully evaluated. A common way toinvestigate the trapping effects of radiation defects is to damage tungsten by heavy ions andmeasure hydrogen isotope retention after exposure to high flux plasma in divertor simulators(linear plasma machines). Heavy ions, however, can create defects only in near-surface regions,and hence defects induced by irradiation of heavy ions may be modified by exposure to high flux
plasma. If such modification by plasma takes place, the effects of neutron irradiation in the bulk oftungsten cannot be simulated well by implantation of heavy ions. In the present study, deuterium
was introduced in tungsten damaged with heavy ions by exposure to gas or low energy neutral
atoms at moderate flux, and trapping of deuterium was examined by thermal desorption technique.
The obtained results were carefully compared with those provided by plasma exposure tounderstand the effects of damage modification by plasma.
This study has been supported in part by Grants-in-Aid for Scientific Research on Priority Areas, 476, Tritium
for Fusion, from MEXT, Japan, and Japan-US Joint Research Project TITAN.

66

P1-29Simulation of Deuterium Retention in Tungsten Exposed to Divertor Plasmas
K. Ohya, Y. Soneda
Institute of Technology and Science, The University of Tokushima, Tokushima 770-8506, Japan
An important issue in the choice of plasma-facing materials for ITER is the tritium retention,
which must be reduced for safety reasons. Dominant retention mechanism for tungsten (W) is
implantation. Implanted T atoms are highly mobile at elevated temperatures and they are retainedin radiation damage sites or defects of the bulk. In the previous study, models for diffusion,
including trapping and detrapping, of thermalized atoms in the bulk and the surface recombination
were incorporated into a binary collision approximation code, EDDY.
In the present study, deuterium (D) retention characteristics for the inner and outer divertor plates,
made of W, are discussed in a detached plasma condition of ITER. The edge plasma parameters
used for simulation are calculated with B2-EIRENE. The magnetic field lines intersect the plate atvery shallow angles between 1 and 3 degs, where the gyro-motion and electric field influence theangle of incidence of the impinging ions. Therefore, the mean angle of the ions at each position of
the plate is calculated using particle-in-cell plasma simulations for different magnetic angles. The
mean energies of D ions exposed to the plate are tens of eV or less. At such low energies, thereflection coefficient at the relevant energy and angle ranges to the divertor plates is calculated bya molecular dynamics code, which reveals that most ions impinging the plate are reflected andthus much less implantation of the ions. Therefore, the incident fluxes to the plates are reduced
according to the reflection coefficient.
At the inner divertor W plate, at the early stage of the discharge, the number of mobile D atoms
reached a steady-state value due to a balance between implanted and re-emitted atoms. During thedischarge, most of implanted D atoms were retained in traps and after saturating available traps,
inward diffusion and subsequent trapping increase the inventory. Most of trapped atoms remainedin the bulk even after the discharge, whereas the number of mobile atoms is steeply reduced.
Further detailed calculation and discussion on surface processes at such low energy will be an
important issue for the use of W as a divertor plate in ITER and devices beyond.

67

P1-30
Simulation Study of Tungsten Exposed to Helium / Neon / Argon Based on Binary Collision
Approximation
A.Takayama1), S.Saito2), A.M.Ito1), N.Ohno2), H.Nakamura1,2)
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)Nagoya University, Nagoya, Aichi 464-8603, Japan
It has been experimentally observed that nanostructured tungsten is formed due to exposure ofhelium plasma [1], and that the fomation of nanostructure differs from irradiation nuclear species,
say, helium, neon, or argon.
We applied a binary collision approximation (BCA) based simulation [2] to the noble gas plasma
exposure to tungsten material and found that the BCA model could qualitatively explain
differences of nanostructure formation between irradiation nuclear species.
In this presentation, we report simulation results based on the BCA model of tungsten exposed tohelium, neon, and argon plasma, and refer to comparisons between simulation and experimental
results.
References[1] D. Nishijima, M. Y. Ye, N. Ohno, and S. Takamura, J. Nucl. Mater., 313-316 (2003) 97;
S. Kajita, W. Sakaguchi, N. Ohno, N. Yoshida, T. Saeki, Nucl. Fusion, 49 (2009) 095005.
[2] A. Takayama, S. Saito, A. M. Ito, T. Kenmotsu, and H. Nakamura, Jpn. J. Appl. Phys., 50(2011) 01AB03.
This research is supported by the National Institute for Fusion Science Collaborative Research Programs
NIFS11KNTS008, the National Institutes for Natural Sciences (NINS) Program for Cross-Disciplinary Study,
and a Grant-in-Aid for Scienctific Research (No. 19055005) from the Ministry of Education, Culture, Sports,
Science and Technology, Japan.

68P1-31
Cross-Field Motion of Plasma Blob-filaments and Related Particle Flux in an Open
Magnetic Field Line Configuration on QUEST
H.Q. Liu, K. Hanada1), N. Nishino2), R.Ogata, M.Ishiguro, H. Zushi1), K. Nakamura1),
A. Fujisawa1), H. Idei1), M. Hasegawa1), and QUEST group

IGSES, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan
1)RIAM, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan2)Department of Mechanical System Engineering, Graduate School of Engineering, Hiroshima
University, Japan
Transport of particles and heat by collective motions at the boundary of magnetically confinedplasmas is an important problem for steady state operation in controlled thermonuclear fusion
research [1]. Recently experimental and theoretical work suggested the importance of ‘blob’ typefluctuation in plasma edge, scrape-off layer (SOL). Blob causes huge amount of energy andparticle losses [2]. Blob-filaments have been observed by combined measurement with a fastcamera and a movable Langmuir probe in an open magnetic field line configuration of electron
cyclotron resonance (ECR) heating plasma in QUEST [3]. The experimental observation show thatthe intermittent visible blob-filaments extended along field lines do correspond to over-dense
plasma structures and propagated across the field lines to the outer wall. The typical radial velocityof the blob-like structure, Vb, was measured as ~ 1 km/s at the intermediate area and increased
along major radius, R (outward). The cross field movement of blob was dominantly driven by the
E × B force. The radial velocities of blob-like structures in QUEST corresponds to roughly 0.02 ~0.07 of the local sound speed (Cs). Vb/Cs decreases with local plasma density and plasma
temperature.
The large blob structures, occurring only 10% of the time, can carry more than 60% loss of theentire radial particle flux. This indicates blob structures will cause the dominant losses channel in
the SOL regions. It strongly interacts with the first wall, and thus the first wall might be damaged.
The particles of main plasma loss by blob radial motion and thus the particle confinement time
will be affected by blob losses. The radial velocity, size of the blob and radial flux driven by blob
showed good agreements with the results obtained by sheath-connected interchange theoreticalmodel. In most of big tokamak devices, ITER, JET, DIII-D etc., the blobs closure schemes are alsosatisfied with sheath-connected interchange model. A roughly prediction for the blob regime of
ITER can be given according to the QUEST and other tokamak device's results.
References[1] P.C. Stangeby Phys. Plasmas 9 (2002) 3489
[2] D. A. D’Ippolito et al., Phys. Plasmas 18 (2011) 060501[3] H. Q. Liu et al., J. Nucl. Mater. (In Press, Available online 29 December 2010)
This work was performed under the auspices of the NIFS Collaboration Research Program (NIFS08KUTR024).

69

P1-32Analysis of end-loss-ion-flux for application studies of the plasma flow from the end-mirror
exit of GAMMA 10.
K. Ichimura, Y. Nakashima, K. Hosoi, T. Hisato, T. Ishii, H. Ueda, S. Kigure, S. Takahashi,
S. Hotaka, M. Ichimura R. Ikezoe, M. Yoshikawa, M. Sakamoto, T. Imai
Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577, Japan    To perform a good simulation experiment for the divertor system of the actual fusion reactor,
the test plasma should achieve very high energy and density. As one of the device which iscapable of generating such high-performance plasma, the world's largest tandem mirror device,
GAMMA 10 has been noted. Since the mirror-magnetic field is open-ended, there exists an axial
loss flux of the plasma which is called the end-loss flux witch is quite useful for a diverter
simulation experiment. More detailed study of the end-loss flux is needed before we proceed to thedivertor simulation experiment. The end-loss ion energy analyzer (ELIEA) is used and the energy
distribution and the current density of the ion flux are examined in the west end-cell of GAMMA
10. The objective of this study is to obtain the knowledge of characteristics and controllability ofparameters in the end-loss ion flux.
   GAMMA 10 is an axisymmetric tandem mirror which consists of a central cell, two anchors andplug/barrier cells. Parameters of the loss plasma will be changed due to the change of parametersof the main plasma. In the GAMMA 10 experiments, the power of ICRF heating input and theamount of gas-puffing particle supplied to the main plasma are varied and the change of end-loss
ion flux is investigated. In a typical GAMMA 10 plasma (without confining potential), the ion
current density is measured to be about 0.4mA/cm2. It is found that the least square method fitting
to the measured energy distribution of the end-loss ion flux works better if we assume that theenergy distribution to be double-component Maxwellian rather than single-component
Maxwellian. Here, we defined the effective temperature of ions from the ratio of two components
with respect to the total ion flux. The evaluated effective ion temperature is about 400eV in typicalplasma of GAMMA 10. The particle flux of the end-loss ion can be increased by increasing theamount of gas-puffing. However, injecting too much gas causes the degradation of the energy orcollapsing of the plasma. Therefore, in order to increase the particle flux to the west end-cellwithout increasing the amount of gas, the additional heating in the east anchor-cell using a newly
installed ICRF antenna is proposed and tested. With an input of the new-ICRF in the east anchorcell, significant increase of the ion-particle flux was ofserved. In the paper, detailed results will be
described and discussed.

70

P1-33
Material dependence on plasma shielding induced by laser ablation
Takuya Kono, Y.Hirooka1), K.Kikuyama, S.Misaki, T.Oishi, T.Yabuuchi and K.A.Tanaka
Graduate School of Eng., Osaka University, Suita, Osaka 565-0871, Japan
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
In nuclear fusion research, it is one of the important themes to study the damage on plasma facing
components (PFC). For example, expected heat load are 10 to 100 MW/m2 at magnetic fusionenergy (MFE) divertor and 109 W/cm2 or higher at inertial fusion energy (IFE) first walls at
plasma densities 1019 /m3 or higher and temperature 1 eV to keV or even to MeV. With the plasma
thermal loading, any solid material may be ablated, therefore, some way to protect the reactor wallis necessary. However there are some active functions to protect themselves like vapor shielding[1] and plasma shielding effects. Shedding light on this effect can be of use.
Plasma shielding is studied using the experimental set up “LEAF-CAP” (Laboratory Experiments
on Aerosol Formation by Colliding Ablation Plumes) [2]. In this set up, the ablated Tungsten or
Molybdenum plumes by a third harmonic beam of YAG laser (6ns, 10Hz) can be crossed
orthogonally. This laser ablation scheme can cover pretty wide range of intensity regime as a heatsource at its laser focus spot from 103 to 1014 W/cm2 and the ablated plasma plume temperatureand densities can range from 0.1 to a few eV from 1011 to 1013 /cm3.
Plasma plume will give a material deposition when the quartz thickness monitor is placed in theforward direction. When another plasma plume crosses this plasma plume, the original plume may
collide at the crossing point and may be absorbed through the collision. If any energy of the plumeis absorbed, the plasma is considered to be shielded. Shielding rate is reduced from the deposition
rates with or without the shielding plasma plume.
As a result when shielding Tungsten plume intersects the original Tungsten plume, the deposition
rate decreases from 5.2×1013 to 3.7×1013 particles/pulse/cm2 at the laser energy density 10 J/cm2.
About 30% Tungsten are intercepted by the intersecting plume. In the case of Molybdenum, thereduction of deposition rates is similar, and shielding rate is 20%. In the ICCD camera fast movie
observation shows that there is clearly a collision of two plasma plumes. In case of Carbon, thecollided part stagnates for more than microsecond.
The dependence of deposition of Molybdenum on laser energy density is higher than that ofTungsten. However, the dependence of shielding rate is regardless of laser energy density, 20 to
30%. The dependence of the material considering for the PFCs is measured by comparing valuesof quartz thickness monitor.
These experimental data indicate the possibility of plasma shielding effects to heat load flux at thedivertor or first wall.
References[1] J.G. van der Laan, M. Akiba, A. Hassanein, M. Seki and V. Tanchuk, Fusion Engineeringand Design 18 (1991) 135-144; A. M. Hassanein, G. L. Kulcinski and W. G. Wolfer, Nuclear
Engineering and Design/Fusion 1 (1984) 307-324[2] Y. Hirooka, K. A. Tanaka, H. Sato, K. Ishihara and A. Sunahara, Journal of Physics:
Conference Series 244 (2010) 032033

71

P1-34
Fundamental study on hydrogen retention in carbon aerosol relevant to inertial fusion
reactor
Tatsuya OISHI, Y. HIROOKA1), N. OHMOTO. K.A. TANAKA
Graduate School of Eng., Osaka University, Suita, Osaka 565-0871, Japan
1)National Institute for Fusion Science, Toki, Gifu 506-5292, Japan【Introduction】Chamber wall of an IFE reactor is subject to the bombardment of high-energy, short-pulse X-ray, unburned DT and He particles from implosions. Therefore solid wall could be
ablated leading to the formation of aerosol in the target chamber. If aerosol is formed in thehydrogen atmosphere including tritium and attached to chamber wall, this affects its radiation
safety. Therefore, it is extremely important to understand the carbon aerosol behavior of retaininghydrogen using a laboratory-scale experimental setup.
【Experimental】 In our work, a YAG laser beam (1064nm, 6nm, 10Hz) is first converted into the
3rd harmonic (355nm), and it’s half-split into two beam and line focused to two arc-shaped carbon
targets positioned in vacuum <10-3 Pa. Therefore ablation plasma plumes collide with each otherand create aerosol in the center-of arc region. The laser energy density is 10J/cm2/pulse relevant to
the IFE reactor situation.[1] The chamber is filled with hydrogen gas as the substitution of tritium.
Aerosol is collected in different hydrogen pressure condition 10, 20, 50Pa, and its hydrogen
retention is examined by TDS.
【Results and discussion】Examining carbon aerosol created in non-hydrogen atmosphere by Qmass, C2+, C3+, C4+, C5+, C6+ cluster were detected. As the laser energy density was increased, lownumber were clusters decreased, and high number ones were increased. The result indicates low
number clusters were used for composing high number molecules through the collision processes.
In fact, collecting carbon aerosol created in laser energy density 2.2J/cm2/pulse and observing its
surface and structure by digital microscopy, SEM and TEM, CNTs were clearly found in thecollected samples. It indicates that the combination reaction of cluster occured because CNT canbe made by C5, C6. Colliding carbon plume experiments have also been conducted in the LEAFCAP facility with a backfilled hydrogen gas. Under the conditions which already explained“Experimental”, carbon aerosol have been collected and examined its hydrogen retention by TDS.
Expressing result as atomic ratio of H/C, it is proportional to hydrogen gas pressure, and indicatesmaximum H/C≈0.28 at the sample collected in gas pressure is 50Pa. This is still low compared tothe saturated co-deposition ratio of H/C~0.4.[2] However our result indicates a possibility to retain
fair amount of tritium in IFE chamber.
References
1. T. J. RENK et al., Fusion Eng. & Design 65 (2003)399-406.
2. D. POST and R. BEHRISCH(Eds.), NATO AISI Ser. Plenum Press, New York (1986).

72

P1-35
Optical emission characteristics of an atmospheric Ar/N2 microwave plasma jet in relation
to superhydrophilic surface treatment of stainless steel
Tracy Estella Tuballa 1) , Ma. Camille Lacdan 1) , Julie Anne Ting 1) , Henry Lee, Jr. 1,2) ,Leo Mendel Rosario 1,3) , Henry Ramos 1) and Roy Tumlos 1,4)
1) Plasma Physics Laboratory, National Institute of Physics, University of the Philippines,
Diliman, Quezon City 1101, Philippines
2) IBF Electronic GmbH & Co. KG, Dr. Robert MurjahnStrasse, 12 D-64372 Ober-Ramstadt,
Germany
3) Sciences Department, College of Arts, Science, and Education, FEATI University, Manila 1003,
Philippines
4) Department of Physical Sciences and Mathematics, University of the Philippines, Manila 1000,
Philippines
An atmospheric-pressure microwave plasma jet of Ar/N2 was used to enhance the wettability ofstainless steel samples. The contact angles of the samples were investigated in terms of varying
gas ratios of Ar and N2 in the plasma. Optical emission spectroscopy was used to determine andcompare the relative populations of the active species in the Ar/N2 plasmas. The observed
wettability changes were related to the contact angle measurements.The morphology, elemental
composition, and roughness of the treated sample surfaces were examined using SEM, EDX, andAFM, respectively.
References[1]. M.C. Kim, D.K. Song, H.S. Shin, S.H. Baeg, G.S. Kim, J.H. Boo, J.G. Han, and S.H.
Yang, Surface & Coatings Technology Volume 171, pp. 312-316, 2003.
[2]. A. Schutze, J.Y. Jeong, S.E. Babayan, J. Park, G.S. Sewyn, and R.F. Hicks, IEEE
Transactions on Plasma Science Volume 26, p. 1685, 1998.
[3]. S. Tang, N. Lu, S.W. Myung, and H.S. Choi, Surface & Coatings Technology Volume
200, pp. 5220-5228, 2006.
[4]. J.W. Lin and H.C. Chang, Nuclear Instruments and Methods in Physics Research B
Volume 269, pp. 1801-1808, 2011.
[5]. T. Yuji, S. Fujii, N. Mungkung, H. Akatsuka , IEEE Transactions on Plasma Science
Volume 37 No. 6, pp. 839-845.
The authors would like to thank Mr. Reinhard Fischer of IBF Electronic GmBH& Co. KG. and DOSTPCIEERD for providing the financial support used in this study.

73

P1-36
Proposal of New Type Diplexer for ECCD System
M. Saigusa, S. Sugawara, K. Atsumi, Y. Oda1), T. Yamaguchi, and K. Sakamoto1)
Ibaraki University, Hitachi, Ibaraki 316-8511, Japan
1)Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan
Neoclassical tearing modes (NTM) can be controlled by the local current drive in a magnetic
island with electron cyclotron current drive (ECCD) [1]. For improving a stabilizing efficiency ofNTM, the fast directional switch (FADIS) had been developed [2, 3].
In this paper, the new type diplexer as a fast switching device of high power millimetre wave is
proposed for ECCD system in fusion devices. The principle is a ring resonator of oversizedcorrugated circular waveguide. The two half mirrors (dielectric disks or slotted antennae) areintegrated in a ring resonator. The switching operation of a diplexer has been simulated with thecode using finite differential time domain (FDTD) method. The switching operation of thediplexer by frequency shifting has been confirmed in the cases of dielectric disks or slotted
antennae, where the loss caused by mode conversion and misalignment in a resonant ring are
considered. A diplexer for 1 MW short pulse at a frequency of 170 GHz is designed andfabricated.
References[1]H. Zohm, G. Gantenbein, G. Giruzzi, S. Gunter, F. Leuterer, M. Maraschek, et al., Nucl.
Fusion 39 (5) (1999) pp. 577-580[2]M.I. Petelin, “Quasi-optics in high-power millimeter-wave systems,” AIP Conf. Proc 691(2003) pp. 251-262[3]W. Kasparek, M. Petelin, V. Erckmann, D. Shchegolkov, A. Bruschi, S. Cirant, et al.,Fusion Sci. Technol. 52 (2007) pp.281-290
This work has been supported by the Grants-in-Aid for Scientific Research (c) of 22560818.

74

P1-37ECCD Experiment Using an Upgraded ECH System on LHD
Y. Yoshimura, S. Kubo, T. Shimozuma, H. Igami, H. Takahashi, M. Nishiura, S. Ogasawara,
R. Makino, K. Ida, M. Yoshinuma, S. Sakakibara, K. Tanaka, K. Narihara, T. Mutoh, H. Yamada,K. Nagasaki1), N. B. Marushchenko2), Y. Turkin2)
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
1) Institute of Advanced Energy, Kyoto University, Uji 611-0011, Japan2) Max-Planck-Institute for Plasma Physics, Greifswald, Germany
Electron cyclotron current drive (ECCD) is an attractive tool for controlling plasmas. ECCD cancontrol the plasma current and rotational transform profiles, which affect magnetohydrodynamic(MHD) activity. In tokamak-type plasma confinement devices, the effectiveness of ECCD for
stabilizing the neoclassical tearing mode has been demonstrated. Also, for stellarators/heliotrons
that do not need plasma current for plasma confinement, the current profile control capability
enables fine plasma control. In the large helical device (LHD), ECCD experiments have been
performed by using EC-waves with the frequency of 84 GHz. The capability of long-pulse
operation of the 84 GHz gyrotron and the wide setting range of EC-wave beam direction enabled
by a 2-D widely tiltable mirror antenna at a bottom port were suitable feature for ECCD. However,
the available injection power for the long-pulse operation was low, and the maximum driven
current was ~10 kA with 100 kW and 8 s power injection. These years, an upgrade of ECH system
on LHD is ongoing. A high-power, long-pulse gyrotron was newly installed. The oscillation
frequency, the output power for several seconds and the output power for continuous operation are
77 GHz, 1.5 MW and 0.3 MW, respectively. The mirror antenna system at a horizontal port used
for the 77 GHz gyrotron also furnishes widely tiltable plane mirror so that the 77 GHz ECH
system can be used for ECCD experiments with high-power and long-pulse. An ECCD experimentwith 775 kW injection power was performed in the second harmonic condition, that is, themagnetic field on the magnetic axis was 1.375 T. The discharges were started with 300 ms pulsesof 84 GHz and 82.7 GHz EC-waves, and the 77 GHz waves of 8 s pulse duration sustained theplasmas. The EC-wave beam direction was scanned toroidally, keeping the beam direction aiming
at the magnetic axis in X-mode polarization. In spite of the change in the EC-wave beam direction,
the electron density and the plasma stored energy were kept nearly the same values for thedischarges, ~0.3×1019 m–3 and ~30 kJ, except for the plasma current. The plasma current showeda systematic change with the change in the beam direction, and at an optimum direction the plasma
current reached its maximum, ~40 kA.

75

P1-38
Improvement of output power and electric efficiency for 77 GHz gyrotrons by externallycontrolled anode voltage in LHD
H.Takahashi, T.Shimozuma, S.Ito, S.Kubo, Y.Yoshimura, H.Igami, M.Nishiura, S.Kobayashi,
Y.Mizuno, K.Okada, Y.Takita, T.Mutoh, S.Ogasawara1), R.Minami2), T.Kariya2) and T.Imai2)
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)Department of Energy and Technology, Nagoya University, Nagoya, Aichi 464-8463, Japan
2)Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577, Japan
In the Large Helical Device, three 77 GHz gyrotrons with each output power of over 1 MW are
operational for plasma generation and heating [1]. The MIG type of the 77 GHz gyrotrons is triode
and the anode voltage (VA) can be controlled using the preset waveform. High power and highefficiency operation was successfully achieved such as the output power of 1.41 MW with theelectric efficiency of 49.1 % for 1 sec by applying the two-step-rise VA. In the first step period,
enough small VA was applied for adequate duration (> 50 ms), in which the electron beam flows
inside the tube but RF was not generated. After that, VA was increased to optimum value foroscillation startup.
The charge neutralization of the gyrotron electron beam in the cavity [2] is considered to be a key
for the improvement of the electric efficiency by the stepwise VA applying. In order to examine theeffect of the beam charge neutralization on the gyrotron oscillation, the gyrotron output
characteristics of the stepwise VA operation were investigated and compared with those of thenormal operation. In the stepwise VA operation, the output power increased with the extension of
the first step duration and was saturated for the longer duration than 50 ms. This indicates that thedrop of the accelerating voltage recovered through the charge neutralization process in the firststep period and the space charge was enough neutralized for ~50 ms. In the normal operation withthe rectangular VA waveform, it was found that 5~6 kV higher acceleration voltage was required togenerate the similar output power with the stepwise VA operation with the first step duration of
100 ms. The increment of the beam acceleration voltage in the normal operation case corresponds
approximately to the drop of the acceleration voltage due to the space charge effect. These results
suggest that the gyrotron operational parameters are able to be optimized for the fully accelerated
electron beam by applying the stepwise VA with the adequate duration of the first step, leading to
the improvement of the output power and the electric efficiency.
References[1]H. Takahashi, et al., Fusion Sci. Technol. 57 (2010) 19.
[2]G. S. Nusinovich et al., Phys. Plasmas 13 (2006) 083106.

76

P1-39Experimental results of electron Bernstein wave heating on the LHD and consideration for ahelical type fusion device
H. Igami, S. Kubo, T. Shimozuma, Y. Yoshimura, H. Takahashi, H. Idei1), M. Nishiura,
S. Ogasawara, R. Makino, R. Kumazawa, T. Mutoh, A. Sagara, T. Goto
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
1) Research Institute for Applied Mechanics, Kyushu Univ. Kasuga, 816-8580, JapanAs a heating method in fusion devices, electron cyclotron heating (ECH) has the great advantagethat the launcher of the electron cyclotron (EC) wave can be placed apart from the plasma. In an
envisioned high density operation in a helical type fusion reactor, the EC wave cannot access theelectron cyclotron resonance (ECR) layer because it is shielded by the cutoffs. ECH by theelectron Bernstein wave (EBW) is expected in such a high density operation, because the EBW
has no density limit in propagation and is absorbed at the ECR layer by cyclotron damping.
In the large helical device (LHD), two methods of EBW excitation and heating have been studied
by launchers placed in ports apart from the plasma. One is to inject the extraordinary (X-) mode
extremely obliquely from the high field side [1]. This method is applicable even the electron
density is less than the cutoff density. In the plasma where ne_bar= 5 x 1019m-3, when the 84GHz,
0.43MW EC wave was launched, increase of the stored energy (Wp) and the ECE signals wereobserved. ( ne_bar : line averaged electron density.) The result of the ray tracing calculationsuggests 10~30% of the launched power was absorbed as the EBW and the rest is absorbed as the
X-mode. Almost all of the launched power can be absorbed by this method. In the helical typefusion device, this method may be applicable in its high density operation if the electron density isless than two times of the cutoff density. Another is to inject the O-mode from the low field side
obliquely to the external magnetic field [2]. Precise adjustment of the launching direction isrequired whereas this method has no density limit. Increase of Wp and the second X-mode ECE
signals were observed when the 77GHz, 1MW EC wave was launched into a plasma where ne_bar =9 x 19m-3. However the estimated heating efficiencies were about 15%. Improvement of theheating efficiency including changing the EC wave frequency and position of the launcher is now
under consideration. Application of these methods to a helical type fusion device is also underconsideration with reference to these experimental results.
References[1] H. Igami et al., Nuclear Fusion, 49, 115505 (2009)[2] H. P. Laqua et al., Physical Review Letters, 78, 3467-3470 (1997)77

P1-40ICRF heating experiment using phasing antenna in LHD
T. Seki, T. Mutoh, R. Kumazawa, K. Saito, H. Kasahara, G. Nomura,
Y.P. Zhao1), S.J. Wang2) and LHD Experiment Group
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
1)Institute of Plasma Physics, Chinese Academy of Science, Hefei,, Anhui, 230031, P.R. China
2)Korea Atomic Energy Research Institute, Daejeon 305-600, Korea Rep.
Ion cyclotron range of frequencies (ICRF) heating is one of the heating methods for fusion reactor
and has been carried out successfully in large helical device (LHD). In LHD, ICRF heating has
been used for long pulse plasma discharge and the pulse length of more than 54 minutes wasachieved so far [1]. Ion heating was observed in minority ion heating scheme and high-energy ion
tail more than 2 MeV was detected. During the long pulse discharge by the ICRF heating, thedivertor temperature near the antenna location increased especially and the plasma was terminatedby influx of metallic impurity into the plasma. It is important for the steady state operation by the
ICRF heating to reduce the local heat load near the antenna and edge interaction caused by RF
waves.
We designed [2] and installed new type of ICRF antenna in LHD in order to control the wave
number in the direction of the magnetic field line. The antenna has two straps side-by-side and canchange the parallel wave number by phasing of the strap current. We call the new antenna as HASantenna named after the antenna shape and Hasu-Seigyo (wave number control) in Japanese.
HAS antenna has an improved design by experience of the experimental results of existingpoloidal array antenna. There is a possibility of control of sheath potential at plasma edge region
generated by antenna electric field by phasing of antenna. It is expected to suppress theunnecessary acceleration of edge particles and reduce the impurity generation caused by RF
injection. Carbon side protectors were replaced by CFC material. The shape of current strap andFaraday shield and water-cooling channel were re-designed in consideration of the steady state
operation.
In the experiment, the phasing effect was studied in the minority heating experiment. The heating
efficiency of 0-pai phasing was higher than that of 0-0 phasing. Achieved electron density was
higher in 0-pai phasing in the same heating power. These experimental results are better than thatof the poloidal array antenna. We will conduct the high power and long pulse discharge andevaluate the antenna performance in steady state operation.
References[1]T. Mutoh, et al., Nucl. Fusion. 47 (2007) 1250.
[2]H. Kasahara, et al., J. Plasma Fusion Res. 5, S2090 (2010).

78

P1-41
Study of ICH Heating Power Profile Evaluation Method in LHD Type Plasma
R. Seki, K. Saito, T. Seki, A. Fukuyama1), K.Y. Watanabe, and T. Muto
National Institute for Fusion Science, Toki 509-5292, Japan
1) Graduate School of Engineering Kyoto University, Kyoto 606-8501, Japan
In the LHD, long plasma discharges with a temperature of several keV are maintained for morethan one hour mainly by ICH. In these plasmas, a total plasma heating power by the ICH isevaluated from a loading resistance of ICH antennas. With the heating power, global confinement
properties are investigated, and it is found that a global energy confinement performance in the
ICH plasma is almost same with that in the ECH and NBI plasmas. However, the ICH heating
power profile and the local thermal transport property in the LHD has not been clarified yet. To
establish the method for evaluating the ICH heating power profile and the local thermal transport
property is an important issue for development of the long time plasma discharges with the higher
confinement capability, especially in the comparatively high density plasma.
In order to evaluate the ICH heating power profile in the LHD experiments, a full wave fieldsolver, TASK3D/WM, on the basis of the Maxwell’s equation is being developed [1]. On the other
hand, in a LHD type fusion reactor, which is about three times larger than the LHD, the ICH
heating power profile was evaluated by a simple ICH heating power evaluation code based on the
Ray-tracing method [2].
As a pre-stage for establishing the ICH heating power profile evaluation code in the LHD, the ICH
heating power profiles in the LHD type plasmas, of which sizes are larger than that of the LHD,
taking the finite beta effects on MHD equilibrium into account are calculated by using theTASK3D/WM and the Ray-tracing code. In this presentation, the validity of the Ray-tracing codeby changing the device size is shown. The effect of the plasma kinetic effects on the propagation
and absorption of the ICRF wave will be discussed.
References[1]A. Fukuyama and T. Akutsu Proc. 19th Int. Conf. Fusion energy 2002, THP3-14.
[2] K. Saito et.al., Plasma Fusion Res. 5 (2010) S1030.

79

P1-42
Improvement of ICRF antenna loading in the minimum-B configuration on GAMMA 10
T.Yokoyama, M.Ichimura, R.Ikezoe, Y.Ugajin, T.Iwai, T.Sato, T.Iimura, Y.Saito, M.Hirata,
Y.Yamaguchi1), A.Fukuyama2), T.Imai
Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki, JAPAN
1)Research Center for Development of Far-Infrared Region, University of Fukui, Fukui, JAPAN
2)Department of Nuclear Engineering, Graduate School of Engineering, Kyoto University, Kyoto,JAPAN
On the GAMMA10 tandem mirror, waves in the Ion-Cyclotron Range of Frequency (ICRF) areused for the plasma production, the ion heating and keeping the magneto-hydrodynamic (MHD)
stability. High-β plasmas in the minimum-B configuration at the anchor cells supply the MHD
stabilizing effect to the whole of GAMMA 10. It is necessary for high performance plasmas thatthe ions are heated efficiently in the anchor cell. In the standard discharge, ICRF waves (around
10MHz), which have resonance layers in the anchor cell, are excited by NAGOYA Type-III
antennas installed in the central cell and propagate to the anchor cell. It is difficult to control only
the heating effects in the anchor cell because these waves are used also for the plasma productionin the central cell.
In this paper, the direct anchor heating by using antennas in the minimum-B configuration is
studied. Firstly, a simple bar-type antenna has been used in the experiment and confirmed thestabilization of GAMMA 10 plasmas [1]. However, the loading of the bar-type antenna has beenfound to be very small values (< 20%). To calculate the loading resistance of antennas in theminimum-B configuration, a three-dimensional full wave numerical code is used, which isdeveloped by one of the authors (A. Fukuyama). To improve the loading, a new antenna has been
designed and replaced in the anchor cell. The loading of the new antenna is compared with that of
the old bar-type antenna in the experiment and calculation. According to calculation, changes ofthe shape from the single strait line to double elliptic arcs and the antenna location are effective forthe improvement. In the initial experiment, the loading of the new antenna is improved up to 50%.
Secondly, the phase control between the anchor antenna and the NAGOYA Type-III antenna in
the central cell has been used and confirmed clear anchor heating effects. For the effective anchor
heating, it is planned to calculate wave excitation in the minimum-B configuration and optimize
the anchor antenna configurations.
References[1] Y.Yamaguchi et al., Fusion Science and Technology, 59, No.1T (2011) 253-255

80

P1-43
Quasi-Optical high purity HE11-mode Exciterfor Oversized Corrugated Waveguide Transmission
M. Sakaguchi, H.Idei, H.Okada1), S.Kobayashi1), S.Kubo1), T.Saito2)
Kyushu University, Kasuga, Fukuoka 816-8580, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)FURUKAWA C&B CO., LTD, Yamato, Kanagawa 242-0018 , Japan
Oversized circular corrugated waveguides will be mainly used for electron cyclotron heating 20
MW transmission lines in the ITER burning fusion-plasma experiments. The higher order modes
as well as a main HE11 mode are excited due to the misalignment of the coupled beam and due tothe manufacturing error of the waveguide components. These cause undesirable events such as
overheating and/or arcing in the transmission lines. The modes excitation is concerned with thetransmission of a high power millimeter wave without remarkable transmission losses and arcing
events. Therefore the aim of the study is to experimentally evaluate of the unwanted mode
excitation. A scalar feed horn antenna for Gaussian beam excitation was designed with anumerical analysis based on the method of moment. A field radiated from the designed antenna
was measured and confirmed to be Gaussian filed distribution. The HE11 mode in the waveguide ismatched to the TEM00 mode (Gaussian beam) in a free space with the conversion efficiency of
98%. The Gaussian beam (TEM00 mode) was often used for the HE11 mode excitation in the
waveguides. The HE11 mode purity over 99% should be prepared to study the loss and the mode
excitation with 1% in the transmission line. The conversion loss from Gaussian beam into theHE11 mode comes from the differences of kurtoses in field distributions.
Two Gaussian beams may be prepared to excite a high purity HE11 mode. The beams should have
opposite phase curvatures but the same size. These were combined at the waveguide aperture withphase matching technique. The excited HE11 mode purity results 99.6% in theoretical analysis. We
designed a quasi-optical system for the HE11 mode exciter system. It is composed of the scalar
feed horn antenna and quasi-optical mirrors. The beam from the antenna was separated with a half
mirror. Each beam was led into ellipsoid reflective mirrors for focusing and expanding. Finally thebeams were integrated with the half mirror again and were shaped with the desired beam profiles
at the aperture. The theoretical analysis and experimental configuration was considered for thehigh purity HE11 mode exciter. In the paper, the system design and experimental results will bereported in detail.
This work was realized in the collaboration between FURUKAWA C&B and Kyushu University.

81P1-44
Finite-Difference Time-Domain Simulation on Millimeter-Wave Propagation in Corrugate
Waveguide
H.Nakamura1,2), K.Ukita3), Y.Tamura3), T.Okamura2), S.Sawada4), S.Fujiwara5), S.Kubo1,2)1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)Nagoya University, Nagoya, Aichi 464-8603, Japan3)Konan University, Kobe, Hyogo 658-8501, Japan4)Shinshu University, Nagano, Nagano 380-8553, Japan
5)Kyoto Institute of Technology, Sakyo-ku, Kyoto 606-8585, Japan
We have developed the Finite-Difference Time-Domain (FDTD) simulation with the DrudeLorentz model [1] to investigate optical phenomena related to the electromagnetic response of
metals using near-field scanning optical microscope (NSOM) [2-5]. As one of our FDTD
simulation results, the oscillating pattern for the transmission image of a gold nanorod using
NSOM was obtained by the FDTD simulation [3].
In the present paper, we propose the FDTD simulation, which has been developed in the nanooptics, to estimate the propagation loss of the millimeter-wave through the corrugate waveguide.
In this proposal, we pay an attention to the following fact: the ratio of the wave length to thetypical structure spatial scale between the nano-optics phenomena and the millimeter waveguide
phenomena becomes the same order.
First we model the corrugate waveguide, which is actually made of Al, as the perfect electric
conductor (PEC). Using the PEC, it is expected to investigate the behavior of the electromagnetic
wave propagation qualitatively. Moreover, as the next step, we also include the Drude-Lorentz
model in the FDTD simulation, to obtain the propagation loss inside material in the waveguide.
References[1]A. Vial, A.-S. Grimault, et al., Phys. Rev. B 71 (2005) 085416.
[2]K. Imura, T. Nagahara. H. Okamoto, J. Chem. Phys., 122 (2005) 077402.
[3]K. Sawada, H. Nakamura, T. Maruoka, Y. Tamura, K. Imura, T. Saiki, H. Okamoto,Plasma and Fusion Research, Vol.5 (2010) S2110.
[4]K. Deguchi, M. Okuda, A. Iwamae, H. Nakamura, K. Sawada, M. Hasuo, J. Phys. Soc.
Jpn., Vol. 78 (2009) 024301.
[5]H. Nakamura, T. Saiki, H. Kambe, K. Sawada, Comp. Phys. Com., Vol. 142 (2001) 464.
This work was supported by NINS Program for Cross-Disciplinary Study.

82

P1-45
Study of magnetic reconnection heating in UTST measured by Doppler spectroscopy
S. Kamio, K. Takemura, Q. H. Cao, H. Itagaki, N. Suzuki, T. G. Watanabe, K. Yamasaki1),
K. Ishiguchi2), R. Imazawa3), T. Yamada, M. Inomoto, Y. Takase, Y. Ono1)
1)The University of Tokyo, 5-1-5, Kashiwanoha, Kashiwa, Chiba, 277-8561, Japan
2)The Open University of Japan, 2-11, Wakaba, Mihama-ku, Chiba, 261-8586, Japan
3)Japan Atomic Energy Agency, 801-1, Mukouyama, Naka, Ibaraki, 311-0193, JapanThe merging start-up of spherical tokamak (ST) plasma has been accomplished in the UTST
device by using external poloidal field coils [1]. The plasma current up to 180 kA has been
attained with assistance of the center solenoid (CS) coil. In this merging start-up method,significant ion heating caused by magnetic reconnection is expected to help the formation of highbeta equilibrium. The reconnection outflow velocity measured by carbon impurity spectroscopy
showed good agreement with the bulk plasma’s Alfven velocity, suggesting the conversion frommagnetic energy to plasma kinetic energy [2].
We have developed a Doppler spectroscopy system using an 8 × 8 channel PhotoMultiplier Tube
(PMT) assembly to observe the temporal evolutions of ion temperature and flow. Line-integrated
emissions from the merged ST plasma were analyzed by using the developed spectroscopic
system. Increase of the bulk ion temperature was observed by helium impurity spectroscopy at thesame time as the plasma merging was completed. The heating efficiency of the plasma merging
will be discussed quantitatively.
References[1] T. Yamada, R. Imazawa, S. Kamio, R. Hihara, K. Abe, M. Sakumura, Q. Cao, T. Oosako,
H. Kobayashi, S. Kiyama, Y. Hirano, M. Inomoto, A. Ejiri, Y. Takase and Y. Ono, Plasma
Fusion Res. 5, S2101 (2010).
[2] S. Kamio, Q. H. Cao, K. abe, M. Sakumura, N. Suzuki, T. G. Watanabe, K. Ishiguchi, R.
Imazawa, T. Yamada, M. Inomoto, Y. Takase and Y. Ono, Plasma Fusion Res. 6, 2402033(2011).

83

P1-46
Effect of Gas Fueling Control on Plasma Performance in Heliotron J
T.Mizuuchi, S.Kobayashi, K.Mukai1), S.Ohshima2), S.Yamamoto, T.Minami, K.Nagasaki,
H.Okada, K.Mizuno1), H.Y.Lee1), L.Zang1), H.Yashiro1), S.Arai1), T.Kagawa1), T.Minami1),
Y.Wada1), H.Watada1), K.Kasajima3), N.Nishino4), Y.Nakashima5), K.Hanatani, Y.Nakamura1),
S.Konoshima, F.Sano
Institute of Advanced Energy, Kyoto Univ., Gokasho, Uji, 611-0011 Japan
1)Graduate School of Energy Science, Kyoto Univ., Gokasho, Uji, Japan
2)Kyoto Univ. Pioneering Research Unit for Next Generation, Gokasho, Uji, Japan
3)Faculty of Engineering, Kyoto Univ., Yoshida-Honmachi, Kyoto, Japan4)Graduate School of Engineering, Hiroshima Univ., Higashi-Hiroshima, Japan
5) Plasma Research Center, Univ. of Tsukuba, Tsukuba, JapanThe optimization of gas-fueling scenario has been studied to improve the plasma performance inHeliotron J, which is a low magnetic-shear device with an L/M = 1/4 helical coil (<R0> = 1.2m,<B0> < 1.5T) based on the helical-axis heliotron concept [1]. This paper discusses the effects ofthe gas-fueling scenario on the plasma performance based on the experimental observations.
A gas fuelling by supersonic molecular beam injection (SMBI) was applied to ECH/NBI plasmas[2]. When the plasma density is increased with conventional gas-puff fueling (GPF), the stored
energy Wp becomes saturated or decreased at a density level, probably due to the edge coolingcaused by excess neutrals. Local fueling with a short pulse of SMBI can increase the core plasma
density avoiding the confinement degradation. In a combination heating condition of ECH and CoNBI, Wp reached about 50% higher value than the maximum one achieved under the similar
heating condition with GPF. On the other hand, it was observed that a large increment of Wp after
turned-off of an intense GPF in a high power NBI plasma [3-4]. By applying an intense GPF, Wp
initially increases with increase of the plasma density and then decreases, as usually observed in acase of excess GPF. If the intense GPF is stopped at a proper timing during the NBI pulse,
however, Wp started to recover and finally reached to a much higher value than the first peak of
Wp. This suggests that the plasma density/temperature profile during the decreasing phase of Wp ischanged to a preferable one for NBI.
Detailed analyses of these observations will give us an insight into the optimization of fuelingscenario.
References[1]M. Wakatani, et al., Nucl. Fusion 40, 569 (2000).
[2]T. Mizuuchi, et al., Contrib. Plasma Phys. 50, 639 (2010).
[3]T. Mizuuchi, et al., in FEC2010, EXC/P8-11.
[4]S. Kobayashi, et al., in EPS2010.
This work is performed underthe auspices of the NIFS Collaborative Research Program.

84

P1-47
Improvement in the fueling efficiency of supersonic gas puffing in LHD
A. Murakami, J. Miyazawa1), C. Suzuki1), I. Yamada1), T. Morisaki1), R. Sakamoto1),
H. Yamada1) and LHD Experiment GroupThe Graduate University for Advanced Studies, Toki 509-5292, Japan
Research Fellow of the Japan Society for the Promotion of Science1)National Institute for Fusion Science, Toki 509-5292, Japan
Two kinds of improvement in the fueling efficiency of supersonic gas puffing (SSGP) have been
observed in the Large Helical Device (LHD). The fueling efficiency improves when the edgedensity is kept high and a strongly hollow density profile is maintained. The fueling efficiency
also improves suddenly when the target plasma is close to the density limit. In SSGP, fuel gasses
are ejected through a fast solenoid valve equipped with a Laval nozzle. Because of the solenoid
valve and the Laval nozzle, higher working pressure of < 8 MPa and more convergent gas flow
than those of ordinary gas puffing are available. Three solenoid valves with different Laval
nozzles are installed in the SSGP device. The efficiency of SSGP, which is defined as a ratio ofincrease in the total number of electrons in the plasma due to SSGP to the number of electrons in
the injected gas, is about 10 - 30 % in the case of hydrogen gas injection. As the electron densityof the target plasma before SSGP injection increases, the fueling efficiency decreases. After SSGP
injection, a hollow electron density profile is formed. The particles are transported from theplasma edge region to the core region through the diffusion process. The fueling efficiency
strongly depends on the edge electron density before SSGP. In order to obtain high efficiency, asteep inward gradient should be maintained. This can be realized when the edge electron density
before SSGP is low enough, or the amount of density increase due to SSGP is higher than the edge
electron density before SSGP. In the case of helium SSGP, where high recycling from wall is
expected, inward density gradient was maintained for a longer time than in the case of hydrogen
SSGP. The fueling efficiency of helium SSGP was higher than that of hydrogen SSGP. In the caseof hydrogen SSGP, abrupt increase in the density increase rate was observed when the pulselength of SSGP was extended. As a result of this, the fueling efficiency was improved two times
even though the difference in the number of supplied particles was less than 20 %. Similar
phenomena are frequently observed when the target plasma is close to the density limit.

85

P1-48
Production and Measurement of Negative Hydrogen Ions in a FET based RF Ion Source
N. Tanaka, T. Matsuno, T. Funaoi, A. Ando, Y. Tauchi 1), H. Nakano 2), K. Tsumori 2)
and Y. Takeiri 2)
Tohoku University, Sendai, Miyagi 980-8579, Japan
1) Yamaguchi University, Ube, Yamaguchi 755-8611, Japan
2) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
High performance ion sources are required for neutral beam injection heating in the next
generation fusion devices. The use of metal-oxide-semiconductor field effect circuit transistors(MOSFET) for the RF power supply enables us to operate RF ion sources with lower frequency(<1 MHz) that is feasible because of large skin depth for high-density plasma production. We havedeveloped a compact FET based negative hydrogen (H-) ion source [1]. The characteristics of theion source have been studied, and the source performance has been improved.
The ion source consists of a driver, an expansion region, and a set of three beam extraction grids[1]. Helmholtz coils are attached around the driver for the improvement of the plasma
confinement. A pair of permanent magnets is attached near the plasma grid to generate a magnetic
filter field. A Cesium oven is attached to the expansion chamber, and the temperatures of the ovenand the plasma grid can be controlled. The typical RF operation frequency is 0.3-0.5 MHz. Plasma
parameters were investigated in the driver, and upstream and downstream of the expansion region.
The H- ions near the plasma grid were measured by laser photodetachment.
The effects of the axial magnetic field and the magnetic filter field were investigated. The electron
density increased as the axial magnetic field increased implying improvement of the plasma
confinement in the driver. The electron densities attained more than 1019 m-3 in the driver region
and 1018 m-3 in the expansion region. The electron temperature downstream of the filter field
drastically decreased to ~1 eV, and the density ratio of H- ions to the electron doubled to 10%. The
H- beam was extracted at ~15 keV. The RF power loss at the isolated transformer should be
minimized for effective plasma production. The developed transformer recorded high RF powertransmission efficiency of almost 90%. Cs effects on the H- production and H- beam extraction
will also be presented.
References[1] A. Ando et. al, Rev. Sci. Instrum. 81, (2010) 02B107

86

P1-49Study of Edge Fluctuation Characteristics using Multiple Langmuir Probes in Heliotron J
S. Ohshima1), K. Hashimoto2), S. Yamamoto, K. Nagasaki, T. Mizuuchi, H. Okada, T. Minami,
S. Kobayashi, K. Hanatani, S. Konoshima, K. Mukai2), H.Y. Lee2), M. Takeuchic3) and F. Sano
Institute of Advanced Energy, Kyoto University, Gokasho, Uji 611-0011, Japan
1)Kyoto University Pioneering Research Unit for Next Generation, Gokasho, Uji 611-0011, Japan2)Graduate School of Energy Science, Kyoto University, Gokasho, Uji 611-0011, Japan
3)Japan Atomic Energy Agency,801-1, Mukaiyama, Naka 311-0193, Japan
Turbulence behavior around last closed flux surface (LCFS) and the contribution to transport are
key issues in plasma fusion researches since performance of plasma confinement is substantiallyinfluenced by turbulence characteristics around edge region, as observed for transitions toimproved confinement modes. A variety of phenomena regarding to edge fluctuations, forexample, transport barriers, meso-scale structures and blob transport, have been focused in many
years. Recently, various advanced analysis techniques and multi-channel diagnostic tools havemarkedly progressed and contributed to the characterization of the phenomena in these studies.
In an advanced helical axis heliotron device, Heliotron J, physics of edge fluctuation is being
studied using multiple Langmuir probes installed at different toroidal/poloidal positions[1-3].
In the plasma heated by neutral beam injection, it was found that low frequency fluctuation
component less than 100 kHz plays a dominant roll in turbulent transport and characteristics of thefluctuation changes depending on the radial location. From the perspective of statistical approach,the distortion of probability density function, characterized by the parameters of kurtosis andskewness, was observed in signals of ion saturation current and floating potential, which may berelative to the intermittent transport such as blob transport. Moreover, phase relationship between
ion saturation current and poloidal electric field fluctuations, which is directly linked with particle
flux caused by fluctuation, varies with probe location although the power spectra of thefluctuations at different radial positions are similar. These results may imply that mechanism ofparticle flux has different characteristics at different location around LCFS. In the presentation,
details of characteristics and the spatial dependence of the fluctuation behavior around LCFS will
be shown and discussed.
References[1] T. Obiki et al. Plasma Phys. Control. Fusion 42, 1151 (2000).
[2] F. Sano et al. Nucl. Fusion., 45, 1557 (2005).
[3] S. Ohshima et al. Proceedings of 38th EPS Conference on Plamsa Physics, P4.134 (2011).

87

P1-50Configuration of flows in a cylindrical plasma device
S. Oldenbürger1), K. Uriu2), T. Kobayashi2), S. Inagaki1,3), Y. Nagashima1,3), T. Yamada1,4),
M. Sasaki1,3), M. Yagi1,3,5), A. Fujisawa1,3), S.-I. Itoh1,3) and K. Itoh1,2,6)
1)Itoh Research Center for Plasma Turbulence, Kyushu University, 6-1 Kasuga-Koen, Kasuga
816-8580, Japan
2)Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-1 KasugaKoen, Kasuga 816-8580, Japan
3)Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasuga-Koen, Kasuga 8168580, Japan4)Graduate School of Frontier Science, University of Tokyo, Tokyo 277-5861, Japan
5)Japan Atomic Energy Agency, 801-1 Muko-yama, Naka-city, 311-0193, Japan6)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
The investigation of flow velocity fields in turbulent magnetized plasmas is a crucial issue for theunderstanding of a wide range of phenomena. Recently, the observation of toroidal flows has
increased the interest in studies of parallel momentum flux and its link to plasma turbulence.
Detailed investigation of flows is a challenging task, especially in bigger devices, due to thecomplexity of the plasma configuration and the limitations for diagnostics. Laboratory plasma
devices with their high reproducibility and easier experimental access can help to understandfundamental relationships between flows and turbulence.
In the cylindrical magnetized plasma device PANTA, the ion flow in the helicon argon plasma isstudied with three different methods. The time delay estimation technique can be used to measure
the apparent phase velocity of fluctuations between two electrostatic probes. This technique willtake into account not only the mean plasma flow, like the E× B rotation in azimuthal direction, butalso the propagation of waves [1]. The cross-field convective particle flux due to plasma
instabilities is measured using three-tip Langmuir probes. Finally, the mean ion flux is studied inazimuthal and axial direction using Mach probes [2]. Measurements of mean axial ion flow
measurements show a reversed flow direction in the outer part of the plasma column in highneutral pressure conditions. In the same conditions an inwards directed radial cross-field particle
transport is observed close to the radial position of axial flow inversion. This contribution presents
results of flow studies in different discharge conditions and discusses possible links between
plasma turbulence and flows in the radial, axial and azimuthal directions.
References[1]J. H. Yu, C. Holland, G. R. Tynan, G. Antar, Z. Yan, J. Nucl. Mater, 363-365 (2007) 728[2] I. H. Hutchinson, Principles of Plasma Diagnostics, Cambridge University Press (2002),
page 82
This work is partly supported by a grant-in-aid for scientific research (S) from JSPS (21224014), the grant-in-aid
for Scientific Research B (23360414), the collaboration programs of RIAM Kyushu University and NIFS
(NIFS10KOAP023).

88

P1-51Time Evolution of Power Spectrum Density during Spontaneous Transition in Cylindrical
Magnetized Plasma
T.Kobayashi, S.Inagaki1,2), S.Oldenbürger2), K.Kawashima, N.Ohyama, Y.Tobimatsu,
H.Arakawa3), M.Sasaki1,2), Y.Nagashima1,2), T.Yamada2,4), S.Sugita1), M.Yagi1,2,3), N.Kasuya2,5),
A.Fujisawa1,2), S.-I.Itoh1,2) and K. Itoh2,5)Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Kasuga, Fukuoka
816-8580, Japan
1)Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580,
Japan2)Itoh Research Center for Plasma Turbulence, Kyushu University, Kasuga, Fukuoka 816-8580,
Japan3)Japan Atomic Energy Agency, Naka-city, Ibaraki 311-0193, Japan4)Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa, Tokyo 277-8561,
Japan5)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Several stable turbulent states exist in magnetized plasmas and fast transitions between them areobserved both in linear and toroidal plasmas. Understanding of such turbulent states andtransitions is an important issue both in physics of non-equilibrium system and in nuclear fusion
research. In the Large Mirror Device-Upgrade (LMD-U), which is a low temperature cylindricalplasma device, mainly two kinds of turbulent states were successfully generated depending on theneutral gas pressure condition Pn [1,2], i.e., a state of solitary waves in the high Pn case and that ofbroadband in the low Pn case. A transition phenomenon between them was studied in detail.
Recently, a new transition among states of broadband, which have different turbulent spectra, isfound in the Plasma Assembly for Nonlinear Turbulence Analysis (PANTA, the successor deviceof the LMD-U) in low neutral gas pressure condition at the case of Pn = 1.00 mTorr. Under thiscondition, the fluctuation spectra and radial profiles transit between two different states. At lowerand higher neutral pressure conditions, spectra and profiles are stable. First, we analyzed transition
properties for radial profiles. The ion saturation current profile (which can be index of the electron
density fluctuation in the PANTA case [3]) changes ∼ 40% within the order of millisecond. The
transition propagates axially with a speed of ∼ 1 km/s that is 30–50% of the ion sound speed.
Second, we obtained a time evolution of two-dimensional frequency and azimuthal mode number
power spectrum at the transition, using short time FFT technique. We discuss the time scale ofboth growth and decay of fluctuation amplitudes.
References[1] H. Arakawa, et al., Plasma Phys. Control. Fusion, 51 (2009) 085001[2] H. Arakawa, et al., Plasma Phys. Control. Fusion, 52 (2010) 105009[3] K. Kawashima, et al., Plasma and Fusion Res., (2011), in press
This work is partly supported by a Grant-in-Aid for Scientific Research S from JSPS (21224014), the Grant-inAid for Scientific Research B from JSPS (23360414), the collabo- ration programs of RIAM Kyushu University
and NIFS (NIFS10KOAP023).

89

P1-52
Evaluation of Electron Temperature Fluctuations Using a Conditional Technique on
PANTA
S.Yamada, S.Inagaki1,2), S.Oldenbürger1), T.Kobayashi, K.Kawashima, N.Ohyama, Y.Tobimatsu,
H.Arakawa3), Y.Nagashima1,2), T.Yamada1,4), M.Sasaki1,2), M.Yagi1,2), N.Kasuya1,4),
A.Fujisawa1,2), S.-I.Itoh1,2) and K.Itoh1,5)Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-1 Kasuga-Koen,
Kasuga 816-8580, Japan
1)Itoh Research Center for Plasma Turbulence, Kyushu University, 6-1 Kasuga-Koen, Kasuga
816-8580, Japan
2)Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasuga-Koen, Kasuga 8168580, Japan
3)Japan Atomic Energy Agency,Tokai-mura,Naka-gun,319-1195,Japan4)Graduate School of Frontier Sciences, The University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa
277-8561, Japan5)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanTo understand the turbulence structures and their formation mechanism in magnetized plasma,
fluctuations are measured with multi-Langmuir probe arrays in the PANTA linear plasma device.
Observation of temperature fluctuations is an important issue to evaluate fluctuation-driven heattransport. Langmuir probes are useful tools to measure plasma parameters and electron
temperature can be evaluated from their current-voltage (I-V) characteristic curve. However, it isdifficult to obtain I-V curves with high temporal resolution.An additional technique to evaluate
electron temperature fluctuation is required. One candidate is the conditional averaging method[1][2].
In the conditional method, the I-V curves are re-sampled based on the phase of the fluctuation ofion saturation current or floating potential measured with another probe. From the reconstructed IV curves, we could obtain simultaneous measurements of electron temperature fluctuation andother fluctuating plasma quantities. In PANTA, correlation between electron temperaturefluctuation and floating potential fluctuation was observed using this method in conditions where
the fluctuation spectrum has multiple peaks[3] . In this paper we establish the electron temperaturefluctuation measurement based on the conditional averaging method. For simplification, we
excited a coherent fluctuation which has single peak in spectrum (∼3KHz). We extracted
correlation between electron temperature fluctuation and floating potential fluctuation and the one
between electron temperature fluctuation and ion saturation current fluctuation, and compared theresults with the triple probe method. Effectiveness of the conditional method is thus verified inplasmas with two different fluctuation patterns. In addition, we will discuss the application of thismethod to plasmas where broad fluctuation spectra are observed.
This work is partly supported by grant-in-aid for scientific research from JSPS (21224014,
23360414), the collaboration programs of Research Institute for Applied Mechanics Kyushu
University and NIFS (NIFS10KOAP023).
References[1] K.Kawashima, Plasma Fusion Res.6(2011) 2406118[2] Y.Nagashima, Plasma Fusion Res.5(2010) S2049[3] H.Arakawa submitted to Plasma Phys. and Controlled Fusion(2011)90

P1-53Magnetic Helicity Injection Experiments for Double-null Startupof the UTST spherical tokamak
T. G. Watanabe, S. Kamio, Q. H. Cao1) , H. Itagaki, N. Suzuki,
K. Takemura, K. Yamasaki, K. Ishiguchi2), T. Yamada, M. Inomoto and Y. Ono
Graduate School of Frontier Scineces, The University of Tokyo,5-1-5 Kashiwanoha, Kashiwa 2778561, Japan
1)Graduate School of Engineering, The University of Tokyo,7-3-1 Hongo, Bunkyo 113-8656,
Japan2)The Open University of Japan, 2-11, Wakaba, Mihama-ku, Chiba, 261-8586, JapanA spherical tokamak (ST) plasma is produced by double-null merging (DNM) method by using
two pairs of external poloidal field coils in the University of Tokyo Spherical Tokamak (UTST)
device [1]. In the late phase of the DNM formation, a single ST is connected to the external
poloidal coil flux, which enables magnetic helicity injection from the helicity source (the coil flux)
into the helicity sink(the ST plasma).
We observed the relationship between magnetic helicity and magnetic reconnection by twodimensional pickup coil arrays, which are located in the r-z plane [2].
We performed a parameter scan to check how crucial and effective the plasmoid ejection is forhelicity injection. Thus, we investigated on what parameter the plasmoid formation depends andhow helicity injected becomes when the formation is changed.
References[1] T. Yamada et al., Plasma Fusion Res. 5, (2003) S2100.
[2] T.G. Watanabe et al., Plasma Fusion Res. 6, (2011) 1202131.

91

P1-54Modeling of Vacuum Field in Start-up in EAST
Xue Erbing1,4), Luo Jiarong2), Liu Xiaolong3), Kazuo Nakamura4)1)College of Information Science and Technology, Donghua University, Songjiang, Shanghai,
China2)Department of Physics, Donghua University, Songjiang, Shanghai, China
3)Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Kasuga,
Fukuoka, Japan
4)Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
Enough loop voltage and large null field region is necessary for EAST to break down and initiate
the plasma current. Eddy currents in conducting structural elements perturb the loop voltage andfield not only spatially, but also temporally. Due to the individually powered poloidal fieldmagnets with specially designed compensation coils for the field, currents in each coils have to be
adjusted to achieve enough large null field region and desired loop voltage for the breakdown andinitial ramp-up. This paper summarized the key factors which influence the plasma start-up[1].
Three algorithms were used to calculate the vacuum field. The first algorithm uses both magnetic
flux loop and magnetic probe to calculate the vacuum field by dividing the vacuum vessel into 80
elements and estimating the eddy current distribution. The second algorithm using only themeasured values of the magnetic flux loop is used to calculate the vacuum field based on theboundary integral equations[2]. The third algorithm is used to calculate the vacuum field directlyfrom the magnetic sensor signals similarly with Cauchy condition surface method[3]. Calculation
results of these three algorithms agree well by comparing the results of the three algorithms, which
proves the reliability of the models.
References[1] E.B. Xue, J.R. Luo, S.B. Shu, Optimization for Plasma Breakdown Based on Genetic
Algorithm in EAST, Proc. International Conference on Computational and Information
Sciences, Chengdu, 2010, pp.1285-1288.
[2] B.J. Braams, Interpretation of Tokamak Magnetic Diagnostics, Plasma Phys. Controlled
Fusion 33 (1991) 715-748.
[3] K. Kurihara, A New Shape Reproduction Method Based on the Cauchy-Condition
Surface for Real-Time Tokamak Reactor Control, Fusion Eng. Des. 51-52 (2000) 1049-1057.

92

P1-55
FDTD simulation analysis for improving the two dimensional electron density distribution
measurement by using phase imaging method in GAMMA10
T. Yamada, M. Yoshikawa, J. Kohagura, Y.Shima, Y. Morikawa, D. Akita, M. Sakamoto, T. Imai
Plasma Research Center, University of Tsukuba, Ibaraki 305-8577, JapanGAMMA10 is the tandem mirror device utilizing plasma confinement by magnetic mirrors andelectrostatic potentials. It consists of the central, anchor, and plug/barrier mirror cells. In
GAMMA10, plasma is produced and contained by the ion cyclotron heating. The ion and electron
confining potentials are formed by applying the electron cyclotron heating at plug/barrier cells,
respectively. Microwave interferometers are placed at each cells for measuring the plasma electron
densities. A phase imaging interferometer system is set horizontally to measure upper half of theplasma in the plug cell. It consists of two microwave oscillators (69.85 and 70.00 GHz), lenses,
mirrors, and two dimensional detector. The detector position can be moved in the radial or axial
directions. The phase imaging method is the method to measure the phase difference between the
microwave beam passing through the plasma and the reference microwave beam which depends
on the line integrated electron density. We can obtain two-dimensional line integrated electron
density distributions by using the two dimensional detector. With application of the Abeltransform technique to the line integrated electron density distributions, we can obtain the electron
density radial distributions. Normally, the Abel transform is assumed axisymmetric distribution ofplasma density. However the plasma profile is not always axisymmetry. Then, using Abeltransform method for nonaxisymmetric plasma may have large error in the calculated electron
density. The purpose of this study is to optimize the assumption of density distribution of the Abeltransform with the use of Finite Difference Time Domain (FDTD) method, and to improve themeasurement accuracy of two dimensional density distribution by use of the phase imaging
method. We compare the phase differences obtained by experimental measurement and thoseobtained by FDTD simulation. Next, we use calculated density distribution which was good
agreement to the assumption of the Abel transform and FDTD simulation. In this paper, we report
the improvement of the electron density distribution measurement by using this method.

93

P1-56
Observation of spectral broadening of LH waves in Alcator C-Mod
S.G. Baek, R.Parker, S. Shiraiwa, G. Wallace, A. Dominguez, G.J. Kramer1)
Plasma Science and Fusion Center, MIT, Cambridge, MA, USA
1) Princeton Plasma Physics Laboratory, Princeton, NJ, USA
Lower hybrid current drive (LHCD) experiment in Alcator C-Mod indicates that lower hybrid(LH) waves are not strongly Landau-damped inside separatix in high density plasmas with the line
averaged density over 1020 m-3[1]. While the collisional loss of LH power in the scrape off layer(SOL) is a working model to explain this phenomenon in Alcator C-Mod, little attention has been
given to the effect of spectral broadening of LH waves.
In order to characterize the spectral width of LH waves near separatix in Alcator C-Mod, we haveup-graded the existing O-mode reflectometer system into a scattering diagnostic to detect LH
waves away from the LH launcher. The measured full width at half maximum (FWHM) is order of
1 MHz, which is wider than FWHM of incident waves (500 kHz). This broadened width of
scattered signals can be explained by the model proposed by Andrews and Perkins in which
multiple scattering of LH waves by low frequency density fluctuations determines the frequency
width of LH waves as LH waves cross turbulent SOL [2].
In addition to scattering diagnostics, the spectral width observed by the Langmuir probe indicatesthat spectral width of LH waves at -35dBc becomes narrower from 17 MHz to 7 MHz as the
distance between the inner wall and the last closed flux surface decreases from 1.5 cm to 0 cm inthe upper null configuration. Interestingly, strongest non-thermal emission is observed when the
spectral broadening is minimum. This broadening is more likely due to parametric decay
instabilities (PDIs) because it is typically accompanied with enhancement of sidebands with 30
MHz spacing. A numerical study of PDIs will be performed to estimate the role of spectral
broadening on the spectrum of parallel refractive index.
References[1]G. Wallace, Proceedings of the 19th Topical Conference, 2009[2] P.I. Andrews and F.W. Perkins, Phys. Fluids 26(1983) 2546
This work supported by US DOE award DE-FC02-99ER54512.

94

P1-57
Identification of MHD Mode Structure using ECE measurement
H.Tsuchiya, Y.Takemura1), Y.Nagayama, K.Watanabe, S.Sakakibara
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)The Graduate University for Advanced Studies, Toki, Gifu 509-5292, Japan
It is recognized that the controlling of MHD instability is important issue to improve the plasma
performance. However the mode structure is less well understood even though the study using themeasurement of SX(soft-Xray) has been done in LHD [1]. In this study, the result of themeasurement of mode structure used by ECE (Electron Cyclotron Emission) [2] is reported.
Usually, MHD or high-beta experiment are conducted in the condition of low magnetic filed Bt >1.0 T in order to clear the effect on the magnetic field. In that condition, ECE measurement cannot
be used because of the cut-off in LHD. Instead, SX emission measurement is a helpful tool forfluctuation studies. However it is not adequate to identification of mode structure, in principle,
since the obtained SXR emission intensity is included the integral effect along the line-of-sight. If
the result of identification of mode structure used by SX is supported by the ECE measurement.
This study helps the confinement study of the low filed discharge.
The experiment is conducted using ECE measurement in the condition of Rax=3.6m,Bt=1.5T,
which is suitable for ECE to measure in 0.1<ρ<1. The MHD fluctuation of low-order mode which
is seem to be m/n = 1/1 are observed by SX measurement, magnetic probe (MP) and ECE. Thereis the significant coherence among SX fluctuation, magnetic fluctuation and electron temperaturefluctuation around ρ = 0.85, which corresponds to the area ι = 1. To clarify the mode structure, we
calculated the displacement amplitude ξ. When SX signal can not show the structure clearly, we
can see the mode structure around ρ ∼ 0.9 using ECE data, which is local measurement tool. The
coherence radial profile calculated by the ECE shows also the mode width clearer than that of SX
measurement.
References[1] S.Masamune, 23rd IAEA Fusion Energy Conference, Daejon, Republic of Korea, EXS/P5-11[2] H. Tsuchiya et.al. Plasma Fusion Res. 6 (2011) 2402114.
The work is supported by NIFS (NIFS11ULPP008,NIFS11ULPP020).

95

P1-58
Fast, nonlocal, radially coherent changes of Te during ELM crashes observed by 2D ECE
imaging in KSTAR
I. Hong, G.S. Yun, W. Lee, H.K. Park, C.W. Domier1), N.C. Luhmann, Jr.1), A.J.H. Donné2) and
KSTAR Team3)
POSTECH, Pohang, Gyeongbuk 790-784, Korea
1)University of California at Davis, Califorina 95616, USA
2)FOM Institute for Plasma Physics Rijnhuizen, Nieuwegein, The Netherlands3)National Fusion Research Institute, Daejeon 305-806, Korea
Local electron temperature (Te) changes throughout the entire cross-section that is correlated withlarge ELM crashes of H-mode plasmas have been observed by a 2D ECE imaging system in
KSTAR. During the pedestal collapse, the core region showed a fast Te drop and recovery in atime scale less than 100 μs whereas the intermediate region showed the opposite Te trend.
Interestingly, the core density and stored energy during the ELM crash were observed to be
recovered in a diffusive time scale (~ 100 ms). The preliminary analysis in the perspective of nonlocality will be discussed.

96P1-59
Electron Density Fluctuation Measurement with 3-D Microwave Imaging Diagnostics in
LHD
Y. Nagayama, D. Kuwahara1), Y. Hamada, H. Tsuchiya, T. Yoshinaga2), S. Yamaguchi3),
Y. Kogi4), A. Mase5), S. Tsuji-Iio1)
National Institute for Fusion Science, Toki 509-5292, Japan
1) Tokyo Institute of Technology, Tokyo 152-8550, Japan
2) National Defence Academy of Japan, Yokosuka 239-8686, Japan
3) Kansai University, Suita 564-8680, Japan
4) Fukuoka Institute of Technology, Fukuoka 811-0295, Japan.
5) KASTEC, Kyushu University, Kasuga 816-8580, Japan
By measuring the local electron density, the micro and macro instabilities can be investigatedexperimentally. The Microwave Imaging Reflectometry (MIR) is one of the most promising
diagnostics of the local electron density. In MIR the plasma is uniformly illuminated with themicrowave, and the plasma wave reflects it at the cutoff density. Imaging optics makes the imageof fluctuating plasma on the imaging detector. The electron density fluctuation causes thefluctuation of phase of the reflected wave primarily. Also the amplitude fluctuation is caused by
the density fluctuation. Therefore both the phase (&phi) and the amplitude (A) of the reflectedwave are detected in this system. The phase is detected with the I-Q demodulator that makes the I(~cos &phi) and Q (~sin &phi) signals.
In the large helical device (LHD), we have developed 3-D MIR system [1]. In this system, theillumination wave consists of 4 frequencies (RF1: 60.410 GHz, RF2: 61.808 GHz, RF3: 63.008
GHz, RF4: 64.610 GHz) in order to observe 4 different plasma layers. The imaging detector is a
newly developed 2-D Horn-antenna Mixer Arrays (HMA). By using 4 frequencies illuminationand the 2-D HMA, we can measure the electron density fluctuation 3 dimensionally.
In the case of low density plasma, where the cutoff surface is absent, the I and Q signals are
sinusoidal, of which amplitude is constant. The phase of I signal is 90 degrees ahead of that of Q
signal, as I and Q correspond to cos &phi and sin &phi, respectively. The microwave is reflected
at the vacuum vessel wall, and the phase change indicates the electron density change.
In LHD, the edge harmonic oscillation (EHO) is often observed. The feature of EHO is thefrequency spectrum that consists of equally separated higher harmonics. EHO is a fluctuation in
the edge plasma, and it was observed in the VH mode plasma in DIII-D when the ELM is absent.
Usually, LHD plasma has a steep electron density gradient at the edge. The MIR image shows that
EHO has a narrow structure along a field line. The EHO is localized near the iota=1.5 surface,where R=4.525 m. Since the channel separation is 2 cm, the width of EHO may be 5 cm along theflux surface. Thickness of EHO in the radial direction is 2 cm. As the phase signals (I, Q) change
rapidly, the reflection surface moves rapidly in the radial direction. This is the first 3-D MIR
measurement in the world.
References[1] T. Yoshinaga, Y. Nagayama, D. Kuwahara, et al, PFR. 5, 030 (2010).
[2] D. Kuwahara, S. Tsuji-Iio, Y. Nagayama, T. Yoshinaga, et al., Rev. Sci. Instrum 81,
10D919 (2010).
The work is supported by KAKENHI (No.21246140), NIFS (NIFS11KEIN1111, NIFS11ULPP008) and NINS
Imaging Science Project (NIFS10KNSI001).

97

P1-60The development of the Infrared TV system for KSTAR
Dongcheol Seo1), Seungtae Oh1), Chansoo Kang2) and KSTAR team1)National Fusion Research Institute, 169-148 Gwahangno, Daejeon 305-806, Korea
2)University of Science & Technology, 217 Gajungro, Daejeon 305-350, Korea
In the KSTAR tokamak, the thermal imaging camera was developed for to measure heat flow to
the poloidal limiter, to the ICRH antenna, to the inboard limiter and to the passive stabilizer andhave been installed. The long view ports make a periscope inevitable to get a field of view for
diagnosing the internal space of the vacuum vessel and better observation of plasma discharge on
KSTAR. The specifications of the Infrared camera(FLIR/ThermoVision SC6000) are 640×512
pixels resolution, full frame rate of 125Hz, and Noise equivalent temperature difference(NETD) of<25mK. We can determine both the location and the intensity of the heating on the poloidal
limiter, on the passive stabilizer and inboard limiter during heating power into plasma. The
ExminIR which is provide from FLIR systems was used for the camera control and Data storage.
Data is stored in digitized forms.

98P1-61
Cylindrical Surface Wave on Periodically Corrugated Metal Cylinder
H.Iiduka, K.Ogura, K.Yambe
Graduate School of Science and Technology, Niigata University, Niigata 950-2181, Japan
Electromagnetic waves from microwave to terahertz wave are demanded for widespread
applications such as plasma heating, plasma diagnostics, telecommunication systems, radarsystems. In such a wide range, conceivable electromagnetic wave sources are Smith-Purcell freeelectron laser (SP-FEL) [1,2] and backward wave oscillator (BWO) [3]. In SP-FEL, an electron
beam passes near the grating surface and a spontaneous radiation may be excited. SP-FEL in Ref.
[1] realizes a stimulated radiation and attract attention to develop a compact source of highintensity terahertz wave. The stimulated SP emission requires a beam bunching by BWO operation
at the beam interaction with the surface wave of the grating. The SP devices mentioned arecommonly based on plane geometry. Another realistic geometry is cylindrical one, which is very
effective for high-power operations of BWOs. An idea of SP-FEL using cylindrical SWS seems tobe very attractive to improve the radiation intensity as shown in Ref. [4]. To realize intense SP
emissions, properties of cylindrical surface waves should be studied more definitely.
In this work, we study cylindrical surface waves on metal cylinders with rectangular corrugations.
The properties of the surface waves due to the corrugation are examined. In the practicalexperiments, the metal cylinder has a finite length. Reflections from both ends quantize theelectromagnetic modes into axial resonant modes. The axial modes formed by the surface wave
are measured using an experimental setup based on a network analyzer. There is another kind ofcylindrical surface wave caused by surface plasmons, i.e., Sommerfeld wave. We confirm theaxial modes by Sommerfeld wave and discuss two kinds of cylindrical surface wave. For
Sommerfeld wave, the dispersion characteristics are fixed by the physical property of metal
surface. On the other hand, the cylindrical surface wave due to periodic structure can have an
arbitrary dispersion [5], and might be very useful to cylindrical SP-FEL and electromagnetic
waveguiding up to terahertz wave.
References[1] J. Urata et al., Phys. Rev. Lett. 80, 516 (1998).
[2] H.L. Andrews et al., Phys. Rev. ST Accel. Beams, 12, 080703 (2009).
[3] K. Ogura et al., IEEJ Trans. FM, 127, 681 (2007).
[4] K. Ogura et al., Plasma Fusion Res. 6, 2401039 (2011).
[5] J.B. Pendry et al., Science, 305, 847 (2004).

99

P1-62
Vertical profiles of EUV spectral emissions from edge impurity ions observed at different
toroidal angles in LHD
Erhui WANG1), Shigeru MORITA1,2), Motoshi GOTO1,2), Chunfeng DONG1)
and Masahiro KOBAYASHI1,2)
1)Department of Fusion Science, Graduate University for Advanced Students, Toki 509-5292,
Gifu, Japan2)National Institute for Fusion Science, Toki, 509-5292, Gifu, Japan
Magnetic field configuration of Large Helical Device (LHD) has three-dimensional structure. The
elliptical poloidal cross section mainly formed by two helical coils rotates five times during a
toroidal turn. The presence of thick stochastic magnetic field layer surrounding the last closed flux
surface (LCFS), so called 'ergodic layer', is a typical character indicating the three-dimensional
structure of magnetic filed lines in LHD. The magnetic field lines in the ergodic layer are radiallydeviated when they toroidally move, while the connection length of field lines is long enough to
confine the edge plasma. Therefore, the three-dimensional measurement of edge plasmas is very
important to study the transport mechanism of the stochastic magnetic field layer. A spaceresolved EUV spectrometer working in range of 50-500A has been recently improved formeasuring plasma poloidal cross section at different toroidal angles by adding a toroidal scanningmechanism, although the vertical profile of EUV emissions has been observed at horizontallyelongated plasma cross section, in which the observation chord passes through both the inboard
and outboard X-points. Then, direct observation of the inboard and outboard X-points is possible
at different toroidal angles by scanning horizontal angle of the observation chord in the EUV
spectrometer. Vertical profiles of several spectral lines emitted from ergodic layer have been
observed at different toroidal angles with vertical spatial resolution of 15 mm. The edge boundaryof LHD plasmas is studied near inboard and outboard X-points using CIV (312.4A) withionization energy of 64eV located in the farthest edge of ergodic layer. Analysis on the edge
boundary near both the X-points is presented with vertical profiles of EUV emission lines in theergodic layer at different toroidal angles.

100P1-63Extension of wavelength range in absolute intensity calibration of space-resolved EUV
spectrometer for LHD diagnostics
C. F. Dong1), S. Morita1,2), M. Goto1,2) and E. H. Wang1)
1)Department of Fusion Science, Graduate University for Advanced Studies, Toki 509-5292, Gifu,
Japan2)National Institute for Fusion Science, Toki 509-5292, Gifu, Japan
Absolute intensity calibration of a flat-field space-resolved extreme ultraviolet (EUV)
spectrometer has been carried out in wavelength range of 80-400A by comparing the radial profile
of EUV bremsstrahlung continuum with that of visible bremsstrahlung continuum at 5300A
emitted from high-density plasmas in Large Helical Device (LHD). Recently, the EUV
spectrometer has been improved to explore the impurity line emissions existing in shorter andlonger wavelength ranges beyond the original wavelength range, and the observable wavelength
range could be successfully extended to 30-500A. The extension of wavelength range makes
possible to measure the first resonance lines of BIV (61.1A), BV (48.6A), CV (40.3A) and CVI(33.7A) in addition to NeVII (465.2A: 2s2p-2s2), which are essential to edge impurity transport
study in LHD. The absolute intensity calibration has been also done in such the extended
wavelength range of 30-500A with the same manner as before. The result on the calibration ispresented and compared with the previous result. The present calibration extended to shorter
wavelength range can also improve an alternative approach to Zeff diagnostics based on EUV
bremsstrahlung continuum. In particular, the stronger bremsstrahlung intensity in shorter
wavelength range can lower the density limit (ne ∼ 3×1013cm-3) in the Zeff measurement. Detailed
examination on the calibration factor in shorter wavelength range can also bring an interesting
result on the reflectivity of EUV emissions for presently using holographic grating.

101

P1-64
EUV Spectroscopy of Highly Charged Tungsten Ions with Electron Beam Ion Traps
H. A. Sakaue, N.Yamamoto1), D. Kato, I. Murakami and N. Nakamura2)
National Institute for Fusion Science, Toki, Gifu 509-5292, JAPAN
1)Chubu University, Kasugai, Aichi, 487-8501, JAPAN2)Institute for Laser Science, The University of Electro-Communications, Chofu, Tokyo 182-8585,
JAPAN
Tungsten is a strong candidate for the material of the plasma-facing components, such as the firstwall and divertor plates, in ITER because it has excellent physical properties, such as a low
sputtering yield, a high melting point, a low tritium inventory, and so on. However, since highparticle and heat fluxes would cause serious damages to such components, tungsten is considered
to be one of the most abundant impurities in the ITER plasma. Emission lines of highly charged
tungsten ions thus play an important role in the spectroscopic diagnostics of the ITER plasma, andconsequently the spectroscopic data of tungsten ions are strongly needed. To date, thespectroscopic studies of highly charged tungsten ions have been carried out by using hightemperature (several keV) plasmas [1-2] and electron beam ion traps (EBITs) [1, 3-4] withrelatively high electron energies (several to several tens keV). However, there is not yet enoughdata. In particular, the spectral data of moderate charge state tungsten ions (W10-30+) are quite
limited.
For the systematic spectroscopic studies of tungsten ions, we have been using the low-energy
compact EBIT (CoBIT) [5,6] developed for spectroscopic studies of moderate charge state ions.
We have constructed two CoBIT, and installed them at the UEC and National Institute for Fusion
Science. The electron energy range of CoBIT is 0.1 - 2.5 keV, and the accessible charge states areabout 10 – 40. In this paper, we present extreme ultraviolet (EUV) obtained with the electronenergy range 100 - 2000 eV. Grazing-incidence flat-field grating spectrometers especiallydesigned for CoBIT were used for EUV emission in the 1 - 20 nm range. The electron energy
dependence of line intensity and its application to the plasma diagnostics are also discussed.
This work is performed with the support and under the auspices of the NIFS Collaboration
Research program (NIFS09KOAJ003).
References[1]T. Putterich et al., J. Phys. B 38, 3071 (2005).
[2]M. B. Chowdhuri et al., Plasma and Fusion Research 2, S1060 (2007).
[3]J. Clementson and P. Beiersdorfer, Phys. Rev. A 81, 052509 (2010).
[4]Y. Ralchenko et al. J. Phys. B 40, 3861 (2007).
[5]N. Nakamura et al., Rev. Sci. Instrum. 79, 063104 (2008).
[6]H. A. Sakaue et al. J. Phys.:Conf. Ser. 163, 012020 (2009).

102

P1-65Collisional-Radiative Modeling of moderate ionized Tungsten Plasma
Xiaobin Ding, Izumi Murakami, Daiji Kato, Hiroyuki A. Sakaue, Fumihiro Koike1),
Chezhong Dong2)
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
1)School of Medicine, Kitasato University, Sagamihara, Kanagawa 252-0373, Japan
2)College of Physics and Electronic Engineering, Northwest Normal University, Lanzhou,
730070, China
Tungsten (W) attracts the research interests recently because it will be a good candidate for thedivertor or wall material in the next generation magnetic confinement fusion reactors due to it’sfavorable properties. It also has motivated a number of theoretical and experimental investigationsof its radiative properties over a rather large temperature range. Plasma electrons collide withtungsten ions that produce excited states and radiate high energy photons in the main core plasma,
while it can promote significant cooling in the divertor region.
A detailed fine structure collisional radiative model with thousands of levels is developed to
calculate the radiative properties of W27+ tungsten plasma based on the atomic data from
relativistic configuration interaction calculation by using FAC (Flexible Atomic Code) code. The
radiative transition, electron collision excitation and their inverse process are taken into account toconstruct the rate equation. The intensity ratio between some neighboring transitions in EUV and
X-ray region has been found to be sensitive to the electron density which could be good candidate
to diagnose the electron density of plasma. The present work should also be of interest to thecontinuing efforts to model the radiative losses in large magnetic fusion devices.
References[1] Fujimoto, T., Plasma Spectroscopy,Oxford University Press, 2004[2] J Abdallah Jr, J Colgan, R E H Clark, C J Fontes and H L Zhang, J. Phys. B: At. Mol.
Opt. Phys. 44 (2011) 07570[3] Gu M.F., ApJ, 582 (2003) 1241
[4] J Yanagibayashi, T Nakano, A Iwamae, H Kubo, M Hasuo, K Itami, J. Phys. B: At. Mol.
Opt. Phys. 43 (2010) 144013[5] Jun Yanagibayashi, Tomohide Nakano, Atsushi Iwamae, Hirotaka Kubo, Masahiro
Hasuo, Kiyoshi Itami, Nucl. Instru. and Meth.in Phys. Res. Sec. A: 623 (2010) 741
This work performed partly as one of the subjects of IAEA’s Co-Ordinated Research Project (CRP) in‘Spectroscopic and Collisional Data for Tungsten from 1 ev to 20 kev’ and partly with the support and under the
auspices of the NIFS Collaboration Research Program (NIFS09KOAJ003 ).

103

P1-66
Steady state recombining plasma in a radio-frequency plasma device for divertordetachment study
A. Okamoto, H. Takahashi, Y. Kawamura, A. Daibo, T. Kumagai, S. Kitajima, M. Sasao
Tohoku University, Sendai, Miyagi 980-8579, Japan
Transient increase of high energy particle flux flowing into the divertor region, which is inducedby the edge localized modes (ELMs), is a topic of fusion reactor development. Those high energy
particles interact with ions and electrons in the plasma. The plasma density and temperature arethen concerned to be changed transiently, which might results in reduction of recombination ratein the detached plasma[1].
Use of a table top linear device with an energy controlled ion/electron beam source is expected tosupport the divertor development in a viewpoint of quantitative understanding of atomic processes.
However, compatibility among (1) beams penetrating plasma source, (2) high density plasma
production, and (3) low temperature plasma in a test region is an troublesome matter.
In this presentation, development of an steady state recombining plasma using a radio-frequency(RF) plasma device, DT-ALPHA is described. The DT-ALPHA device has the beam penetratingcompatibility since the high density plasma is produced using a cylindrically winded RF antenna.
The electron temperature required for detached/recombining plasma study is on the order of 1 eV,
while typical electron temperature in RF plasmas are about 10 eV. In order to obtain such a lowtemperature plasma from the RF plasma source, a gas cooling method was applied in the test
region. Orifices for gas conductance control and an differential pumping system were installed in
the DT-ALPHA device to prevent the cooling gas from affecting plasma production and beam
penetration. When the target gas pressure increased up to 50 Pa, change of the source pressure waswithin a factor of two showing good isolation of plasma production condition from the coolinggas. Then a plasma with about 2 eV of the electron temperature was obtained. Recombining
spectrum is also discussed in the presentation.
References[1]A. Okamoto, et al., J. Nucl. Mater. 363-365, (2007) 395
This work was partly supported by the Ministry of Education, Culture, Sports, Science and Technology of Japan(MEXT) Grant-in-Aid for Young Scientists (B), 22740357.

104

P1-67
Initial Studies of Microwave-Induced Atmospheric Plasma Jets Using Optical Emission
Spectroscopy
M.C.Lacdan1), J.A.Ting1), H.J.Lee1)2), L.M.Rosario1)3), H.Ramos1), R.Tumlos1)4)1)Plasma Physics Laboratory, National Institute of Physics, University of the Philippines,
Diliman, Quezon City 1101, Philippines2)IBF Electronic GmbH & Co. KG, Dr. Robert MurjahnStrasse, 12 D-64372 Ober-Ramstadt,
Germany
3)Sciences Department, College of Arts, Science, and Education, FEATI University, Helios St., Sta
Cruz, Manila 1003, Philippines4)Department of Physical Sciences and Mathematics, College of Arts and Sciences, University of
the Philippines, Manila, Taft Avenue, Manila 1000, Philippines
Preliminary studies on a microwave-induced atmospheric plasma jet were done using opticalemission spectroscopy (OES). The microwave plasma jets were generated and sustained by acontinuous-wave magnetron. The active species from various atmospheric microwave-induced
plasma discharges of Ar, Ar-N2, Ar-O2, and He were determined. The effects of the absorbed
power and gas flow rate on the relative population of the active species were also investigated. The
results were found to be useful in various future plasma applications.
References
1. E. A. H. Timmermansa, J. Jonkersa, I. A. J. Thomasa, A. Roderob, M. C. Quinterob, A.
Solab, A.
Gamerob and J. A. M. van der Mullena, “The behavior of molecules in microwave-induced
plasmasstudied by optical emission spectroscopy. 1. Plasmas at atmospheric pressure,”
Spectrochimica Acta Part B: Atomic Spectroscopy, Volume 53, Issue 11, Pages 1553-1566. 2
October 1998
2. T. Yuji, S. Fujii, N. Mungkung, H. Akatsuka , “Optical Emission Characteristics ofAtmospheric-Pressure Nonequilibrium Microwave Discharge and High-Frequency DC Pulse
Discharge Plasma Jets,” IEEE
Transactions on Plasma Science, Volume 37 No. 6, pp. 839-845.
The authors would like to thank IBF Electronic and DOST-PCIEERD for the financial support used in thisstudy.

105

P1-68
Optical emission spectroscopy studies on the effect of negatively biased copper shield to low
pressure glow discharge plasmas
J. Daseco, K. Pabelina, C. Romero, J. Guhit, H. Ramos
Plasma Physics Laboratory, National Institute of Physics, University of the Philippines, Diliman,
Quezon City, Philippines
This paper investigates the effect of a negatively biased copper shield on the low-pressure DC
glow discharge oxygen, water and hydrogen peroxide plasmas generated using a Plasma Enhanced
Chemical Vapor Deposition (PECVD) facility. The negatively biased copper shield served as afocusing mechanism to influence the electric field intensities, affect plasma formation on thesystem and confine the active plasma species responsible for bacterial inactivation. Thismechanism can be utilized to improve plasma sterilization techniques on materials with deep
cavities and complex geometries. Optical emission spectroscopy was used to identify plasma
species present and examine the effect of the copper shield on spectral line intensities ofchemically active plasma species involved in sterilization processes.

106P1-70
Improved Charge Exchange Measurements in Large Helical Device
T.Ozaki, N.Tamura, S.Sudo, High-Energy Particle Group and LHD Experimental Group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
The pellet charge exchange technique (PCX), which is a combination of the compact neutral
particle analyzer (CNPA) and an impurity pellet, is a unique method to observe the radial
energetic particle distribution. The polystyren pellet of 1 mm-diameter is ablated after the injection
to the plasma and it occurs a charge exchange reaction between the plasma ion (mainly proton).
There are not only the charge exchange reaction between the hydrogen in pellet and the proton, butalso between the ionize carbon in pellet and the proton. We measure the energetic particle
enhancement at the resonance layer in the ion cyclotron resonance heating. However the current
system has a problem about the poor spatial resolution. According to the CCD imaging with filter,
the pellet cloud is expanded to 5 cm or more along the magnetic line. However the main core,
which consists of mainly neutral atoms, is still within 2 cm. This means that highly time resolved
measurement is meaningful. We prepare 10 times high repetition clock system to obtain the 10micro seconds resolution. If the pellet velocity is 500 m/s, the spatial resolution of 0.5 cm can beobtained.
Three-dimensional camera system is also introduced to observe the accurate pellet trajectory. At
asymmetric tangential injection of the neutral beam injection, the pellet trajectory is notstraightforward. More accurate spatial distribution of the energetic particle can be expected byusing the system.

107

P1-71
Concept for Numerical Calculation of 3D MHD Equilibria with Flow and FLR Effects
D.Raburn, A.Fukuyama
Kyoto University, Kyoto, Kyoto, 606-8501, Japan
Equilibrium flows and 3D effects can significantly impact plasma stability and energy
confinement. Further, in equilibria with flow, FLR effects can play an important role. Presently,
there exist a number of codes which can calculate MHD equilibria with a subset of the above
effects, such as: the FLOW code,[1] the PIES code,[2] and a recently developed code forcalculating axisymmetric equilibria with flow and hot ions in the large aspect-ratio limit.[3] Using
insights gained from these codes, the concept for a new code for calculation of 3D MHD equilibria
with flow and FLR effects is being being developed; the code is called the Kyoto ITerative
Equilibrium Solver (KITES).
References[1] L. Guazzotto, R. Betti, J. Manickam, and S. Kaye, Phys. Plasmas 11, 604 (2004).
[2] H. S. Greenside, A. H. Reiman and A. Salas, J. Comput. Phys., 81(1): 102-136, (1989).
[3] D. Raburn and A. Fukuyama, Phys. Plasmas 17, 122504 (2010).
This work is supported by Grant-in-Aid for Scientific Research (S) (20226017) from JSPS, Japan.

108P1-72MHD Equilibrium Analysis of SDC Plasmas in LHD
Y.Suzuki and N.Ohyabu
National Institute for Fusion Science, Toki, Gifu 509-5292, JapanThe MHD equilibrium is the basis of both most theoretical considerations and physics
interpretation of the experimental results. As a standard technique to calculate the 3D MHD
equilibrium, an inverse equilibrium solver VMEC, assuming the existence of perfect nested flux
surfaces, is widely used. In such a technique, a magnetic coordinate system is directly constructedso as to satisfy the force balance, or, MHD equilibrium equation. For low-β equilibrium, since themagnetic field sustains clear flux surfaces, the standard technique is acceptable. However, by
nature, 3D MHD equilibrium has magnetic islands and stochastic regions in the plasma because of
the absence of toroidal symmetry. For high-β equilibrium, the degradation of flux surfaces by thefinite β effect is not avoidable, so that the standard technique based on the nested flux surfaces
could not be directly applicable to them. On the other hand, in recent experiments, various types of
3D MHD equilibrium are obtained, namely, low-shear 3D MHD equilibrium with magnetic
islands, 3D MHD equilibrium with multiple magnetic axes, 3D MHD equilibrium with zerorotational transform, two-dimensional (2D) MHD equilibrium with current hole near the magnetic
axis. The standard technique based on the nested flux surfaces is not suitable in such situations. In
order to analyze such MHD equilibria with magnetic islands and stochastic magnetic field, othertechniques are required such as the HINT2 and PIES codes.
As mentioned above, the HINT2 code is a powerful and user-friendly code. In this study, we apply
the HINT2 code to study SDC (Super Dense Code) plasmas in the LHD. Since the SDC plasma
has the peaked pressure profile, the Shafranv shift is large. In addition, the weak magnetic shear
leads the stochastization in the peripheral region. We study the stocahstization due to finite β
effects.

109

P1-73
MHD equilibrium analysis with anisotropic pressure in LHD
Y.Asahi1), Y.Suzuki1,2), K.Watanabe1,2), W.A.Cooper3)
1)The Graduate University for Advanced Studies, Toki, Gifu 509-5292, Japan2)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan3)Ecole Polytechnique Federale de Lausanne, CRPP Lausannne, Switzerland
In the LHD (Large Helical Device) experiments, high beta plasmas are sustained only bytangentially-injected neutral beam (NB) lines under the low density and the low magnetic fieldstrength. Because of the long slowing-down time of the energetic ions by tangential NBs and thelarge beam pressure comparing with the thermal one, the velocity distribution of the plasmas is
expected to be anisotropic. From a Monte Carlo simulation, the beam pressure is estimated ～30%of the total pressure in a high beta plasma with ～5% volume averaged beta value[1]. The MHD
equilibrium theory predicts that anisotropic plasma pressures parallel and perpendicular to fieldlines are not flux surface quantities. The method to identify the distortion of pressure from thesurface average in experimental plasmas is not established. In this study, the effects of the
distortion on measurable values are investigated in LHD plasmas. The MHD equilibria with theanisotropic pressure are analyzed using a 3D MHD equilibrium code, ANIMEC[2]. We study the
position of magnetic axis[3] and the magnetic flux obtained by magnetic diagnostic as candidatesof characteristic parameter in the MHD equilibrium. As the results, we find that the position ofmagnetic axis is not sensitive to the distortion from the flux surface average, but the magnetic flux
is sensitive to the distortion. This suggests that the magnetic diagnostics has a possibility toestimate the distortion from flux surface average.
References[1] K. Y. Watanabe, et al. Fusion Science and Technology 58, 160 (2010).
[2] W. A. Cooper, et al. Comp. Phys. Comm. 180, 1524 (2009).
[3] Y. Asahi, et al. Plasma and Fusion Research Volume 6, 2403123 (2011).

110

P1-74
Numerical analysis of axisymmetric toroidal equilibria with flow in single-fluid and twofluid MHD models
A.Ito, N.Nakajima
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Numerical analysis of axisymmetric toroidal equilibria with flow is performed based on singlefluid and two-fluid magnetohydrodynamic (MHD) models. Effects of toroidal and poloidal flow
comparable to the poloidal sound velocity, two-fluid, ion finite Larmor radius (FLR), pressure
anisotropy and parallel heat fluxes on high-beta toroidal equilibrium is studied by solving reduced
MHD equilibrium equations. The equations to solve are the Grad-Shafranov (GS) type equilibrium
equations for high-beta tokamaks derived from the fluid moment equations for collisionless,magnetized plasmas [1]. The equations for the anisotropic pressures derived in the formalism
include the heat fluxes. The gyrovisosity and other FLR effects cause the so-called gyroviscouscancellation of the convection due to the ion diamagnetic flow induced by the two-fluid effects inthe equilibrium equations of momentum balance, pressure and heat fluxes. We have solved the
reduced GS equations numerically by means of the finite element method. The two-fluid effects
induce the diamagnetic flows, which result in asymmetry of the equilibria with respect to the signof the E×B flow. Higher order terms of quantities like the pressures and the stream functions show
the shift of their isosurfaces from the magnetic surfaces due to effects of flow, two-fluid andpressure anisotropy. We will also show the numerical solutions for the generalized GS equation
for non-reduced single-fluid MHD equilibria with flow and compare with the analytical solution of
the reduced GS equations in the single-fluid limit [2] to examine the validity of the reduced GS
equations.
References[1] A. Ito and N. Nakajima, submitted to Nucl. Fusion.
[2] A. Ito and N. Nakajima, Plasma Phys. Control. Fusion 51 (2009) 035007.

111

P1-75
MHD equilibrium including a static magnetic island for the reduced MHD equations instraight heliotron configuration
K. Saito1), K. Ichiguchi1,2), R. Ishizaki1.2)(1)The Graduate University for Advanced Studies, Toki, Gifu 509-5292, Japan(2)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
We study the interaction between static magnetic islands and interchange modes in straightheliotron configurations by using the reduced MHD equations[1]. In our previous studies for theinteraction[2,3], a pressure profile corresponding to nested magnetic surfaces was used as theequilibrium pressure profile. In the results, interchange modes grow as in the case without theislands. This is due to the fact that the finite pressure gradient exists in the region of the magnetic
islands. In general, the equilibrium pressure corresponding to the magnetic surfaces including
magnetic islands can be flat inside the magnetic islands. Therefore, the magnetic islands can affectthe stability of interchange modes through the local change of the pressure profile. In order tostudy island effect on the stability, an MHD equilibrium having the pressure profile consistent
with the magnetic islands is required.
In this work, we numerically calculate such an MHD equilibrium including a static magnetic
island for the reduced MHD equations in a cylindrical geometry by using a two-step approach. In
the first step, we obtain the plasma pressure which is constant along the field lines with themagnetic field fixed. In the second step, we solve the equilibrium equation corresponding to thevorticity equation with the pressure fixed. The two steps are iterated until the island width isconverged. We utilize three kinds of method for the first step. At first, we employ a diffusion
equation parallel to the field to obtain the pressure. As a result, the obtained pressure profile has
local flat structure at both the O-point and the X-point of the magnetic island. Next, we include thepressure diffusion perpendicular to the field in the first step. In this case, a pressure profile with aflat region only at the O-point, not the X-point, is obtained. Furthermore, we employ an averaging
method as a different approach in the first step. In this approach, the resultant pressure profile isalso locally flat only at the O-point.
References[1]K. Ichiguchi, N. Nakajima, M. Wakatani, B. A. Carreras and V. E. Lynch, Nucl. Fusion
43, (2003) 1101.
[2]K. Saito, K. Ichiguchi, and N. Ohyabu, Phys. Plasmas 17, (2010) 062504.
[3]K. Saito, K. Ichiguchi, and R. Ishizaki, Plasma Fusion Res. 6, (2011) 2403072.
This work is supported by budget NIFS11KNXN222 of the National Institute for Fusion Science and by a Grantin-Aid for Scientific Research (C) 22560822 from JSPS, Japan.

112

P1-76
Simulation of R-T and K-H hybrid instabilities in a 2D slab
R.Goto1), H.Miura1) 2), A.Ito1) 2), M.Sato2)
1) The Graduate University for Advanced Studies (Sokendai), 322-6 Oroshi, Toki, Gifu 509-5292,
JAPAN2) National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, JAPANLinear and nonlinear evolutions of Magnetohydrodynamic (MHD) unstable modes such as
interchange/ballooning modes can play key roles to understand experimental results in magnetic
confinement devices. While these unstable modes starting from an equilibrium without a flow
have been well studied (see [1,2], for example), it can often happen that the unstable modes arecoupled with other kinds of unstable modes, such as the Kelvin-Helmholtz instability, when a flow
is driven.
In this paper we study a simultaneous growth of the Rayleigh-Taylor and the Kelvin-Helmholtz
modes in a 2D slab. Numerical simulations of the compressible MHD equations in a 2D slab
geometry are carried out for various parameters to clarify various aspects of interactions between
the two kinds of modes. Influences of the wave numbers of the two modes, magnitudes of theinitial perturbations, the ratio of the mass density, the velocity difference on the growth rates of theunstable modes, evolution of the mixing range, and transition to turbulence are concerned.
We are also concerned about some two-fluid effects on the hybrid evolution since both of the two
types of the instability have the same nature such that the growth rate of the unstable linear eigenmode is larger for a larger wave number. Because of the nature, the wave length of the unstablemodes can be comparable to or smaller than the ion skin depth and/or the ion Larmor radius,
influences of which are not included in the single-fluid MHD equations. We will try to clarify a
few aspects of the two-fluid effects on the hybrid evolution of the two unstable modes.
References[1] H. Miura, N. Nakajima, T. Hayashi and M. Okamoto, Fusion Science and Technology 51(2007) 8.
[2] H. Miura and N. Nakajima, Nuclear Fusion 50 (2010) 054006.

113

P1-77Development of a portable AMR module applicable to various fluid/particle simulations
H. Miura1,2), M. Nunami1), H. Usui3), T. Moritaka3), T. Matsui3), R. Goto2)
1) National Instutite for Fusion Science, Toki, Gifu 509-5292, JAPAN
2) Graduate University for Advanced Studies (SOKENDAI), Toki, Gifu 509-5292, JAPAN3) Kobe University, 1-1 Rokko-dai Nada-ku, Kobe 657-8501, JAPAN
Enabling a high-resolution simulation is a prerequisite subject to study a complex dynamical
behavior of hot plasma. It is required not only to understand detailed physics of small-scale events
such as saturations of short-wave instabilities, micro-turbulence and magnetic reconnections, butalso to construct a macro-scale model to predict long-time behaviors of the plasma because smallscale events often affect the long-time behaviors through changing local pressure gradients and/or
magnetic configurations.
For the purpose of carrying out numerical simulations with a high resolution and a high efficiency,
we develop an adaptive mesh refinement (AMR) module which is easily transplanted to various
numerical codes, especiall in the Numerical Simulation Research Project (NSRP) of NIFS. While
the module is based on the block-based domain decomposition approach and the self-similarrefinement, it is designed to be applicable to a cell-based decomposition easily. Computational
domains generated as a consequence of the refinement are distributed to the computational nodes
according to the Morton method[1]. It is aimed that the ion skin depth, the ion Larmor radius andsome other scales are well resolved by intrducing the AMR module while macroscopic behaviors
are also simulated simultaneously.
In this paper, we demonstrate the effectiveness of the developed AMR module on fluid
simulations by carrying out magnetohydrodynamic (MHD) simulations. The MHD equations areapproximated either by the use of the 4th order central finite difference scheme or by the use of the
8th order compact finite difference scheme, which have been applied for the nonlinear MHD
simulations of LHD[2,3]. The suitability of the compact finite difference scheme to the AMR
approach will be discussed.
References[1] M. S. Warren, and J. K. Salmon, Proceedings of Supercomputing, IEEE, 12 (1993) 21.
[2] H. Miura, N. Nakajima, T. Hayashi and M. Okamoto, Fusion Science and Technology 51(2007) 8.
[3] H. Miura and N. Nakajima, Nuclear Fusion 50 (2010) 054006.

114

P1-78
Numerical MHD Analysis of LHD Plasma in Magnetic Axis Swing Operation
K.Ichiguchi, S.Sakakibara, S.Ohdachi, B.A.Carreras1)
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)BACV Solutions Inc. Oak Ridge, Tennessee 37831, USA
When we consider the steady state operation in a fusion reactor, we have to avoid disruptive
phenomena. Heliotron configurations including Large Helical Device have an advantage that they
are free from the disruptions which inherently occur in tokamaks. However, minor pressurecollapses have been observed even in the heliotrons even in the LHD configurations with thevacuum magnetic axis shifted inward. They are not as dangerous as the disruptions, however, theyshould be suppressed for the stable steady state operations. Recently, experiments with a real time
control of the vertical field was carried out for the precise study of the collapse property in the
inward shifted configurations[1]. In the experiments, the position of the corresponding vacuum
magnetic axis is shifted toward the inward side during each discharge. This operation is called amagnetic axis swing operation. Collapse in the profile of the electron temperature, therefore in thebeta profile, is observed when the corresponding vacuum magnetic axis becomes smaller thanabout Rax=3.55m.
In the present work, we simulate the experiment with the magnetic axis swing by means of thenonlinear MHD calculation. In the simulation, we take the change of the equilibrium quantities
due to the vertical field change into account. However, the time scale of the equilibrium change isquite slow compared with the time scale of the MHD dynamics of the perturbation. We have totreat both time scale simultaneously in the simulation. On the other hand, we have developed amulti-scale simulation scheme for nonlinear MHD simulations for beta-increasing heliotron
plasma[2]. In the scheme, we can follow the development of the equilibrium due to the increase ofbeta as well as the nonlinear dynamics of the interchange modes. Thus, we apply this multi-scale
scheme to the simulation of the magnetic axis swing experiment. This can be achieved byexchanging the equilibrium development due to the increase of beta to that due to the vertical field
variation. We discuss the mechanism of the collapse of the pressure and the stability limit in theview point of the nonlinear collapse based on the simulation results.
References[1] S.Sakakibara, et al., 23rd IAEA Fusion Energy Conference, 11-16 October 2010, Daejon,
Korea, Exps/P5-13.
[2] K.Ichiguchi, B.A.Carreras, Nucl. Fusion 51(2011) 053021 1-7.
This work is supported by a budget NIFS11KLTT001 of the National Institute for Fusion Science, and Grant-inAid for Scientific Research (C) 22560822 of JSPS Japan.

115

P1-79Dynamic Domain Decomposition for 3D PIC simulation with Adaptive Mesh Refinement
H. Usui2), Y. Yagi, T. Matsui2), M. Nunami1)2) and T. Moritaka2)
Graduate school of system informatics, Kobe University, Kobe 657-8501, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)JST/CREST
To investigate multi-scale phenomena in space plasma including plasma kinetic effects, we started
to develop a new electromagnetic Particle-In-Cell (PIC) code with Adaptive Mesh Refinement(AMR) technique. In AMR simulation, spatial grid size and time step intervals are definedaccording to the hierarchy levels, where high and low levels correspond to the fine and coarse grid
systems, respectively. As the simulation system evolves, some complex micro-scale phenomena
can locally and intermittently occur in a hierarchical domain (Level L). If the grid size in Level L
is too coarse to simulate the local complex phenomena, A higher hierarchical domain (Level L+1)
is adaptively created in which the grid spacing size and the time step interval become half of thoseused in the domain of (Level L). AMR-PIC simulations require smaller amount of memory andshorter computation time than conventional PIC simulations.
In parallelizing the code for a distributed many-core system, we adopt a scheme of dynamic
domain decomposition which realizes the load balance between distributed sub-domains. In
conventional domain decomposition scheme, one of the issues is load balance between processors
because the number of spatial grids and particles belonging to each sub-domain is not always
constant due to the AMR procedure. Since the cost of particle calculation is dominant, which isabout 70-80% of the total cost, the number of particle calculation loops should be balanced
between each processor. For this purpose we introduced dynamic domain decomposition called
DDD. We are currently incorporating this scheme in the three dimensional version of AMR-PIC
code and are performing a test simulation to examine the parallelization efficiency as well as thescalability with respect to the number of processors. In the presentation, we will show you thedetail of DDD scheme as well as some preliminary results of AMR-PIC

116

P1-80Nonlinear stability of externally driven magnetic islands
S. Nishimura, M. Yagi1,2), S. Toda, Y. Narushima
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1) Japan Atomic Energy Agency, Rokkasho, Aomori, 039-3212, Japan
2) Research Institute for Fusion Science, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan
Physics of resonant magnetic perturbations (RMPs) in torus plasmas is of great interest to themagnetic confinement fusion, since it might be applicable to the control method of themagnetohydrodynamic (MHD) instability. The RMPs due mainly to external current coils produce
current sheets at corresponding resonant magnetic surfaces, and under certain condition they drive
the forced magnetic reconnection and generate magnetic islands, even in the absence of the
intrinsic tearing instability. In the Large Helical Device, which is a stellarator, sudden transition
between equilibria with and without the externally driven magnetic islands is observed[1]. It isalso observed that such bifurcation affects the amplitude of the pressure-gradient-driven MHD
instability[2].
The goal of the present study is theoretical understanding of the influence of the RMPs on
stellarator plasmas. For this purpose, the following approaches are examined.
Firstly, a theoretical model of the forced magnetic reconnection in a stellarator is developed,
which is an extended version of that in our previous work[3]. In the model, zero-dimensional
equations of the width and the phase angle of magnetic islands are coupled with an onedimensional equation of the neoclassical poloidal flow produced by the rippled toroidal magnetic
field. It is found that nonlinear saturation state of the externally driven magnetic islands has two
branches, i.e., non-rotating, small magnetic islands and non-rotating large magnetic islands, andthe transition and the hysteresis between these states are observed. These phenomena are causedby the screening effect of the poloidal flow and the slip of the magnetic island rotation in the rest
frame of rotating plasmas.
In the next step, the three-dimensional nonlinear dynamics of the externally driven magnetic
islands is simulated, using a reduced set of two-fluid equations for a stellarator. The nonlinear
bifurcation state of the externally driven magnetic islands is also observed in the simulation, andthe results of the theoretical model are confirmed. Influence of the RMPs on the pressure-gradientdriven MHD modes is also discussed.
References[1] Y. Narushima et al., Nucl. Fusion, 48 (2008) 075010.
[2] F. Watanabe et al., Nucl. Fusion, 48 (2008) 024010.
[3] S. Nishimura et al., Plasma and Fusion Research, 5 (2010) 040.

117

P1-81
Effect of resistivity on mode structure of interchange instability in heliotron plasma
R. Ueda, Y. Matsumoto, M. Itagaki, S. Oikawa, K. Y. Watanabe1), M. Sato1), W. Takado
Graduate School of Engineering, Hokkaido University, Sapporo 060-8628, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
It is one of the important subjects for realization of the fusion reactor to investigate theMagnetohydrodynamics (MHD) instabilities. The interchange mode which is one of the pressuredriven instabilities is considered to play an important role in plasma confinement property in the
heliotron plasma.
In the Large Helical Device (LHD), the magnetic fluctuations around the peripheral rationalsurface are observed. The level of such fluctuations increases with decreasing magnetic Reynolds
number [1]. It is important for understanding the interchange instabilities to investigate
relationships between the magnetic islands produced by the magnetic fluctuations and the mode
structures of instabilities. However, such relationships have not been investigated especially forcases with the high magnetic Reynolds number.
In this study, we numerically analyze the interchange instability of the straight heliotron plasma
with various magnetic Reynolds number based on the linearized reduced MHD equations. The
relationships between the width of the magnetic island and the mode structure will be presented.
We focus on interchange instability with (m, n) = (1, 1) mode. (m is the poloidal mode number
and n is the toroidal mode number.) In addition, the dependencies on the magnetic Reynolds
number and the β value will be discussed.
References[1] K. Y. Watanabe, et al., Physics of Plasmas 18, 056119(2011).

118

P1-82Development of an MHD Code Based on the CIP Method
to Consider the Advection with Alfvén Velocity
W. Takado, Y. Matsumoto, R. Ueda, M. Itagaki and S. Oikawa
Graduate School of Engineering, Hokkaido University, Sapporo 060-8628, JapanThe analysis of Magnetohydrodynamics (MHD) instabilities is one of the important issues torealize
the nuclear fusion reactor. The advection of the Alfvén wave must be solved with small numericalerrors in the simulations of the instabilities since the fluctuations propagate at the Alfvén velocity.
The advection term described with the Alfvén velocity explicitly appears in the MHD equations
when Elsässer variables are introduced [1]. It is well known that the Constrained Interpolation
Profi le (CIP) method [2] is useful for stably and accurately solving the advection equation.
In this study, we have developed an MHD code in which the MHD equations using Elsässer
variables are solved with the CIP method. Using the developed code, it is investigated whether the
initial equilibrium of tokamak plasmas is maintained. We first apply the developed code to ahypothetical rectangular cross sectional tokamak plasma. The developed code is also applied to the
JT-60U plasma.
The detail of the developed code will be described. The time evolutions of several quantities willbepresented. The comparison will be made between our results and those by use of the CIP-MOCCT
method [3].
References[1] W. M. Elsässer, Phys. Rev. 79, 183 (1950).
[2] H. Takewaki, A. Nishiguchi and T. Yabe, J. Comput. Phys. 61, 261 (1985).
[3] T. Kudoh, R. Matsumoto and K. Shibata, Astrophys. J. 508, 186 (1998).

119

P1-83Modeling of Formation of Helical Structures in Reversed-Field Pinch
N.Mizuguchi1)2), A.Sanpei3), S.Fujita3), K.Oki3), H.Himura3), S.Masamune3), K.Ichiguchi1)2)
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)The Graduate University for Advanced Studies (Sokendai), Toki, Gifu 509-5292, Japan3)Kyoto Institute of Technology, Kyoto, Kyoto 606-8585, Japan
From a point of view of the magnetohydrodynamic (MHD) stability, spontaneous formation, or
self-organization, of a quasi-steady state is a promising nature of plasma for a long-termsustainment of the magnetic confinement. In this study, we aim at modeling the formation of three-dimensional structures in the reversed-field pinch (RFP) plasma, which is experimentallyobserved with several types of aspect, by means of a nonlinear MHD simulation. In RFPs, thetearing modes tend to be destabilized because of the internal current by nature. The confinement istherefore deteriorated by the overlaps of multiple rational surfaces. In order to avoid this situation,a control method is proposed both experimentally and theoretically, in which the perturbations areconcentrated at a small number of the modes, and kept to quasi-steady states termed the quasisingle helicity (QSH) and the single helical axis (SHAx) states. We apply the general-purpose
nonlinear MHD codes, MIPS, to the RFP configuration with the reconstructed equilibria of the
RELAX device solved by the RELAXFit code[1] as the initial condition. The long-term solutionsof the simulation reproduce the growth of the resistive modes dominated by the m/n=1/4
component and the helical deformation of the whole torus. There is also a qualitative agreement in
the dependency on the equilibrium parameters to the concentration to the single n=4 mode. We
also discuss the transition processes between the single- and multi-helicity states.
References[1]Sanpei A., et al., J. Phys. Soc. Jpn., 78 (2009) 013501

120

P1-84
Study of resistivity effect on MHD stability beta limit in LHD using TASK3D-code
M.Sato, K.Y.Watanabe, Y.Nakamura1), A.Fukuyama2), S.Murakami2), R.Seki, M.Yokoyama,
H.Funaba, S.Sakakibara, S.Ohdachi, H.Yamada and N.Nakajima
National Institute for Fusion Science, Toki 509-5292, Japan
1)Graduate School of Energy Science, Kyoto University, Kyoto 611-0011, Japan2)Graduate School of Engineering, Kyoto University, Kyoto 606-8501, JapanThe MHD Stability beta limit in the Large Helical Device (LHD) has been analyzed using ahierarchy-integrated simulation code TASK3D[1] in order to explore the capability of the LHD
configuration. For the analysis of the MHD stability beta limit, the MHD equilibrium module
VMEC, the transport module TR and the linear MHD stability module MSSH are used. The
numerical model for the effect of the MHD instabilities is introduced such that the pressure profile
is flattened around the rational surface due to the MHD instabilities. The width of the flattening of
the pressure gradient is determined form the width of the eigenmode structure of the MHD
instabilities. It is also assumed that there is an upper limit of the poloidal mode number of the
MHD instabilities; mc which directly affect the pressure gradient. Recent experimental results
suggest that eigenmodes of the MHD instabilities with poloidal mode number of m<5 dynamicallyaffect the pressure profile. In our analysis, mc=4 is used as the standard mode of the calculation of
the achievable beta value. In the previous study[2,3], ideal interchange modes, which are mostimportant instabilities in helical plasmas, were considered to give the MHD stability beta limit. It
has been found that the achievable volume averaged beta value is expected to be beyond about 6%
for 1.5 < σ < 1.6, where σ is the peaking factor of the pressure profile. In this study, the linear
MHD stability module MSSH has been extended to analyze resistive modes and the effect of theresistive interchange modes on MHD stability beta limit is investigated. Dependence of theachievable beta value on the plasma resistivity will be discussed.
References[1] M.Sato et al., Plasma Fusion Res. 3 (2008) S1063.
[2] M.Sato et al., Proc. of 22nd IAEA Fusion Energy Conf. (Geneva, Switzerland,2008)IAEA-CN-165/TH/P9-18.
[3] M.Sato et al., Contrib. Plasma Phys. 50 (2010) 665.

121

P1-85Development and experimental application of integrated transport code, TASK3D, forhelical plasmas, ~equipment and extension of neoclassical transport module~
M.Yokoyama, A.Wakasa1), S.Murakami1), M.Sato, R.Seki, H.Funaba, K.Watanabe,
Y.Nakamura2), N.Nakajima, A.Fukuyama1) and LHD Experiment Group
National Institute for Fusion Science, Toki 509-5292, Japan
1)Graduate School of Nuclear Engineering, Kyoto University, Kyoto 606-8501
2)Graduate School of Energy Science, Kyoto University, Uji 611-0011
The integrated code for helical plasmas, TASK3D, has been developed both by modifying
modules in TASK [1] to e applicable to three-dimensional magnetic configurations, and by adding
new modules for stellarator-heliotron specific physics. In this paper, these module developments
so far are introduced, and those experimental applications are collectively described.
The Large Helical Device (LHD) experiments have steadily expand the parameter regime ofhelical plasmas such as electron and ion temperatures, beta value, electron density, pulse lengthand total input energy to plasma in a long-pulse operation as summarized in Ref.[2]. The widerange applicability of the integrated code is mandatory, and, in the mean time, experimental
verification of each module and its integration should be performed by utilizing the experiment
database.
The radial transport aspect of TASK3D employs the diffusive transport equation [TR] with thelocal diffusivity (particle and heat) and the particle and heat source/sink term. Taking threedimensional configuration’s impact into account such quantities have been main focus of module
modification/supplement in TASK3D development.
The presence of the ripple transport (so called 1/ν diffusion) is one of specific features in nonsymmetric magnetic configurations. The ambipolarity of neoclassical (NC) particle fluxes has
been demonstrated to well describe the bifurcation nature of the radial electric field (Er) in LHD
and other helical plasmas [3,4]. Thus, it is of a great importance to accurately calculate theneoclassical fluxes, and then Er. For this purpose, the diffusion coefficient database, DGN/LHD[5], has been prepared for a wide-range of equilibria by utilizing advantages of NC transport
codes, DCOM [6] and GSRAKE [7]. This highly-accurate NC module has made it possible tocompare with the power-balance analysis (TR-snap package, composed of several modules of
TASK3D) [8]. Recently, DGN/LHD database has been extended to higher-beta range, so that
accurate NC transport calculations for LHD high-beta plasmas and predictive analyses forparameter regime in a reactor such as FFHR [9] can be performed. Such example calculations will
be reported in the paper.
References[1] A.Fukuyama, http://bpsi.nucleng.kyoto-u.ac.jp/task/[2] A.Komori, H.Yamada et al., Fusion Science and Technology 58 (2010) 1[3] K.Ida et al., Nucl. Fusion 45 (2005) 391.
[4] M.Yokoyama et al., Nucl. Fusion 47 (2006) 1213.
[5] A.Wakasa et al., Contributions to Plasma Phys. 50 (2010) 582.
[6] A.Wakasa et al., J. Plasma Fusion Res. 4 (2001) 408.
[7] C.D.Beidler et al., Plasma Phys. Control. Fusion 36 (1994) 317.
[8] A.Wakasa et al., 23rd IAEA-FEC (2010).
[9] N.Yanagi et al., in this conference.
This work is performed with the support and under the auspices of the NIFS Collaboration Research program(NIFS11KNTT008).

122P1-86
The radial electric field formation in high Te plasmas with the electron internal transportbarrier in LHD
S. Matsuoka, S. Satake, H. Takahashi, A. Wakasa1), M. Yokoyama, T. Shimozuma,
T. Ido, A. Shimizu, LHD Experiment Group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan
Very high electron temperature (Te &simeq; 15 keV) plasmas are successfully obtained with the
ECH heating in recent LHD experiments [1]. These plasmas are called Core Electron-Root
Confinement (CERC) plasmas [2] since they are accompanied by the strong positive (electronroot) radial electric field (Er) in the core region. CERC plasmas also have the characteristic steep
Te gradient (electron internal transport barrier, eITB) at the core. It is also observed that theelectron-root Er has the steep shear there corresponding to the eITB formation. The finite orbit
width (FOW) effect of electrons becomes important in such high Te helical plasmas [3]. As theelectron-root Er and the steep Te gradient are formed in the core of CERC plasmas, the width ofthe local flattening, namely, the region of approximately zero temperature gradient, becomesnarrower. Since Er in helical plasmas are predominantly determined by the ambipolar condition of
the neoclassical (NC) transport, the numerical evaluation of the NC transport and the ambipolar Er
is the key issue to understand the relation between the improvement of the transport and Er in
CERC plasmas.
In this study, we focus on the NC ambipolar Er formation in the core region of CERC plasmas.
Since the poloidal flow velocity is deeply related to Er through the E × B drift and it affects the NC
transport, the radial profile of the poloidal flow is examined. This enables ones to investigate thereduction processes of the transport by Er and its relation to the poloidal flow in CERC plasmas.
For this purpose, NC transport analyses for CERC plasmas are carried out using δf Monte-Carloparticle code, FORTEC-3D [4], which has been extended to be applicable to electron species andcan calculate the NC transport including the electron FOW effect. The numerically obtained
ambipolar Er is compared to the experimental observations. The formation of the Er and its shear
in the core region, where either the local flattening or the steep gradient of Te is observed, are also
investigated. It is also investigated how the particle orbit affects the NC transport improvement in
the core.
References[1] H. Takahashi, et al., IAEA Fus. Energy Conf. (2010) pp.EXC/p8-15[2] e.g., M. Yokoyama, et al., Fus. Sci. Tech., 50 (2006) 327[3] S. Matsuoka, et al., Phys. Plasmas, 18 (2011) 032511[4] S. Satake, et al., Plasma Fus. Res., 1 (2006) 002

123

P1-87
Dependence of thermal diffusivity on plasma parameters in perturbed magnetic field in
toroidal plasma
R.Kanno1,2), M.Nunami1), S.Satake1,2)1)National Institute for Fusion Science, Toki 509-5292, Japan2)Department of Fusion Science, the Graduate University for Advanced Studies, Toki 509-5292,
Japan
We consider the modeling of energy transport under effect of noise caused by resonant magnetic
perturbations (RMPs) in a toroidal plasma. It is found in the previous study [1] that the radial
thermal diffusivity χr in the perturbed region is represented as χr=χr(0)(1+c|δBr/Bt|2), where |δBr| isthe strength of RMPs, |Bt| is the strength of toroidal magnetic field at the magnetic axis, χr(0) isthermal diffusivity when |δBr|=0, and c is a coefficient. In the present paper, dependence of thecoefficient c on plasma parameters is investigated by a Monte-Carlo simulation code calculatingtime evolution of an ion guiding center distribution function, where the guiding center distributionfunction evolves with time from the Maxwell distribution under effects of both the Coulomb
collision and noise caused by RMPs. Recently, we develop a transport simulation code without theassumption of nested flux surfaces [2]; the code is named KEATS. The code is programmed byexpanding the well-known Monte-Carlo particle simulation scheme based on the δf method. By
using KEATS code, it is possible to execute the investigation. We find that the coefficient c
depends on the collision frequency ν in wide range of collisionality and on the particle mass m.
References[1] R. Kanno et al., Plasma Phys. Control. Fusion, 52 (2010) 115004.
[2] M. Nunami et al., Research Report NIFS Series, No.NIFS-871 (2007).
This work was performed with the support and under the auspices of the NIFS Collaborative Research Program
NIFS11KNST019.

124

P1-88
Research on burn control of core plasma with the transport code
Y.MIYOSHI, Y.OGAWA1), M.NAKAMURA2)
1)Graduate School ofFrontier Science, 5-1-5 Kashiwanoha, Kashiwa-shi, Chiba 277-8568, Japan2)Fusion Reactor Design Group, Devision of Advanced Plasma Research, Naka Fusion Institute,Japan Atomic Energy Agency, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho, Kamikita, Aomori
039-3212, JapanFor the commercial fusion reactor, one will be required to control the reactor parameter, likefusion power Pfus, diverter heat flux qdiv, and avoid the disruption. One can’t satisfy these requests
directly, so one need to control some mid parameters like plasma shape (κ,δ), position, density n,current density j, temperature T, HH-factor, troyon factor βN and etc. To control these parameters,
many diagnostics and actuators are needed. In the future reactor, one must take COE (Cost Of
Electricity) or large fusion neutron flux into account and the diagnostics and actuators which is
extrapolated into the reactor are limited. In addition, the actuators, like gas puff, impurity
injection, NBI injection, RF, external coil flux, and etc, and plasma parameters are not necessarily
one-to-one correspondence. For these reasons, construction of control theory is required for the
future reactors. The control of plasma shape or position are discussed in [1], then, in this research,
we will consider the control of the core plasma and the control theory is expressed as follows,
Au(t)=x(t)y(t)=Bx(t)where x(t),u(t),y(t) are reactor parameter vector, actuator vector, mid parameter vector
respectively, and A, B are tensors. The choice of the components of x(t), y(t), u(t) are very
important and difficult problem. In this research, as the first step of the construction of control
theory, we use the simple model.1) We assume B tensor as a unite tensor, in other words, we canmeasure the plasma parameters accurately.2) We assume the dimension of u(t) and x(t) are 2, andA tensor is a 2×2 tensor.3) We assume the components of u(t) are Ppuff and PLH, and x(t) are Pfus
and qmin. In most of previous studies, they control one plasma parameter, and the study of multiple
parameters control with multiple actuators has not been much done. For this reason, we use two two model. Pfus is the most important parameter to make electricity, and qmin has strong influenceon high-β. The analysis of safty factor q control is done in [2-3]. In this research we will analyze
the A tensor with 1-D transport code, and in this conference, we will present about the control
simulation.
References[1]Y.MIYOSHI et al, PFR Vol6, 2405110 (2011)[2]T.SUZUKI et al, Nucl. Fusion, 48, 0450002 (9pp) (2008)[3]D.MOREAU et al, Nucl. Fusion, 43 870-882 (2003)

125P1-89The neoclassical transport and viscosity in helical plasma consisting of multi-species ions
including high-Z impurities
Kenji Nishioka1), Yuji Nakamura1) and Shin Nishimura2)

1) Graduate School of Energy Science, Kyoto University, Kyoto, Japan
2) National Institute for Fusion Science, Gifu, JapanThe neoclassical transport has important roles on the transport phenomena in helical plasma, suchas the radial electric field, the poloidal and toroidal viscosities, particle flows, bootstrap current. In
addition, this transport becomes significantly large at high temperature. Though high-Z ions is
considered to have greater influence on the neoclassical transport and viscosity, the effects ofmultispecies ions on the neoclassical transport in helical plasmas have not been well known yet,.
Therefore in this paper, by using the moment method, those effects on the neoclassical transport
and the neoclassical viscosity in helical plasmas are investigated.

126

P1-90Electron fluid perturbation and related electron transport in a field-reversed configuration
Fusaki P. IIZIMA, Toshiki TAKAHASHI and Tomohiko ASAI1)
Department of Electronic Engineering, Gunma University, Gunma 376-8515, Japan
1) College of Science and Technology, Nihon University, Tokyo 101-8308, JapanThe transport mechanism of a field-reversed configuration (FRC) plasma is still unclear.
Resistivity is anomalously high, and a trapped poloidal flux soon decays in less than 100 microseconds. The rapid flux decay may cause toroidal spin-up of an FRC. We have proposed that
direct conversion from the magnetic flux to the kinetic angular momentum takes place as long as
axisymmetry holds. The resultant rotational instability with the toroidal mode number n=2 of an
FRC plasma is known as the most often observed global instability. The FRC current just after
formation is primarily carried by electron current, while ions are approximately at rest. The ions,however, gradually gain angular momentum in the ion diamagnetic direction before the onset of
the instability. Conversion to the angular momentum, however, also occurs for electrons, and it
plays the electron current drive and consequently maintains the magnetic flux.
Therefore, we suppose that the presence of an anomalous loss of the electron angular momentum
due to an electron fluid perturbation with higher frequency than the ion cyclotron frequency.
Spatial structure of the perturbation in an FRC is studied analytically. Calculation of electron
orbits in the prescribed fields is carried out in order to find the loss of the angular momentum, i.e.,
the resistivity and the end-loss rate.

127

P1-91
Computation of Neutral Gas Flow Generation from a CT Neutralization Fuel-Injector
T. Watanabe, T. Ohshima, T. Takahashi, N. Fukumoto1), M. Nagata1)
Gunma University, Kiryu, Gunma 376-8515, Japan
1)University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280, Japan“CT Neutralization Fuel Injection” is an expansion of the “CT Injection” method.1,3 The “CT
Injection” method was proposed as a fuel-injection method for large-scale power generation
plasmas such as the ITER. However, feasibility is still unclear because of the difficulty concerning
the translation process of a plasmoid across the vertical magnetic field of the fuel injection target.
To resolve this problem, the “CT Neutralization Fuel Injection” method utilizes a neutralization
cell which transforms the injection plasmoid into an ultra-fast neutral gas flow. It is expected to befaster than conventional methods such as the “Gas-Puff” or “Pellet Injection”. Therefore, there is ahigh possibility of reaching the core of the fuel injection target.2,4
In this study, we will check if the neutral gas generation is enough to meet above quotas. We run acomputer simulation on the neutralization cell of the “CT Neutralization Fuel Injection” device. In
our calculations, we focus on the plasmoid itself moving through the cell. The plasmoid is treatedas a stationary target upon which the neutral gas particles of the cell collide and charge exchange
with. Therefore only gas pressure inside the cell and injection speed of the plasmoid is changed.
The collision processes are reproduced by the Monte Carlo calculation. The calculation determines
the plasmoid’s degree of neutralization, and as a result optimum cell length is found.
When the central ion temperature is 100 [eV] and the injection speed of CT at 50-300 [km/s], it
was determined that 0.4 [m] is the optimum cell length due to thermal diffusion of the generated
gas flow. Any cell longer will be not efficient, owing to the increased chance of the neutral gas
particles colliding with the cell and translation tube walls. This will likely cause a reduction of thegas injection speed. Furthermore, it was calculated that the gas pressure inside the neutralization
cell and the plasmoid’s injection speed does not largely affect the neutralization level and process.
Also, on the subject of interactions with the cell wall, we have determined that the incident angle
for most of the impacts is 30 to 60 degrees.
We also show preliminary results of a hybrid simulation on CT plasma. The hybrid simulation
model under development will take into account the neutralization process to investigate anelectromagnetic behavior inside the neutralization cell. In particular, it should be clarified thatmagnetic flux decay occurs by reducing the plasma current and associated radial expansion of theplasma caused by imbalance of the radial force. We will also study the influence of the axial
electric field generated by the friction force between electrons in the moving plasmoid and coldions.
References[1]L. J. PERKINS et al., Nucl. Fusion 28, 1365 (1988)[2]S. Fujii et al., IEEE Transactions on Plasma Science 38, 1473 (2010)[3]R. Raman et al., Phys. Rev. Lett. 73, 3101 (1996)[4]M. Nagata et al., Nucl. Fusion 45, 5 (2005)

128

P1-92
Toroidal spin-up caused by an anomalous loss of the electron angular momentum of a fieldreversed configuration
T. Takahashi, F.P. Iizima, T. Watanabe, T. Asai1)
Department of Electronic Engineering, Gunma University, Gunma 376-8515, Japan
1) College of Science and Technology, Nihon University, Tokyo 101-8308, JapanThe rotational instability with the toroidal mode number n=2 of a field-reversed configuration(FRC) plasma is the most often observed global instability. The current of FRC plasmas just after
formation is primarily carried by electrons, while ions are approximately at rest. The ions,however, gradually gain the angular momentum in the ion diamagnetic direction before the onsetof the instability. The rotation mechanism of the FRC plasma has been often explained by aselective loss of ions [1], or end-shorting [2].
We have proposed another possible spin-up mechanism, which is direct conversion from themagnetic flux to the kinetic angular momentum as long as axisymmetry holds [3]. In this spin-up
mechanism, plasma ions in the core of the FRC plasma can gain the angular momentum.
Therefore, plasma rotation is possible to start near the field-null; the core rotation start-up is good
agreement with experiments. Conversion to the angular momentum, however, also occurs forelectrons, and a rapid loss of the electron angular momentum is needed to explain a rapid loss ofthe magnetic flux.
Therefore, we suppose that the presence of an anomalous loss of the electron angular momentum
due to an electron fluid fluctuation with higher frequency than the ion cyclotron frequency. Hybrid
simulation is carried out to find the time evolution and the spatial profile of the toroidal flow
velocity. Comparison of the calculation results with experimental results is also
References[1] D. S. Harned, D. W. Hewett, Nucl. Fusion 24 (1984) 201.
[2] L. C.Steinhauer, Phys. Plasmas 9 (2002) 3851.
[3] T.Takahashi et al., Plasma Fusion Res. 2 (2007) 008.

129

P1-93Nonthermal effects on the resonant instability of the dust-acoustic wave in semi-bounded
Lorentzian plasmas
Dae-Han Ki, Young-Dae Jung
Department of Applied Physics, Hanyang University, Ansan, Kyunggi-Do 426-791, South Korea
The nonthermal effects on the resonant instability of the surface dust-acoustic wave areinvestigated in semi-bounded Lorentzian dusty plasmas containing elongated rotating dust grains.
It is found that the nonthermal effects suppress the frequency domain and enhance the growth rateof the resonant instability for the case of positively charged dust grains. For negatively charged
dust grains, however, the nonthermal effects on the frequency domain and the growth rate arefound to be negligible. In addition, it is found that the growth rates of the resonant instabilities innonthermal plasmas would be greater than those in thermal plasmas.

130P1-94
Numerical Analysis of Quantum-Mechanical Grad-B Drift II
P. K. Chan, S. Oikawa and E. Okubo
Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Japan.
We have developed a code to solve the time-dependent Schrödinger equation [1], [2], withstandard notations, iħ (∂ψ / ∂t) = (1 / 2m) (−iħ ∇ − qA)2ψ, in the presence of a non-uniform
magnetic.
In the previous paper [2], we have shown that the quantum mechanical variance σ2 in position may
reach the square of the interparticle separation in a time interval of the order of 10−4 sec for typicalmagnetically confined fusion plasmas.
In this paper, as an extension of the paper [2], we investigated the dependence of the variance σ2 inposition on parameters such as m, q, v0, B0 and LB, where m is the mass of the particle, q is the
charge, v0 is the initial speed of the corresponding classical particle, B0 is the magnetic field at theorigin and LB is the gradient scale length of the magnetic field. Here we have assumed a vector
potential A = (− B0 y (1 − y / 2LB ), 0, 0), which corresponds to the non-uniform magnetic field of
B = (0, 0, B0 (1 − y / LB )).
We solved the two dimensional time-dependent Schrödinger equation for a single particle in thepresence of the non-uniform magnetic field for various sets of m, q, v0, B0 and LB. It wasnumerically found that the variance, or the uncertaintly, in position σ 2 is given by, dσr2 / dt =4.04ħv0 / qB0LB.
References[1] S. Oikawa, T. Oiwa, and T. Shimazaki, Plasma Fusion Res. 5 (2010) S2025.
[2] S. Oikawa, T. Shimazaki and E. Okubo, Plasma Fusion Res. 6 (2011) 2401058.

131

P1-95
Quantum Mechanical Plasma Scattering II
E. Okubo, S. Oikawa, P. K. Chan
Graduate School of Engineering, Hokkaido University, N-13, W-8, Sapporo 060-8628, Japan
We have solved the unsteady Schrödinger equation in order to analyze a real particle's motion [1,2]. We will numerically solve the two-dimensional Schrödinger equation for a single particle in
the presence of a uniform magnetic field and field particles.
The unsteady Schrödinger equation for wavefunction ψ(r,t), at a position r and a time t, is given by2iħ∂ψ/∂t=[(-iħ∇-qA) /(2m)+qφ]ψ,where φ and A stands for the scalar and vector potentials, m and q the mass and electric charge ofthe particle under consideration, i≡(-1)1/2 the imaginary unit, and ħ≡h/2π the reduced Planck
constant. When the corresponding classical particle has an initial momentum p0=mυ0 at a position
r=r0, the initial condition for the wavefunction is given byψ(r,0)=exp[-(r-r0)2/(2lB2)+ik0·r]/(π1/2lB),
where r0 is the initial center of ψ, lB≡(ħ/qB)1/2 is the magnetic length, and k0=mυ0/ħ is the initial
wavenumber vector.
Let the vector potential be
A=(-Bzy,0,0),
and the scalar potential be, in the cylindrical coordinates (R,θ,Z),∞2
2
2
φ(η)=Σfφf(η)=Σfqf/(π ε0lB)∫0 K[4ηη′/(η+η′) ]exp(-η′ )η′/(η+η′)dη′,
where f is the label for field particles, ε0 is the vacuum permittivity, K is the complete elliptic
integral of the first kind, and η=R/lB.
References[1] S. Oikawa, T. Oiwa, and T. Shimazaki, Plasma Fusion Res. 5, S2024 (2010).
[2] S. Oikawa, T. Shimazaki, and E. Okubo, Plasma Fusion Res. 6, 2401058 (2011).

132

P1-96
Theoretical study of ultra-relativistic laser electron interaction with radiation reaction by
quantum description
K. Seto, H. Nagatomo, J. Koga1), K. Mima
Institute of Laser Engineering, Osaka University, 2-6 Yamada-oka, Suita, Osaka 565-0871, Japan
1)Advance Photon Research Center, Kansai Research Establishment, Japan Atomic Energy
Research Institute, Kyoto 619-0215, Japan
With the advent of ultra?relativistic laser, it will reach to the Intensity 1022W/cm2. If an electron isunder this laser when laser intensity is larger than 1018W/cm2, an electron will carry out relativistic
behavior. The most important phenomenon in this regime is an effect of ponderomotive force, andan electrons are pushed out in the propagation direction of laser. If we can get the laser intensity
1022W/cm2, the strong bremsstrahlung might be caused. Moreover, simultaneously with it,“radiation reaction force (or damping force)” works to charge particles. We could reach todescribe this reaction force working on an electron in Minkowski spacetime, it is as follows[1]:freactionμ=m0τ0 d2wμ/dτ2 + m0τ0/c2 g(dw/dτ,dw/dτ)wμ - m0τ0/c2 (w0wμ-c2δμ0)/((w0)2-c2) d2w0/dτ2
Where τ0=Q2/6πε0m0c3, w, τ and g is the relativistic 4-velocity, proper time and Lorentz metric.
Before this equation appears, the Landau-Lifshitz (L-L) equation[2] as the approximation ofLorentz-Abraham-Dirac (LAD) equation[3] is used. But, the L-L equation breaks the relation ofrelativistic covariance. Our equation is satisfied with the covariance.
In the next laser intensity generation (over 1024W/cm2), we consider the electron-positron paircreations and something effects appear[4]. This intensity level requires us to treat phenomena as
QED. However, we can consider an electron as quantum description under 1024W/cm2, without
the pair creations. In this presentation, we suggest the quantum model of the radiation reaction
without the electron-positron pair creation using the Dirac equation.
[iħγμ(∂μ-ie/ħAμ)+Pradiation-mcI4×4]ψ=0
Here, Aμ means the external laser driver field. The radiation reaction is one of the self interactions.
Therefore, we put the new interaction term of Pradiation. In our model, the bremsstrahlung-photon
field is treated as quantum field.
References[1] K. Seto, H. Nagatomo and K. Mima, Plasma Fusion Res. 6, 2404099 (2011).
[2] L. D. Landau and E. M. Lifshitz, The Classical theory of fields (Pergamon, New York,
1994).
[3] P. A. M. Dirac, Proc. Roy Soc. A 167, 148 (1938).
[4] G.A.Mourou, C.P.J.Barry and M.D.Perry, Phys. Today 51,22 (1998).

133

P1-97
Laser Guiding Through an Axially Nonuniform Magnetoplasma Channel in the Weakly
Relativistic Limit
Navpreet Singh and Arvinder SinghNational Institute of Technology, Jalandhar, Punjab, India
In this paper, laser guiding through an axially nonuniform plasma channel is investigated in the
weakly relativistic limit, under the effect of static magnetic field. Self defocusing of the prepulse
leads to the axial nonuniformity of the plasma channel. The propagation of the guided laser beamthrough the preformed axially nonuniform plasma channel results into the alternate convergenceand divergence of the beam. Unbalanced diffraction and refraction phenomenon through magnetoplasma channel results into periodic beam width variation with the distance of propagation. The
wave equations governing the propagation characteristics of the ionizing prepulse and delayed
pulse through an axially non-uniform plasma channel have been solved by moment theory
approach. Effect of the Magnetic field, Beam width of the guided laser beam, intensity of theguided laser beam and the axial nonuniformity of the plasma channel on the laser guiding has beenstudied. Laser guiding up to several Rayleigh lengths has been observed. We have seen a very
important effect of magnetic field on the guidance of the laser beam through the plasma channel
and have noticed that laser guidance increases with the increase in magnetic field. Results of thepresent analysis are very important for the successful implementation of the fast Igniter concept, inwhich, it is required that the intense guided pulse must propagate through the plasma region
without being absorbed and without beam break-up due to filamentation up to the edge of the core.

134

P1-98Charge-transfer cross sections of ground state He+ ions in collisions with He atoms andsimple molecules in the energy range below 4.0 keV
T. Kusakabe, S. Kitamuro, Y. Nakai, H. Tawara1), M. Sasao1)
Kinki University, Higashi-Osaka, Osaka 577-8502, Japan
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Helium is the second most abundant element in the universe. Indeed, helium ions were observed in
the outer space of the Earth by an extreme ultraviolet scanner on Mars Orbiter Planet-B [1]. Someobservations of X-ray emissions from comets have attracted much attention and the origin of thisemission has been interpreted as being due to charge transfer into solar-wind ions from thecometary gases [2]. Helium will be also one of typical impurities in D-T fusion plasmas as ashes,
if the D-T fusion reactions really occur in controlled thermonuclear fusion devices. Thus, the+2+accurate knowledge on charge-transfer collisions between He ions as well as He ions andcommon impurity molecules is critically important in proper handling of the helium ashes in thedivertor region, better understanding of the behavior of low temperature edge plasmas, and theirplasma-modeling in the fusion devices [3].
Total cross sections of these charge-transfer collisions have been extensively measured since the1955s [4]. In the low collision energy region, however, serious discrepancy is observed among
existing experimental data. Therefore, in the present work, we have systematically measured thecharge-transfer cross sections for the ground state He+ (1S) ions colliding with He atoms andsimple molecules (H2, D2, N2, CO, CO2) in the energy range of 0.2 to 4 keV by applying an initial
growth rate method. Furthermore, in He+ + He collisions, we will discuss about the detection
efficiencies of a micro channel plate detector for He+ ions and He atoms. The present observations
have been found to provide reasonably reliable cross-section data for the charge transfer of He+
ions in collisions with these molecules in the energy range below 4.0 keV. In He+ + D2 collisions,
any isotope effect was not observed over the present energy range, compared to H2 molecules.
References[1] I. Yoshikawa, et al., J. Geophys. Res., 106 (2001) 26057.
[2] C.M. Lisse, et al., Science, 274 (1996) 205.
[3] R.K. Janev, in Atomic and Molecular Processes in Fusion Edge Plasmas, (Plenum, New
York, 1995) p.1.
[4] J.B.H. Stedeford and F.B. Hasted, Proc. Roy. Soc. London, Ser.A, 227 (1955) 466.

135

P1-99Application of Collocation Meshless Method to Eigenvalue Problem
A. Saitoh, T. Itoh1), N. Matsui, A. Kamitani2) and H. Nakamura3)
University of Hyogo, Himeji, Hyogo 671-2280, Japan
1)Seikei University, Musashino, Tokyo 180-8633, Japan
2)Yamagata University, Yonezawa, Yamagata 992-8510, Japan3)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Many meshless approach [1], [2] have been proposed and have yielded excellent results in thefields of engineering and science. In spite of the convenience, meshless approaches are plagued by
two difficulties. First, the method for implementing the essential boundary condition is different
according to meshless approaches. Second, both the essential boundary condition and the natural
one are not exactly fulfilled on the boundary. If a new implementation method of not only the
essential boundary condition but also the natural one were proposed without dependence onmeshless approaches, the above demerit could be completely resolved.
In the previous studies, Kamitani et al. [3] have proposed the new method for implementing
the essential boundary condition to the meshless Galerkin/Petrov-Galerkin approach. The resultsof computations show that the accuracy of the proposed method is higher than that of the standard
one. In addition, Saitoh et al. [4] have investigated the applicability of the proposed method to thenonlinear Poisson problem.
The purpose of the present study is to develop numerical code for solving the eigenvalue
problem on the basis of the collocation meshless method and to investigate the performance of thecollocation EFGM by use of the code.
The numerical results will be shown in the conference.
References[1]T. Belytchko, Y.Y. Lu and L. Gu, Int. J. Numer. Methods Eng., 37 (1994), 229[2]S.N. Atluri and T. Zhu, Comput. Mech., 22 (1998), 117[3]A. Kamitani, T. Takayama, T. Itoh and H. Nakamura, Plasma and Fusion Research, 6(2011), 2401059[4]A. Saitoh, T. Itoh, N. Matsui, A. Kamitani and H. Nakamura, Plasma and Fusion
Research, 6 (2011), 2401089
This work was supported in part by Japan Society for the Promotion of Science under a Grant-in-Aid for
Scientific Research (B) No. 22360042. A part of this work was also performed with the support and under the
auspices of the NIFS Collaboration Research Program (NIFS11KNTS011, NIFS11KNTS014).

136

P1-100
Numerical Simulation of Electromagnetic Wave Propagation using Time Domain Meshless
MethodS. Ikuno1), Y. Fujita1), T. Itoh2), S. Nakata3), H. Nakamura4), A. Kamitani5)
1) Tokyo University of Technology, Katakura 1404-1, Hachioji, Tokyo 192-0982, Japan
2) Seikei University, Kichijoji-kitamachi 3-3-1, Musashino, Tokyo 180-8633, Japan3) Ritsumeikan University, Nojihigashi 1-1-1, Kusatsu, Shiga 525-5062, Japan
4) National Institute for Fusion Science, Shimoishi 322-6, Toki, Gifu 509-5292, Japan
5) Yamagata University, Johnan 4-3-16, Yonezawa, Yamagata 992-8510, Japan
In the Large Helical Device (LHD), the electron cyclotron heating device is used for plasma
heating. The electrical power which is made by the gyrotron system transmits to LHD by using
long corrugated waveguide. However, it is not clear that the shape of curvature of the waveguideor transmission gain of electromagnetic wave propagation theoretically.
Generally, Finite Differential Time Domain (FDTD) method is applied for electromagnetic wave
propagation simulation. FDTD method has provided the solution of Maxwell equation directly.
Furthermore, FDTD method has great advantages in terms of parallelization and treatment ofproblems and so on. However, the numerical domain should be divided into rectangle meshes if
FDTD method is applied for the simulation, and it is difficult to treat the problem constructed byarbitrary shapes.
The meshless approach has even developed, such as the element-free Galerkin (EFG) method andthe meshes local Petrov-Galerkin (MLPG) method and the radial point interpolation method(RPIM). As is well known that the meshes approach does not require finite elements or meshes ofa geometrical structure. And these methods are applied to a variety of engineering fields and thefields of computational magnetics. In particular, meshes approaches based on RPIM are applied totime dependent problems[1, 2].
The purpose of the present study is to develop numerical code for analyzing electromagnetic wave
propagation in arbitrary shapes of waveguide using meshes approach based on RPIM.
References[1]Y. Yu, Z. Chen, IEEE Trans. Micorwave Theory and Tech., Vol. 58, No. 3, (2010) 578.
[2]T. Kaufmann, C. Fumeaux and R. Vahldieck, In 24th International Review of Progress inApplied Computational Electromagnetics (ACES 2008), 426.

137

P1-101
Implicit Function with a Natural Behavior over the Entire Domain
T. Itoh, A. Saitoh1), A. Kamitani2), H. Nakamura3)
Seikei University, 3-3-1, Kichijoji-Kitamachi, Musashino, Tokyo, 180-8633, Japan
1) University of Hyogo, 2167, Shosha, Himeji, Hyogo 671-2280, Japan
2) Yamagata University, 4-3-16, Johnan, Yonezawa, Yamagata 992-8510, Japan
3) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan      Many kinds of meshless methods have been proposed, and have been applied to numericalsimulations of various fields including plasma physics and fusion science. In the meshlessmethods, although the elements representing a geometrical structure are not necessary, an analysis
domain has to be defined. To define the analysis domain, an implicit function [1, 2] is employed in
the eXtended Boundary-Node Method (X-BNM) [3]. In the X-BNM, it is assumed that, by using
the implicit function f(x), the boundary of the analysis domain is represented as f(x) = 0.
     In general, an implicit function f(x) has properties such as f(x) < 0 and f(x) > 0 for inside andoutside of the surface, respectively. For generating an implicit function f(x) from a large scatteredpoint data, the Partition of Unity (PU) based methods such as the Multi-level Partition of Unity
implicits (MPU) method [1] are often employed. However, the generated implicit function f(x) has
an effective range. Namely, f(x) = 0 is distributed not only on the surface but also outside of theeffective range. This is not a natural behavior of an implicit function, since the above properties
are not satisfied. Especially in the X-BNM, since a procedure of finding the boundary f(x) = 0 is
indispensable for evaluating the influence coefficients [3], it is desirable that f(x) = 0 is only
distributed on the boundary.
     The purpose of this study is to generate an implicit function satisfying the above properties
over the entire domain. To this end, two kinds of implicit functions are generated by using the PU
based method and the radial basis function based method [2], respectively, and these implicit
functions are smoothly combined by using an appropriate weight function.
References[1] Y. Ohtake, A. Belyaev, M. Alexa, G. Turk, H.-P. Seidel, ACM Trans. on Graph. 22,(2003) 463[2] G. Turk, J. F. O'Brien, ACM Trans. on Graph. 21 (2002) 855[3] T. Itoh, A. Saitoh, A. Kamitani, H. Nakamura, Plasma Fusion Res. 5 (2010) S2111

138

P1-102
A new De-noising Method of Laser-produced Plasma Penumbral Images by Principal
Component Analysis
S.Nozaki, A.Kinjo, S.Fujioka1), R.Azuma2), Y.-W.Chen3), Y.Namihira
University of the Ryukyus, Nishihara, Okinawa 903-0213, Japan.
1)Osaka University, Suita, Osaka 565-0871, Japan.
2)Aoyama University, Sagamihara, Kanagawa 252-5258, Japan.
3)Ritsumeikan University, Kusatsu, Shiga 525-8577, Japan
Penumbral imaging technique[1] can be applied to highly penetrating radiations such as neutrons.
The large aperture is used and the convolved image (penumbral image) is recorded on the detector.
The source image can be recovered by the deconvolution. Therefore the technique can be used toinertial confinement fusion research. In the inertial confinement fusion research, the fast ignition[2] approach can increase the number of fusion. However, the γ rays produced by the fast-heating
laser pollute the penumbral image as noise. Since the conventional deconvolution methods like the
Wiener filter cannot obtain the clear reconstructed image from the noisy penumbral image. Other
reconstruction methods are proposed to obtain the reconstructed image from the noisy penumbral
image. But these methods cannot obtain the reconstructed image since the signal-to-noise ratio(S/N) on the detector is extremely low due to the γ rays by the fast-heating laser.
In this paper, we propose a new reconstruction method by the principal component analysis[3].
The methods can efficiently remove the noise by high S/N training images which are obtained
from other experiments. We used the (2D)2PCA[4] method as a noise reduction method, whichcannot obtain clearer images but calculation cost is lower compared with the conventional PCA
methods. The efficacy of the proposed method is demonstrated by the computer simulations.
References[1] K. A. Nugent and B. Luther-Davies,Opt.Commun., 393 & 396(1984)49[2] R. Kodama, et al., Nature, 933 & 934(2002)418[3] Jian Yang, Zhang D., Frangi A.F., Jing-yu Yang, IEEE Trans. Patt. Anal. Machine Intell.,
131 & 137(2004)26[4] Daoqiang Zhang, Zhi-hua Zhou, Neurocomputing, 224 & 231(2005)69
This study is supported by the Rising Star Program for Subtropical Island Sciences of the University of the
Ryukyus, Okinawa, Japan.

139

P1-103
Studies on Plasma Production and Acceleration Using Rotating Electromagnetic Fields forElectrical Propulsion Engine

H. Wakabayashi, Y. Hayashi, S. Nakamoto, Y. Yasaka, H. Takeno
Kobe University, Hyogo, Kobe 657-8501, Japan
In the space engineering, VASIMR (Variable Specific Impulse Magneto-plasma Rocket)
researched by NASA[1] is attracting considerable attention. It has an important feature ofindependent control of thrust and specific impulse by using helicon wave plasma production andion cyclotron heating. This two-stage radio frequency (RF) system makes the equipment bigger
and complicated. The authors have proposed a simple system: an employment of the Rotating
electromagnetic Field Antenna (RFA) for plasma production and heating, which consists of a pairof double half turn antennas staggered by 90 degrees in the azimuthal direction. By controlling thephase difference between currents of two straps (φ), the ratio of excitation fields of left and righthand polarizations can be controlled. Each polarization field will lead excitations of slow and fast
waves, and thus, ion heating and plasma production, respectively.
In the electrical propulsion rocket experimental simulator, the effectiveness of RFA has been
examined. To the base hydrogen plasma produced by another RF power, an RF application by
RFA provides sinusoidal variations of ion saturation current (Iis) and ion temperature (Ti) to φ.
Under the ion cyclotron resonance condition, the Iis and Ti show maximum values of 120% and320% increments for φ=π/2 and -π/2, respectively. Moreover, 200% increment of Ti has also been
observed under the second harmonic resonance condition.
As a next stage, a study of simultaneous performance of plasma production and ion heating is inprogress. Using a helical antenna, forward left hand polarized wave and backward right handpolarized wave will be excited simultaneously, and ion heating and plasma production will be
performed, respectively. The experimental results on this scheme will be presented in theconference.
References[1] F. R. Chang Diaz et al. 42nd AIAA Aerospace Sciences Meeting and Exhibit, AIAA-2004-2150, (2004).

140

P1-104
Exact trajectory of a charged particle in several non-uniform magnetic fields
S. Oikawa, P. K. Chan, E. Okubo
Faculty of Engineering, Hokkaido University,
N-13, W-8, Sapporo 060-8628, Japan
We have derived exact expressions for trajectory of a charged particle in the presence of nonuniform magnetic fields. The types of magnetic field B include:(1) B(x) = ( 1 + a1x + a2x2 + a3x3 )B0, or the cyclotron motion in x-y plane.
In the simplest case, where B(x)=(1-x/LB)B0, the exact drift speed VD is given by
VD/v0 = 1 - (2/m)(1-E(m)/K(m)),
where v0 is the initial speed of the particle, K(m) and E(m) are the elliptic integrals of the first andsecond kinds whose argument m is defines as
m = 4ρ0 / (LB + 4ρ0sin2(α/2))
where ρ0 is an initial cyclotron radius at the origin, α stands for the initial phase angle of the
gyration.
(2) Bφ ∝ 1/R in cylindrical (R,φ,Z) coordinate, or a simple torus.
The exact drift speed VD in Z-direction for the initial pitch angle of π/2 is given by
VD/v0 = I1(-ρ0/R0)/I0(-ρ0/R0),
where I0 and I1 are the modified Bessel functions of the zero-th and the first kinds, R0 stands forthe initial radial position of the particle. Note that the drift speed is independent from the initial
phase angle α.
Also presented will be the exact trajectories and the drift speed for the arbitrary pitch angle
concerning (2).

141

P1-105
Application of the Binary Interaction Approximation to Plasma Oscillation
T. Kamei, S. Oikawa and M. Goto
Graduate School of Engineering, Hokkaido University, N-13, W-8, Sapporo 060-8628, JapanThe authors have developed a code to solve multibody problems for plasma oscillation using the
BIA [1].
In a binary system with an impact parameter b = b0cot(χ/2), a typical relative velocity change Δg is
given byΔg = 2gsin(χ/2) ∼ εg, ε ≡ 2b0 /Δl = U/K, (1)where b0 corresponds to χ = π/2 scattering, Δl is the average interparticle separation, U and K
stands for the potential and kinetic energies. In N-body systems with ε << 1, such as the fusion
plasma, Equation(1) means that three-or-more body interaction is of order of ε2 and can beignored. It should be noted that the Debye lengths λD in fusion plasma generally satisfy λD >> Δl,thus typical interaction is characterized by the nondimensional parameter ε.
Let us assume that there is a single positive charge fixed at the origin and the electrons are
randomly distributed initially around the positive charge.We will solve such a case using the BIA
schem.
References[1]S. Oikawa and H. Funasaka, "Binary Interaction Approximation to N-bodyProblems,"Plasma Fusion Res. 5, S1051(2010)142

P1-106
Accuracy Assurance in Binary Interaction Approximation for N-body Problems
M. Goto and S. Oikawa, T. Kamei
Graduate School of Engineering, Hokkaido University, N-13, W-8, Sapporo 060-8628, JapanThe authors have developed an algebraic model for multibody problems, and have shown that the
momentum transfer cross-section with our model is in excellent agreement with the exact one [1].
In a binary system with an impact parameter b=b0 cotχ/2, a typical velocity change in the relative
velocity Δg is given by
Δg=2gsin χ/2∼εg,ε≡b0/Δl(1)where b0 corresponds to χ=π/2 scattering, and Δl is the average interparticle separation. In N-bodysystems with ε≪1, such as the fusion plasma, Eq. (1) means that three-or-more body interaction isof order of ε2 and can be ignored. It should be noted that the Debye lengths λD in fusion plasma
generally satisfy λD≫Δi, thus typical interaction is characterized by the nondimensional parameter
ε. This parameter coincides approximately with U/K, where U and K stand for the potential andkinetic energies.
For a given time interval Δt and a set of initial conditions ri(0) and vi(0) for N-particles, first
calculate rl*(Δt) and vi*(Δt) using the exiting BIA code.
Second, calculate ri(Δt) and vi(Δt) from the same initial conditions then calculate ri(Δt) and vi(Δt)
from ri(Δt/2) and vi(Δt/2). If |ri*(Δt)-ri(Δt)| and |vi*(Δt)-vi(Δt)| for any i are less than the ERRORTOLERANCE, then calculation will be proceeded for t>Δt. If not, the time interval is reduced toΔt/2 this procedure will be repeated until errors become within the ERROR TOLERANCE.
References
S. Oikawa and H. Funasaka, “Binary Interaction Approximation to N-body Ploblems,”
Plasma Fusion Res. 5, S1051(2010)143

P1-107
Localization method using Microsoft Kinect for indoor structures
Y.Takabatake1), Y.Tamura2), N.Kashima2), T.Umetani;2)1)Graduate School of Natural Science, Konan University,
8-9-1, Okamoto, Higashinada-ku, Kobe, 658-8501, Japan2)Department of Intelligence and Informatics, Konan University,
8-9-1, Okamoto, Higashinada-ku, Kobe, 658-8501, Japan

Most nuclear fusion reactor is generally large and its buildings are also large. In such a large
space, a user cannot easily recognize where he/she is. Therefore it is useful to provide the user
current location information. The popular way to get the location information is to use Global
Positioning System (GPS); however, GPS cannot be used inside the building. For overcoming thisproblem, many researches for localization have been proposed from the stand point of machine
vision with cameras. Many vision-based localization methods using markers have been proposed[1]; however, it is necessary to install and allocate many artificial markers in large space. It costs a
lot of manpower and the markers are eyesore. One method without the artificial markers is thenatural feature point landmark database [2]. This method consists of two stages. In the first stage,
we reconstruct the environment from omni-directional image sequences. We find out naturalfeature points by processing the images and create the natural feature points database in which thecamera position and posture and image data is stored. In the second stages, when the user moves
with camera, the current position is calculated by using this database. As a result, the user canknow the current position and this system can provide suitable information to the user. For
example, if he wants to operate one device, this system can provide the information of operation of
the device in front of the user. An omni-directional multi-camera system is generally used for
grabbing environmental image, however this type of camera is very expensive. Moreover, it isdifficult to measure self-position accurately since experimental error of depth tends to be large
only using a RGB camera. On the other hand, the Microsoft Kinect sensor, which has a RGB
image camera and a depth camera, can improve this problem. Finally, we provide one example,
which reconstructs the room shape.
References[1] L. Naimark and E. Foxlin, Proc. IEEE/ACM Int. Symp. on Mixed and Augmented
Reality, 1 (2002) 27.
[2] M. Oe, T. Sato, and N. Yokoya, Proc. Scandinavian Conf. on Image Analysis, 14 (2005)
171.

144

Plenary / Invited / Oral
Sessions(Tuesday 29th November)PL-42-D Microwave Imaging Experiments on KSTAR
H.K. Park, G.S. Yun, W. Lee, M. Choi, J.H. Lee, N.C. Luhmann, Jr.1), C.W. Domier1),
A.J.H. Donné 2,3), S.W. Yoon4), Y.M. Jeon4), S.I. Park4), and KSTAR Team 4)
Pohang University of Science and Technology, Pohang, Korea
1)University of California at Davis, California, U.S.A
2)FOM-Institute for Plasma Physics Rijnhuizen, The Netherlands
3)Eindhoven University of Technology, The Netherlands4)National Fusion Research Institute, Daejeon, Korea
The 2-D Electron Cyclotron Emission Imaging (ECEI) system has been a major diagnostic tool forthe study of various MHD modes during the last two KSTAR campaigns (2010 and 2011).
Following the first application to sawtooth physics, the detailed study of the growth of ELMs [1]and a newly observed core instability [dual flux tubes only with electron cyclotron heating (ECH)]
in both campaigns will be reviewed in this paper. The preliminary images obtained during the
suppression experiment of ELMs using in-vessel control coils (IVCC) and meso-scale Te
fluctuations during the H/L transition will be addressed. A toroidal separated second ECEI systemfor 3-D reconstruction of the MHD instabilities which will verify the high m and n mode numbers
and rotation speeds (poloidal and toroidal) for 2012 campaign will be discussed. At the same time,a unique 2-D Microwave Imaging Reflectometry system [2] which is capable to measure density
fluctuation will be deployed for the 2012 campaign to study the turbulence based transport physics
on KSTAR.
References[1] G.S. Yun, et al., Phys. Rev. Lett., 107, 045004 (2011)[2] W. Lee et al., Plasma Fusion Res., 4, 2402037 (2011)
This work was supported by NRF, Korea, Euratom, DE-AC02-09CH11466 and DE-AC05-00OR22725.

147

I-5
Super-X Reality: Implementing an Innovative Divertor
H. Meyer and G.M. Fishpool for the MAST and MAST Upgrade Teams
Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK
One of the most prominent challenges for a fusion power plant based on the magnetic confinement
concept is that of handling the exhaust plasma. For this a divertor is implemented in current
devices, where a thin layer (DEMO: ~1cm) of scrape-off-plasma is guided magnetically onto
special targets. In reactor size devices such as DEMO the power density reaching the targets needs
to be reduced significantly to avoid excessive target erosion, e.g. by radiation. At present it is notclear if a viable divertor solution for a full-scale reactor exists. The Super-X divertor concept
could play a significant part in the design of large-scale fusion devices, in particular for tightaspect ratio, by offering a way of mitigating the power flux to the target. In this concept, thepoloidal magnetic field of the divertor is used to take the target strike zone through an intermediate
region of low poloidal field out to a larger major radius. The lower total field at the larger radius
leads to a reduced parallel heat flux, whilst reducing the poloidal field enlarges the plasma wettedarea. In addition, the intermediate region of reduced poloidal field increases the connection lengthof the plasma in a volume some distance away from the core plasma, allowing for volumetriceffects to further reduce the power density. The current project to upgrade MAST incorporates two
closed, actively pumped divertors in double null configuration capable, of realising, for the firsttime, this Super-X concept. Moreover, 2 sets of 8 coils controlling the divertor field allow thecomparison of various divertor geometries with different strike point radius, connection length andflux expansion; including conventional divertor geometries. This flexibility, together with anextensive set of diagnostics, will be used to test the divertor physics leading to a reduced power
flux at the target such as increased connection length, role of collisional processes, reduction ofparallel power density, and the approach to detachment. SOLPS and other tools were used toassess the balance between competing design choices like recycling at the entrance, closure,pumping and achievable connection length. For example the geometry of the divertor throat
represents a compromise between sufficient closure and tolerable recycling and modelling shows
neutral compression factors in excess of 100 with the chosen design, with the promise of improved
performance of the MAST core plasma. Furthermore, engineering challenges needed to be
overcome, such as the control requirements of coil currents to better than 1% out of ~5kA total forsome divertor coils to ensure the stability of the intermediate low poloidal field region. In thispaper we will present the current Super-X design in MAST upgrade, discuss the basic designprinciples outlined above and the challenges they pose for the design and specification of theplant, as well as the expectations of the physics programme that will be possible.
Work funded by RCUK Energy Programme and EURATOM.

148

O-5
Influence of the isotope effect on the charge-exchange process between hydrogen isotopes
and ions and atoms of plasma facing component materials
Inga Yu. Tolstikhina, Daiji Kato1) and V. P. Shevelko
P. N. Lebedev Physical Institute, Leninskii Prospekt 53, Moscow 119991, Russia
1) National Institute for Fusion Science, Oroshi-cho 322-6, Toki-city, Gifu 509-5292, Japan
The plasma-wall interactions remain one of the main areas of research in the physics andtechnology of controlled thermonuclear fusion. The conditions in the peripheral region largely
determine also the processes in the main plasma. Numerical simulations of the near-wall anddivertor region plasma in fusion devices are based on the knowledge of the rates of the elementary
processes involving plasma particles and ions and atoms sputtered from the plasma facing
components (PFC). The energies of collisions in the peripheral region are very low (about 1 – 500eV/u). Under these conditions the charge-exchange becomes the dominant process for theneutralization and population of the excited states of plasma impurities and plays an important rolein radiative cooling, particle transport and ions charge distribution. The isotope effect, foundrecently in charge-exchange reactions between hydrogen isotopes and α-particles [1-3] and PFC’sions (Li, Be, C) [4], manifests in a significant difference (up to several orders of magnitude) in the
charge-exchange cross sections: the heavier the isotope the larger the cross section. This effectoccurs at very low collision energies due to rotational interaction in close collisions – thescattering angle depends on reduced mass of colliding particles. Since DT burning plasma
experiments are planned in ITER, hydrogen isotopic effects on transport of the sputtered particles
is an important issue. We present here calculations of the charge-exchange cross sections in slow
collisions of Li, Be, C and W ions with H, D and T. The isotope effect is studied in adiabatic
approximation [5] which, in the theory of atomic collisions, is used to describe electronictransitions when the collision velocity is small and the nuclear motion can be treated classically. In
this theory, there are no assumptions on the specific form of the electronic Hamiltonian, and only
the smallness of the relative nuclear velocity is used. It results in a deeper understanding of thenature of nonadiabatic transitions. Since the isotope effect occurs at collision energies where theadiabatic theory applies, the adiabatic approximation is a natural theoretical framework forstudying the effect. The cross sections are calculated with ARSENY code [6] based on the hiddencrossing method. The charge-exchange cross sections in collisions of W +(6s) with H, D, T, which
are presented here for the first time, at the collision energy of 30 eV/amu are 1.56E-19 cm2, 1.18E18 cm2, and 1.25E-16 cm2, correspondingly. This significant difference in the cross sections for H
and T targets (three orders of magnitude) indicates the necessity for accounting the isotope effectin edge plasma modeling.
References[1] N. Stolterfoht, R. Cabrera-Trujillo, Y. Ohrn, E. Deumens, R. Hoekstra, and J. R. Sabin, Phys.
Rev. Lett. 99 (2007) 103201.
[2] N. Stolterfoht, R. Cabrera-Trujillo, P. S. Krstic, R. Hoekstra, Y. Ohrn, E. Deumens, and J. R.
Sabin, Phys. Rev. A 81 (2010) 052704.
[3] R. Cabrera-Trujillo, J. R. Sabin, Y. Ohrn, E. Deumens, and N. Stolterfoht, Phys. Rev. A 83
(2011) 012715.
[4] Inga Yu. Tolstikhina, Daiji Kato, and V. P. Shevelko, Phys. Rev. A 84 (2011) 012706.
[5] E. A. Solov’ev, Sov. Phys. Usp. 32 (1989) 228.
[6] E. A. Solov’ev, Workshop on HiddenCrossings in Ion-Collisions and in Other Nonadiabatic
Transitions (Harvard Smithonian Centre for Astrophysics, Cambridge, MA, 1991).

149

O-6
Divertor Simulation study using the GAMMA 10 end-mirror cell
H. Takeda1), Y. Nakashima1), K. Hosoi1), K. Ichimura1), T. Furuta2), T. Higashiyama2), M. Toma2),
A. Hatayama2), T. Ishii1), H. Ueda1), M. Sakamoto1), M. Ichimura1), M. Yoshikawa1),
J. Kohagura1) and T. Imai1)
1) Plasma Research Center, Tsukuba University, Ibaraki 305-8577, Japan
2) Science and Technology, Keio University, Kanagawa 223-8522, Japan   Heat load of plasma particles led to divertor plates approaches tens of MW/m2 on steady-state
operation. Therefore, it is important to develop the divertor plates which attain the level of highheat load from the divertor plasma. The divertor simulation experiments are widely performed byusing linear devices. In the GAMMA 10 tandem mirror, divertor simulation experiments wereplanned and started [1-3]. GAMMA 10 consists of a central-cell, two anchor-cells, twoplug/barrier-cells and two end-cells. Plasma heating systems which are ICRF, NBI and ECH are
installed for plasma production/heating and confinement. By applying ICRF in the central-cell andanchor-cells, end-loss plasma can be generated. In this plasma state, contribution of ion dominatesin heat flux. In the case of superimposing ECH in plug/barrier-cell on ICRF-produced plasmas theeffects of electron become significant. Thus, characteristics of end-loss plasma toward the divertor
plate can be changed by controlling heating systems in GAMMA 10.
  The purpose of the study is analysis of heat and particle flux in end-loss plasmas for divertor
simulation experiments. Heat and particle fluxes were measured by using directional probes andcalorimeters. In the case of only ICRF heating, heat-flux has strong dependence on diamagnetism
which is time integrated in central-cell. While, particle fluxes are proportional to electron linedensity in central-cell. These results mean that the characteristics of end-loss plasma aresignificantly influenced from the central-cell plasma parameters. It is also found that spatial
profiles of heat and particle fluxes have peaks on the axis and are governed by the magnetic field
configuration in the end-mirror region.
  In this paper, we report the detailed results of characteristics of heat and particle fluxes from theend-mirror exit.We also describe an initial result of the background plasma simulation forunderstanding the experimental data by using two-dimensional fluid code
References[1] Y. Nakashima, et al. Fusion Eng. Design 85 issue 6, Nov. (2010) 956-962.
[2] Y. Nakashima, et al. Fusion Sci. Technol. 59 No.1T (2011) 61-66.
[3] Y. Nakashima, et al. 23rd IAEA Fusion Energy Conference (October 11-16, 2010,      Daejeon,Korea)IAEA-CN-180 FTP/P1-33.

150

I-6
Advances in Lower Hybrid Current Drive for Tokamak Long Pulse Operation: Technology
and Physics
G.T. Hoang
CEA, IRFM, F-13108 Saint-Paul-lez-Durance, France
Progress and achievements of LHCD technology for long pulse operation in tokamaks arereported, including the development of CW high power klystrons, and its evolutions towards
ITER.
So far, the 3.7GHz / 700kW CW klystrons have been produced in series by Thales Electron
Devices (TED) for Tore Supra. First series of eight klystrons, are operating now (3 MW/ 43s
achieved after several days of operation). A prototype of 500 kW CW klystron operating at 5 GHz
has been developed for KSTAR by Toshiba Electron Tubes and Devices (TETD) for the steadystate RF source. This prototype klystron has produced RF output powers of 300 kW / 800s and
450kW / 20s using during the commissioning.
The review also reports the situation on wave coupling and antennas, with the latest Tore Supra
results of the new CW Passive-Active Multi-junction launcher; the antennae concept foreseen for
ITER LHCD system.
Finally, the activity in view of ITER of the worldwide LHCD community is updated. An LHCD
system capable of fulfilling several important tasks on ITER is presented. Based on experimental
results, simulations of ITER scenario using LHCD have been revisited, including advanced modesof operation in tokamaks as well as the LHCD-assisted start-up. LHCD is found to be a key tool
for two tasks: i) extend the burn duration, up to 500s (save Volt-seconds, from early plasmaphases); ii) help accessing and sustaining Steady-State (drive far off-axis current, complementarily
to Bootstrap Current, NBCD and ECCD).

151

I-7
ECW / EBW Heating and Current Drive Experiment Results and Prospects to CW
Operation in QUEST
H. Idei1), H. Zushi1), K. Hanada1), K. Nakamura1), A. Fujisawa1), M. Hasegawa1),
N. Yoshida1), M. Sakamoto2), H. Watanebe1), K. Tokunaga1), Y. Nagashima1), A. Ejiri3),
M. Sakaguchi4), E. Kalinnikova4), M. Ishiguro4), S. Tashima4), A. Fukuyama5) H. Igami6),
S. Kubo6), S.K. Sharma4), T. Ryokai4), H.Q. Liu4), M. Isobe6), K. Nagaoka6), H. Nakanishi6),
N. Nishino7), S. Kawasaki1), H. Nakashima1) A. Higashijima1), Y. Takase3), T. Maekawa8),
O. Mitarai9), M. Kikuchi10) and K. Toi6)1)Research Institute for Fusion Science, Kasuga, Japan
2)Plasma Research Center, Tsukuba Univ., Tsukuba, Japan3)Department of Complexity Sci. and Eng., Univ. of Tokyo, Kashiwa, Japan4)Interdisciplinary Grad. School of Eng. Sci., Kyushu Univ., Kasuga, Japan5)Department of Nuclear Engineering, Kyoto Univ., Kyoto, Japan6)National Institute for Fusion Science, Toki, Japan
7)Mechanical System Engineering, Hiroshima Univ., Hiroshima, Japan8)Graduate School of Energy Science, Kyoto Univ., Kyoto, Japan
9)Inst. of Ind. Sci. and Tech. Research, Tokai Univ., Kumamoto, Japan
10)Japan Atomic Energy Agency, Naka, JapanThe Q-shu university experiment with steady state spherical tokamak (QUEST) has been
conducted in Kyushu University. Electron Bernstein wave heating and current dive (EBWH/CD)
is one of attractive candidates to sustain the steady-state ST plasma.The plasma current of ~15kA
with an aspect ratio of 1.5 was started up and sustained by only 8.2GHz RF injection. The longpulse discharges of 10 kA and 15 kA were attained for 37 s and 20 s in the limiter configurations,
respectively. The new density window to sustain the plasma current was observed in the overdense
plasma. The single-null divertor configuration with the high plasma current (<~25kA) was attainedin the 17 s plasma sustainment.
In order to attain the steady state ST configuration, several topics have been focused in the
QUEST. The phased array antenna system has been developed to excite an elliptically polarized O-mode in the oblique injection for the EBW/CD experiments. In the long pulse discharge with thelimiter configuration, the hot spots were appeared, and the metal impurity contamination wasobserved. The plasma position and shaping for the divertor configuration has been optimized to
avoid the hot spot formation. The hot first wall is being designed to reduce hydrogen retention at
the wall which causes the discharge termination due to the hydrogen desorption in the long pulse
discharge. Some trials for the CW operation in the QUEST will be introduced and discussed.

152

O-7
Recent Results from the Development of the Electron Cyclotron Heating
System for JT-60SA toward High-Power Long-Pulse Operations
A. Isayama, T. Kobayashi, K. Yokokura, M. Shimono, M. Sawahata, S. Suzuki, M. Terakado,
S. Hiranai, K. Wada, J. Hinata, Y. Sato, K. Hoshino, S. Moriyama, K. Sakamoto and
K. Hamamatsu
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japan
Electron cyclotron (EC) waves are widely used in tokamaks for many purposes, such as electron
heating, current drive and instability control, due to their capability of highly localized deposition
profile. In addition, EC waves can be used to assist plasma initiation and to clean the first wall. In
JT-60SA, an electron cyclotron heating (ECH) system will be installed and used for the above
purposes. The frequency of the EC wave is 110 GHz, and the total injection power is expected tobe 7 MW using 9 gyrotrons. Although the ECH system for JT-60U is reused as much as possible,
modifications and developments of almost all components, such as power supplies, gyrotrons,
transmission lines, antennas and control systems, are required, because the specifications for JT60SA are much beyond those for JT-60U. In particular, required pulse duration at an output powerof 1 MW is significantly extended from 5 s to 100 s. In 2010, pulse duration for 1 MW output was
extended to 31 s by installing a new mode convertor, which was designed to reduce strayradiofrequency power [1]. An increase in the temperature of coolant water for the DC break of the
gyrotron, which hindered the extension of the pulse duration, was found to decrease to about halfof its previous value. And the temperature reached a steady state in about 10 s with a temperature
increase of ~35°C. From this result, it was concluded that there is no serious limit for the gyrotron
in extending the pulse duration. On the other hand, the temperature on the surface of thetransmission line without cooling continuously increased; the temperature reached about 50-70°C
for a 1 MW, 15 s output. To suppress such a temperature increase, whose typical time constant ison the order of 100 s, the straight section of the transmission line was covered with copper plates
having water pipes. The effectiveness was confirmed in a 99 s operation at 0.3 MW. In 2011, inorder to achieve longer pulse duration, 60.3 mm-diameter components were installed for thetransmission line in place of the previous 31.75 mm-diameter components. A low-powertransmission test at typically 1 mW and a subsequent high-power transmission test at ~0.5 MW
showed no problem. Furthermore, a cathode resistor unit was modified to enable long-duration
operations of the gyrotron power supply, and the capability of heat resistance and heat removal in
the transmission line was improved. After these modifications, gyrotron operations toward thetarget value of 1 MW for 100 s were restarted and are now ongoing. In addition to the upgrade of
the present gyrotrons, a new gyrotron is being developed. The main objective is to deposit the EC
waves in the central region of JT-60SA plasmas for the toroidal magnetic field of 2.25 T; EC wave
frequency of 130-140 GHz is desired for that purpose. In order not to increase the total number ofgyrotrons, a dual-frequency gyrotron was designed. The second EC wave frequency was chosen to
be 138 GHz from the viewpoint of the oscillation mode and the wavelength. In this combination,
the oscillation modes for 110 GHz and 138 GHz are TE22,8 and TE27,10, respectively, and the ratioof these wavelengths is 5:4. A superconducting magnet enabling higher magnetic field is also
being fabricated. A gyrotron output window made of synthetic diamond with a thickness of 2.3
mm will also be fabricated to allow the transmission of both frequencies. The fabrications will befinished in March 2012, and the operations of the new gyrotron will be started after its installation.
References[1] T. Kobayashi et al., to be published in Nucl. Fusion (2011).

153

O-8
Steady State Operation of Ion Sources for Fusion Plasma Heating
M. Wada, T. Kasuya, T. Kenmotsu, S. Maeno1), M. Nishida, M. Nishiura2), K. Shinto3),
H. Yamaoka4)
Doshisha University, Kyotanabe, Kyoto 610-0321, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)Novelion Systems, Co. Ltd., Kyotanabe, Kyoto, 610-0394, Japan
3)Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japan4)Riken, Harima, Hyogo 679-5148, JapanThe neutral beam injection (NBI) system for ITER will extract negative deuterium ions from radio
frequency (RF) driven ion sources instead of cathode arc discharge sources [1]. However, quasisteady state operation of ion sources with RF plasma excitation may also exhibit some problems
like high temperature cathode discharge. A comparative research to excite ion source plasma atvarious RF frequencies has been started to clarify potential problems associated with steady state
RF ion source operation for NBI.
Microwave power at 14 GHz has produced plasma in a test ion source with the volume less than
50 cm3. Change in the magnetic field configuration has largely altered the performance of the ion
source. The magnetic mirror geometry with the axis set perpendicular to both beam extraction andthe electric field of the incident electromagnetic wave, created a dense plasma in the ion source,
but the field hindered extraction of negative hydrogen ions (H-) from the source [2]. A permanent
magnet induces a strong axial magnetic field in the ion source of the current design. It produces afield large enough to realize the electron cyclotron resonance condition, while the field near the
extractor hole realizes the “tent filter” geometry suitable for H- extraction [3]. Experimental resultsof the internal conductor design will be also reported.
References[1] D. Marcuzzi et al., Fusion Engg. Dsgn. 82 (2007) 798.
[2] M. Wada et al., Plasma and Fusion Res. 6 (2011) 2406100.
[3] H. Tobari et al., Rev. Sci. Instrum. 79 (2008) 02C111.
This work has been conducted as the LHD Collaboration Research sponsored by National Institute for Fusion
Science.

154

I-8
Microwave Reflectometry Diagnostics: Present day systems and challenges for future devices
T. Estrada
Laboratorio Nacional de Fusion. As. Euratom-CIEMAT, 28040 Madrid, Spain
Microwave Reflectometry, based on the reflection of a microwave beam at the plasma cut-off
layer, is applied to measure not only the electron density profile but also plasma turbulence andplasma rotation. During the 1990s different approaches were developed to overcome thedeleterious effect of the turbulence on the density profile measurement, becoming nowadays analmost routine diagnostic in many devices. Similarly, turbulence measurements have been
extended from the initial single frequency systems to multiple frequency or ultra-fast sweep
frequency systems. Moreover, in the last decade a new technique, Doppler reflectometry, has beendeveloped to measure simultaneously plasma turbulence and flows with excellent spatio-temporal
resolution. All these advances make reflectometry a very attractive diagnostic presently used inalmost all fusion devices. Although it is not straightforward, the extension to future devices like
ITER is still possible partially due to the limited access needed to accommodate the antennas andthe possibility to use stainless steel or carbon-based materials for the antennas keeping thesensitive elements as microwave sources and detectors outside the radiation area. In ITER several
reflectometry systems are being developed for profile and turbulence measurements [1]. Besides,
reflectometry will be used as an alternative approach to the magnetic systems in long pulse
operation for plasma position control [2]. All these reflectometry systems face some difficulties
linked to the intense radiation and the long pulse operation. This imposes a careful design of thefront-end to keep radiation and thermal and mechanical stresses at acceptable levels and to allow
maintenance by remote handling [3]. Relativistic effects as well as problems associated with nonabsorbed microwave stray-radiation should be also considered as they could hinder theachievement of the required measurement accuracy in some plasma scenarios. Besides, anintelligent handling of the huge amount of data is mandatory to allow real time monitoring of longpulse discharges.
References[1] G. Vayakis et al. Nuclear Fusion 46, S836 (2006)[2] J. Santos et al. Rev. Sci. Instrum. 81, 10D926 (2010)[3] G. Perez et al. Fusion Engineering and Design 84, 1488 (2009)

155

I-9
ITER instrumentation and control system toward long pulse operation
Izuru YONEKAWA, Jean-Yves. JOURNEAUX, Antonio Vergara Fernandez,
Jean-Marc FOURNERON, Anders WALLANDER, Wolf-Dieter KLOTZ.
ITER Organization, Route de Vinon sur Verdon – 13115 St Paul Lez Durance – France
ITER is a long-pulse tokamak with elongated plasma. The nominal inductive operation produces a
D-T fusion power of 500 MW for a burn length of 300-500 s, with the injection of 50 MW of
auxiliary power. With non-inductive current drive from the H&CD systems, the burn duration isenvisaged to be extended to 3000 s.
The term ITER Instrumentation & Control (I&C) includes everything required to operate the ITER
facility. It comprises three vertical tiers; conventional control, interlock system and safety system,and two horizontal layers; central I&C systems and plant system I&C. CODAC (Control, Data
Access and Communication) system forms the upper level of the hierarchy, and is theconventional central control system of ITER architecture. CODAC system is responsible for
integrating all plant system I&C and enable operation of ITER as a single integrated plant. CIS(Central Interlock System) and CSS (Central Safety System) also form the upper level of thehierarchy to supervising and integrating all plant system interlock and safety functions. Plant
system I&C forms the lower level of the hierarchy, and provide dedicated plant data acquisition,
plant status monitoring, plant control and plant protection functions to perform individual plantsystem operation under the supervision of central I&C systems.
Due to the long pulse operation of ITER, pulse scheduling and plasma control system shallprovide flexibility be able to change discharge scenario and plasma control algorithms according
to the phenomenon during the pulse. The concepts of plasma discharge scenario associate with pre-defined control algorithms are considered and further investigating collaborate with ITER parties.
ITER produces two different data streams, engineering data and scientific data. Engineering datacontains plant system status data and alarms during plant system operation. The majority ofscientific data is dominated by plasma diagnostics and some of these data shall also be used forreal-time plasma control.
References[1]A. Wallander et al. “Baseline Architecture of ITER control system”, IEEE TNS 2011 (inpress)

156

O-9
Diagnostic- and other in-vessel components in the high-power μ-wave backgroundof ECR heated steady-state discharges
M. Hirsch, D. Hathiramani, H. P. Laqua, J. W. Oosterbeek 1),
J. Baldzuhn, T. Bergmann, C. Biedermann, H. v. d. Brand 1), R. Burhenn, A. Cardella, K. Grosser,
M. Endler, M. Koeppen, R. Konig, M. Laux, H. Thomsen, D. Zhang and the W7-X Team
Max-Planck Institut fur Plasmaphysik, EURATOM Association, D-17491 Greifswald, Germany1) Technische Universiteit Eindhoven, Den Doelch 2, 5612 AZ Eindhoven, The Netherlands
Incompletely absorbed Electron Cyclotron Resonance Heating (ECRH) beams which are multiplyreflected and scattered at the highly reflective walls result in a nearly isotropic μ-wave backgroundin the plasma vessel. During long pulse operation this stray radiation may cause severe
overheating for μ-wave absorbing in-vessel components such as gaskets, bellows, windows,
ceramics and cable insulations. The level of unabsorbed μ-wave radiation is particularly high in
scenarios with moderate single pass absorption only such as high-density scenarios with X2heating close to cut-off conditions, O2-, X3-heating or OXB mode-conversion heating.
Extrapolations to W7-X with 8MW ECRH (140 GHz) and O2-heating yield values up to 150
kW/m 2. With well absorbing plasma (X2, medium density) the stray radiation level is reduced bya factor of ~100.
In-vessel components of W7-X are qualified for 30 min loads of isotropic 140 GHz radiation at
power flux densities of up to 50 kW/m2 in the MIcrowave STray RAdiation Launch facility,
MISTRAL, operated at IPP Greifswald. Microwaves are provided from a power gyrotron withadjustable duty cycle via a corrugated transmission line. An issue is the development of calibrated
sensors for the isotropic radiation also as an operational diagnostic of W7-X.
Measures to protect sensitive poorly cooled items can be tight metallic shields with the remaining
holes for pumping being smaller than λ/2 (<1mm). Additionally the power load may be reduced by
μ-wave absorbers consisting of ceramic layers on cooled metal structures. The spurious sensitivityof a bolometer diagnostic to the μ-wave radiation background was reduced by a factor of 300 via μ-wave absorber coating in the detector enclosure and - additionally - μ-wave reflecting meshes infront of the detector. Windows are closed for isotropic μ-waves with conducting Indium Tin Oxide(ITO) layers while remaining transparent in the visible and UV. Critical issues under investigation
are the shielding of cable isolations, the rather high effective μ-wave absorption of poorly cooledmeshes, e.g. protecting flexible cooling pipes, the shielding of cryo-pumps and port-bellows.

157

O-10
Fatal damages due to breakdown on a diagnostic mirror located outside the vaccum vessel in
JT-60U
Shin Kajita, Takaki Hatae1), Takeshi Sakuma1), Shuichi Takamura2), Noriyasu Ohno3),
Kiyoshi Itami1)
EcoTopia Science Institute, Nagoya University, Nagoya 464-8603, Japan
1)Japan Atomic Energy Agency, Ibaraki 801-1, Japan
2)Aichi Institute of Technology, Toyota 470-0392, Japan3)Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan
In fusion devices, the initiation of arcing has been an important plasma surface interaction issue,
because it could cause damages on divertor and first wall materials and deteriorate the plasma
performance consequently [1]. Recently, enhancement of the initiation of arcing on divertor
surfaces has been pointed out due to the surface morphology change and transient heat loads, suchas ELMs and disruptions [2]. Inside the vacuum vessel, it seems that the arcing has beenfrequently observed, and the existences are more or less recognized. However, it is likely thatarcing or breakdown phenomena take place not only inside the vessel, but also outside the vessel.
In this study, we will report that a diagnostic mirror in JT-60U, which was located outside thevacuum vessel, for Thomson scattering measurement optics in JT-60U has been seriously
damaged by some discharge phenomena. Because the damages were so fatal that the mirror had tobe replaced after the breakdown. Details of the characteristics of the damage are revealed from theobservation of the trails. Furthermore, the mechanism to cause the damages is discussed based on
the observation. Many fine trails were found on the surface. The trails could be categorized into
two different types with respect to the trail width. It is thought that surface discharge was initiatedon the mirror. The mechanisms to lead the damages were discussed based on the observation. This
study issues warning on the components to be installed in ITER and future fusion devices both
inside and outside the vacuum vessel.
References[1]G. McCracken and D. Goodall, Nucl. Fusion 18(1978)537.
[2]S. Kajita, S. Takamura and N. Ohno, Nucl. Fusion 49(2009)032002.

158

I-10
Summary of Japanese Researches on Tritium Science and Technology for Fusion Reactor
T. TanabeInterdisciplinary Graduate School of Engineering Science,Kyushu University
6-10-1, Hakozaki, Higashi-ku, Fukuoka 812-8581, Japan
In order to establish a D-T fusion reactor as an energy source, economical conversion of fusionenergy to electricity and/or heat, having enough margins in tritium breeding, and insuring tritium
safety must be simultaneously achieved. Scientists and researchers working on Tritium in Japanare now tackling with T related problems. Their research subjects can be categorized into two, (1)Science and technology and (2) Tritium safety.
Researches on (1) Science and technology have been carried out mainly Japanese University and
JAEA under the project of “Tritium science and technology for a fusion reactor” supported by
Grant in Aid for Scientific Research, Ministry of Education, Culture, Sports, Science andTechnology (MEXT), Priority area No.467. (2007-2011) [1]. In addition JAEA is doing its
original research works mainly related to ITER. Some works have also been done under the frame
work of NIFS-LHD collaborations through Toyama Univ.
Researches on (2) Tritium safety have been carried out mainly under the frame work of LHD
Project Research Collaboration of NIFS. Detailed subjects are (2-1) Behavior of Tritium inEnvironment includes (i) Water analysis( rain, river, groundwater, etc) , (ii) OBT behavior(measurement, Incorporation), (iii) Modeling (water, air, OBT, dose) and (iv)) Atmosphericanalysis (Chemical forms) and (2-2)Studies on biological Effects include (i) establishment of a
hypersensitive assay system that can be applied for experiments in radiation biology (cultured
cells, transgenic mice), (ii) Biological responses to low-dose (rate) tritium radiation, especially totritiated water (HTO) exposure and (iii) Molecular mechanisms of DNA damages and repair.
Many researchers from various universities, and institutes such as NIFS, JAEA and IEA (Inst.
Environmental Science) in Japan are involved in the research programs given above. Most of theirrecent results are presented in 9th International Conference on Tritium Science and Technology,
Nara Japan Dec.2010) [2], co-organized by the project “Tritium science and technology for afusion reactor”and NIFS. Present paper summarizes recent results given by the project “Tritium
science and technology for a fusion reactor” focusing on tritium issues to be solved forestablishment of a fusion reactor.
References[1] See web site; http://tritium.nifs.ac.jp/english/index.html[2] Proceedings of 9th International Conference on Tritium Science and Technology. Fusion
Sci. Technol. (2011) in press

159

O-11
Estimation of decay heat in fusion DEMO reactor
Y. Someya, K. Tobita
Japan Atomic Energy Agency, Rokkasho-mura, kamikita-gun, Aomori 039-3212, Japan
Decay heat of activated materials is important in safety of fusion DEMO reactor against loss of
coolant-flow accidents. Effects of decay heat for main components of the SlimCS [1] DEMO
reactor are studied with a one-dimensional code THIDA-2 [2]. The major parameters of SlimCS
are a plasma major radius of 5.5 m, aspect ratio of 2.6 and fusion power of 2.95 GW [1]. The main
reactor components consist of the inboard (IB) and outboard (OB) blanket modules, and divertor.
The decay heats of IB blanket, OB blanket and divertor are calculated under the neutron wallloading of 2.5 MW/m2, 3.5 MW/m2 and 1.5 MW/m2, respectively. Here, the blanket is assumed tobe composed of structural material with F82H, tritium breeder with Li2TiO3 pebbles and neutron
multiplier with Be12Ti pebbles. The blanket is filled with the mixture Li2TiO3 pebbles and Be12Ti
pebbles. The surface of the blanket is covered with 0.2 mm-thick tungsten (W). The divertor isassumed to be made of W mono-block and F82H. For the reactor with a fusion power of 2.95 GW,
the decay heat of IB blanket, OB blanket and divertor are roughly estimated to be as high as 29
MW, 146 MW and 6 MW, respectively, immediately after the shutdown of operation. This meansthat the blanket and divertor need to be actively cooled during several year storage in the reactor or
hot-cell. Three days after the shutdown, the total decay heat decreases to 15 % of the initial one.
However, this value is higher than decay heat for fission reactor. The total decay heat for a fission
reactor with an electrical generating power of 460 MWe is 5.2 MW at three days after shutdown.
In this paper, main radionuclides contributing to decay heat are estimated for different fusion
powers in the fusion DEMO reactor. Finally, a detailed comparison of between fusion and fission
reactor is discussed from view point of decay heat.
References[1] K. Tobita et al., Compact DEMO, SlimCS: design progress and issues, Nuclear fusion 49(2009) 075029 (10pp)[2] Y. Seki et al., THIDA-2: An Advanced Code System for Calculation of Transmutation,
Activation, Decay Heat and Dose Rate, JAERI 1301, 1986

160

O-12
OVERALL BURNING EFFICIENCY OF TRITIUM AND TRITIUM BALANCE
IN A D-T FUSION REACTOR
M. Nishikawa
Kyushu University
Tritium consumed in the plasma vacuum vessel must be supplied from the breeding blanket tomaintain the self-sustainable D-T nuclear burning because no effective tritium resources fromoutside are expected at present. Then, the attainable tritium breeding ratio which is estimated fromthe neutron usage and the recovery efficiency of the bred tritium in the blanket system must belarger than the required tritium breeding ratio which is estimated from the tritium balance in afusion reactor. Accumulation of tritium is also required for preparation of the initial inventory ofnext reactors to be constructed.
Recent plasma-wall interaction studies have revealed that a considerable amount of tritium
introduced into the plasma vacuum vessel is trapped at the re-deposition layer, though the trapping
behavior under the power plant condition is not qualitatively understood yet.
It is considered that trap of tritium to the co-deposits or tritium loss due to plasma driven
permeation increases with proportional to the tritium introduction rate to the plasma vacuum
vessel. The tritium inventory in the fueling system also proportionally increases with increase of
the tritium introduction rate. The tritium introduction rate is decided from the overall burning
efficiency and the energy output. Some tritium is also considered to introduce to the pedestal
plasma region for the ELM control or for the Divertor gas cooling. Then, it is important to know
the efficiency in plasmadization of tritium introduced and the efficiency in transfer of plasmadized
tritium to the burning core together with the ideal burn-up in the core.
Effect of the overall burning efficiency on the tritium balance is discussed in this paper.

161

Poster Session 2(Tuesday 29th November)P2-1
Modification of DOHEAT for optimization of coolant conditions in DEMO blanket
H. Utoh, K. Tobita, Y. Someya, M. Nakamura
Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, JapanFor demonstration of power generation fusion reactors (DEMO), the engineering feasibility ofblanket is a key issue. The difficulty of blanket design is that the blanket has the competing
requirements: (1) fuel self-sufficiency, (2) heat removal, and (3) structural strength against thermal
stress and electromagnetic force. For conceptual blanket design of fusion reactors, a twodimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed [1]. Useof the code has showed outstanding usefulness in the blanket design where detailed evaluation ofneutron flux, nuclear heating rate, tritium breeding ratio (TBR) and the temperature of materials is
required for various blanket concepts and trial-and-error-basis iteration is sometimes necessary.
However, in the previous version of DOHEAT, the coolant temperature is given as an input data.
Actually, a temperature change of the coolant along the cooling tube needs to be determined
implicitly when the inlet temperature, pressure and flow speed are given. The coolant conditions
are the parameters that have effect not only on blanket design, but also on system design including
power generation system. For more systematic design, DOHEAT has been modified. The coolant
condition calculation module was added into the 2-D thermal analysis module, and thetemperature profile in the blanket was provided based on the nuclear heating rate profile andcoolant temperature. A validation calculation indicates that DOHEAT provides reasonable resultson the temperature profile in the blanket.
References[1] H. Utoh et al., Fusion Eng. Des. (2011) in press.

165

P2-2
Removal of deuterium in lithium titanate by gas exposure
A.Tsuchiya, T.Hino, Y.Yamauchi, Y.Nobuta, M.Akiba1), M.Enoeda1)
Laboratory of plasma physics and engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku,
Sapporo, 060-8628 Japan
1)Japan Atomic Energy Agency, 801-1, Mukouyama, Naka, 311-0193, Japan
In the fusion blanket, tritium is produced by nuclear reaction with neutron irradiation. Lithium
titanate (Li2TiO3) pebbles will be employed as tritium breeding material in a water cooled solid
breeder blanket [1], and the tritium generated in and released from the pebbles will be swept andcollected by using sweep gas. As the sweep gas, helium with a little hydrogen gas will be
employed. However, the mechanism on the removal behavior of tritium by the sweep gas has notbeen investigated in detail and must be clarified in order to improve the tritium recovery.
In the present study, Li2TiO3 pebbles were irradiated by deuterium ions in order to simulate thetritium generated in the pebble. The effects on the sweep gas exposure on the removal of
implanted deuterium in the pebbles were evaluated.
The Li2TiO3 pebbles with a diameter of 2 mm were used as a sample. The pebbles were heated at
997K for 1h in a vacuum for degassing. After the degassing, the pebbles were irradiated todeuterium ions by using an ECR ion source. The ion energy was 1.7keV and the ion fluence was
3.0×1018 D/cm2. In order to remove the implanted deuterium, the pebbles were exposed to heliumgas for 1h. The exposure pressure was 0.01 MPa, and the exposure temperature was taken at RT,
373K, 473K, 573K or 773K. After the gas exposure, the amount of residual deuterium in thepebbles was measured by thermal desorption spectroscopy, TDS. The samples were heated fromRT to 973K with a heating rate of 10 K/min and then kept at 973K for 1h. The desorbed gases
which contain deuterium atoms were quantitatively measured by a quadrupole mass spectrometer.
Deuterium retained in the pebbles desorbed in forms of HD, D2, HDO and D2O, and thedesorption of HDO was the largest. The desorbed peak of HDO appeared for 580K with thepebbles after exposure at RT. This peak disappeared for the pebbles after the exposure above
473K. Also, the amount of residual deuterium rapidly decreased with the increase of the exposure
temperature.
For comparison, the amount of residual deuterium in the pebbles after vacuum conservation at
373K was also measured. The residual gas for the vacuum-conserved pebbles was similar to thatwith helium exposure at 373K. These results suggest that helium gas exposure was not effectiveon the removal of deuterium in the present conditions.
The effects on helium gas exposure with a little hydrogen gas at different temperature will be alsopresented.
References[1]Japan Atomic Energy Agency Report, JAERI-Review 2005-011, 2005

166

P2-3
Ion-beam induced luminescence of Er2O3
D.Kato, H.A.Sakaue, I.Murakami, T.Tanaka, T.Muroga, A.Sagara
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Er2O3 electric insulating coating is being developed for reduction of the MHD pressure drop in
Li/V-alloy blanket systems [1]. The Er2O3 coating is a promising candidate also for tritium
permeation barrier of blanket systems made of other breeder and coolant such as Flibe and Li-Pb[2]. However, neutron damages of the coating in fusion reactors are big issues. Ion-beams can beused to introduce defects in sample surfaces to simulate the neutron damages. Changes incrystallinity of the irradiated sample may be inferred from ion-beam induced luminescence
spectra. Tanaka et al. [3] measured the visible luminescence of Er2O3 coating samples by 100 keV
H+ and Ar+ ion-beams irradiation at Osaka Univ. It was demonstrated that the ion-beam induced
luminescence was potentially useful to evaluate changes in crystallinity of irradiated samples.
In the present study, the similar measurements of sintered Er2O3 samples were performed using anapparatus in NIFS for a better understanding about correlations between ion-beam energy
deposition and luminescence spectra. The experimental apparatus consists of an ion-beam source,
a collision chamber, and a CCD spectrometer. The ion source is a part of medium current ion
implantor (ULVAC IM-200MH-FB) used for semiconductor production (Freeman-type). We
could use primary Ar+ beam of about 0.1 - several μA in current at 33, 50, and 70 keV in kinetic
energy. In the collision chamber, a movable stage made of stainless steels was installed to set the
sample disk (25 mmφ × 1 mm). In the present measurements, three luminescence-bands wereobserved in 490-520, 540-580, and 640-690 nm, respectively. Stark components of the band in
640-690 nm are clearly identified as 4F9/2 → 4I15/2 transitions of trivalent Er ions in the Er2O3
crystal. Intensity of the luminescence-band in 640-690 nm increased with ion current density,
while intensities of the other bands decreased as the ion current density increased. Measurements
on stopping power dependence of the spectra are also being planned.
References[1]T. Muroga et al., J. Nucl. Mater. 367-370(2007) 780[2]D. Levchuk et al., J. Nucl. Mater. 367-370 (2007) 1033
[3]T. Tanaka et al., J. Nucl. Mater. (2011) in press
This study is supported by NIFS(UFFF029).

167

P2-4
Electroplating of erbium on steel surface in ErCl3 doped LiCl-KCl
Masatoshi KONDO1), Teruya TANAKA2), Takeo Muroga2), Hiroyuki TSUJIMURA3) andYasuhiko ITO4)
1) Tokai University, 4-1-1 Kitakaname, Hiratuka, 259-1292, Japan
2) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan3) I’MSEP cooperation
4) Doshisha University
The functional layers such as a tritium permeation barrier and an electrical insulation layer coatedon structural materials are the essential technology for liquid breeder blankets of fusion reactors.
One of the candidates is the coating of erbium oxide (Er2O3) layer. The chemical stability of theoxide was extremely high, and this can allow a long-term application in reductive liquid breeders.
Then, the oxide coating on the steel surface has been fabricated with some coating method, and thehydrogen permeation and the compatibility with Li have been investigated. An important issue forthe layers is the improvement of interfacial adhesion strength between the layer and the substrate.
The thermal stress was induced by the difference of the thermal expansion ratio of the oxide layer
and the steel substrate, when the temperature increased or decreased. The stress destroyed theinterface between the layer and the substrate. Then, the peeling off of the layer from the substrate
was promoted. This phenomenon can be improved when the layer has a gradient composition withthe substrate. An electrical plating method can make the gradient composition rather than the
direct coating method of the oxide on the steel substrate. The oxide layer is formed by oxidation of
the surface after the plating of Er metal on the substrate. The metal layer can work as self-healing
layer by the in situ oxidation if some part of the oxide layer was damaged and peeled off from themetal layer in the flowing liquid breeders. The plating technology for the metal, which has
extremely high affinity for the oxidation, was not established so far. The purpose of the present
study is to make clear the feasibility of the erbium- metal plating on the steel surface by anelectroplating method.
The electroplating experiments were carried out in ErCl3 doped LiCl-KCl bath. Three electrodesof a reference electrode, a counter electrode and a working electrode were placed in the molten
salt. The reference electrode and the counter electrode were made of Al-Li and glassy carbon,
respectively. The rectangular shape specimen of 304 type austenitic steel was used as the working
electrode. The temperature of the bath was 450oC. The constant electrical potential of 0.3V
(Li+/Li) was applied on the working electrode. The plating was performed for 1 hour. After theplating, the specimen was taken out from the molten salt. The surface was analyzed using XRD.
The analysis results indicated that the formation of Er metal layer on the specimen surface. The
metal layer was formed due to the electrical precipitation of Er3+, which was made by the
dissolution of ErCl3 in the salt, on the working electrode in the molten salt.

168

P2-5
Microstructure and Mechanical Properties on Oxide Insulator Coating before and after
Thermal Cycling Test
Y. Hishinuma, K. Matsuda1), S. Murakami1), T. Tanaka, T. Nagasaka and T. Muroga
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1) University of Toyama, Toyama, 930-0877, JapanThe oxide coating process is one of the attractive methods of the electrical insulator to restrain of
the for the Magneto-Hydrodynamic (MHD) pressure drop in liquid breeding blanket systems suchas Li and Li-Pb. It is well known that Erbium oxide (Er2O3) was shown to be a high potential
candidate material for electrical insulator coating. Furthermore, Er2O3 coating is also effective to
reduce hydrogen permeation in molten-salt blanket. Er2O3 oxide coating is one of the important
processes to realize an advanced breeding blanket system. We investigated to develop to form
insulator layer on the complicated shape ducts and large-area walls of breeding blanket
components. Recently, we succeeded to form Er2O3 coating layer into large interior surface are ofmetal pipe using Metal Organic Chemical Vapor Deposition (MOCVD) process.
In this paper, we investigated the microstructure and mechanical property of Er2O3 coating layer
on stainless steel 316 (SUS 316) plate before and after heat cycling test with hydrogen permeation.
We found that the permeation reduction factor (PFR) of MOCVD coating layer showed highperformance. From the results of TEM observations, we confirmed that Er2O3 coating layer with
700 nm thickness was formed on the SUS 316 plate and this layer was identified to poly-crystal
phase because the diffraction fleck which was arranged like a ring was observed in the selected
electron diffraction pattern. No macroscopic defects such as crack and peeling in Er2O3 coating
layer were observed before and after permeation test. The change of crystallinity and mechanicalproperties of the Er2O3 coating layer on before and after heat cycling test will be reported indetail.

169

P2-6
Fabrication of the hydrogen recovery unit in the molten salt loop Orosh2i-1:
Operational Recovery Of Separated Hydrogen and Heat Inquiry-1

T. Nagasaka, T. Tanaka, A. Sagara, T. Muroga, M. Kondo1), T. Watanabe2), S. Fukada3),
H. Yukawa4), T. Nambu5), T. Iikubo6)
National Institute for Fusion Science, 322-6, Toki, Gifu 509-5292, Japan
1) Tokai University, 4-1-1, Kitakaname, Hiratsuka, Kanagawa 259-1292, Japan2) The Graduate University for Advanced Studies, 322-6, Toki, Gifu 509-5292, Japan3) Kyushu University, Kasuga, Fukuoka 816-8580, Japan
4) Nagoya University, Furo, Chikusa, Nagoya 464-8603, Japan
5) Suzuka National College of Technology, Shiroko, Suzuka, Mie 510-0294, Japan
Liquid breeder blanket is promising for DEMO and commercial fusion reactors. Molten salt, suchas Flibe (LiF + BeF2) and Flinak (LiF + NaF + KF), is attractive liquid breeder materials becauseof (1) non-reactivity with the air and water, (2) low vapor pressure leading to low operationpressure, (3) low tritium inventory and (4) low MHD resistance. In the liquid breeder blanket
system, the breeder is circulated between the reactor core and the outside as a coolant. The fueltritium and heat generated in the blanket are transferred by the molten salt itself, and thenextracted by a tritium recovery system and a heat recovery system located outside the reactor core.
In order to keep high thermal efficiency for the heat recovery system, the tritium recovery systemshould be operated at high temperature, i. e. 550 oC of blanket outlet temperature.
In order to examine the compatibility between the hydrogen recovery and the heat recovery, an
experiment with a molten salt loop system, Operational Recovery Of Separated Hydrogen andHeat Inquiry – 1 (Orosh2i-1), has been initiated in National Institute for Fusion Science. In thepresent study, a hydrogen recovery unit for Orosh2i-1 was fabricated. The hydrogen recovery unit
consists of a hydrogen permeation tube made from pure Ni. Prior to installation to Orosh2i-1,hydrogen permeation rate was evaluated for the hydrogen recovery unit filled with heliumhydrogen gas mixture.
The feasibility of advanced hydrogen permeation materials, such as pure V, Nb, Ta and their
alloys, for the hydrogen recovery unit is also discussed.
This work was supported by JSPS KAKENHI Grant-in-Aid for Scientific Research (A) 21246141.

170

P2-7
Hydrogen Permeation in Stainless Steel Modified with Surface Nitride by Electrochemical
TechniqueT.Watanabe, A.Sagara1), T.Tanaka1), Y.Edao2), S.Fukada2)
The Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki 509-5292, Japan
1)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan2)Kyushu University, Hakozaki, Higashi-ku, Fukuoka 812-8581,Japan
Molten fluoride salt such as Flibe and Flinak has high potential for liquid blanket in nuclear fusion
reactor [1]. Fluoride salt, however, corrodes structural materials such as steel at high operation
temperature over 500°C. The structural materials require compatibility with fluoride salt. Andleakage of the tritium, which is bred in the blanket system, has to be prevented. In addition, themain materials for heat exchanger have to be as high in thermal conductivity as steel. To begin
with, compatibility with fluoride salt was examined on several materials. It was revealed thatnitride has compatibility with fluoride salt [2]. And a surface modification method with nitride on
metal surface using an electrochemical technique in molten fluoride salt was developed. The
surface modification, which consists of graded composition structure, is expected to be morerobust than such as ceramics coating film against defects such as peeling, crack and pinhole.
Considering in-situ maintenance of structural components in the blanket system, nitriding
treatment in molten salts themselves might be more advantageous than other gas phase nitriding
treatments such as ion nitriding and radical nitriding. Stainless steel 316 (SS316) was successfully
nitrided [3]. A SS316 specimen in LiF-KF-Li3N (49mol% LiF, 49mol% KF, 2mol% Li3N) was
treated by potentiostatic treatment for 100 min at 600 °C and 1.0V vs. Li+/Li. Formation ofmultilayered structure near the surface was examined by EPMA, SEM, XPS and XRD. CrN wasformed from the surface to the depth of 2μm and condensed in the layer. The surface structurechanged drastically in composition and crystal structure with nitrogen diffusion. α-Fex(x>8)N was
also formed as solid solution that nitrogen atoms are inserted between iron atoms. It was formed
from the surface to a depth of 35μm as an inside diffusion layer. Considering tritium permeability,hydrogen permeability is being measured by a build-up method. With surface modification of
nitride on SS316 surface, its hydrogen permeability will be discussed.
References[1] A. Sagara, et al., Nuclear Fusion 45 (2005) 258[2] T. Watanabe, et al, J. Plasma Fusion Res. Ser. 9(2010)342[3] T. Watanabe, et al, Electrochimica Acta, submitted

171

P2-8
The evaluation of heat transfer for the sphere packed pipe by using a molten salt under highheat flux
A.Watanabe, S.Ebara, H.Hashizume, A.Sagara1)
Tohoku University, Sendai, Miyagi 980-8579, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, JapanThe first wall of the helical type fusion reactor FFHR is supposed to be imposed by high heat fluxof 1 MW/m2 at most[1] and to be cooled by a molten salt Flibe. With several attractive features,
however, the Flibe is high Prandtl number (Pr) fluid and has a problem of poor heat transfer
properties. In order to overcome this disadvantage, heat transfer enhancement by a sphere-packed
pipe (SPP) has been proposed and tested for several conditions. Its basic heat transfer
characteristics were evaluated[2][3][4]. For the case of high heat flux above 1 MW/m2 is todetermine empirical correlations between Nu and Re numbers, however, the validity of these
correlations should be verified since there are large temperature change near the heated wall to
induce large changes in the thermal properties like viscosity. In this study, therefore, heat removal
experiment is conducted under the condition of high heat flux of 1 MW/m2 in order to propose aheat transfer correlation equation available under the condition of large viscosity change.
Moreover, fin effect is also evaluated. In the experiment, Tohoku-NIFS Thermofluid loop (TNTloop) is used to circulate a high temperature molten salt (HTS). Heat flux can be imposed up to
about 1 MW/m2 at test section by electrical heating. The test section is a pipe with diameter of
D=14 mm, 20 mm in length and 0.5 mm in thickness filled with pebbles of D/2 diameter made ofaluminum or SS304 whose thermal conductivities is 230 or 18 W/(m·K), respectively. From theexperimental results, there is no significant difference in the heat transfer characteristics between
aluminum and SS304 pebble cases. This means that the fin effect does not appear even in the highheat flux case. In comparison of different heat flux conditions, there is an apparent difference in
the Nu number. It can be considered that this difference is attributed to the change in viscosity due
to the temperature change between the heated wall and bulk fluid. By introducing the ratio ofviscosity at the bulk temperature to that at the heated wall (μb/μw), the heat transfer correlation
equation is modified based on the experimental data.
References[1]A.Sagara et al., Fusion Eng. and Des., 81( 2006) 2703
[2]M.Varahasamy et al. ,Heat Mass Transfer, 39(1996) 3931
[3]T. Satoh et al., Fusion Sci. and Technol., 52(2007)618[4]K.Shimizu et al., Fusion Sci. and Technol., 60(2011)528

172

P2-9
Effects of Argon and Water Plasma Treatments on Philippine Coconut Fibers
J. Guhit, K. Pabelina, H. Ramos 1)
National Institute of Physics, University of the Philippines - Diliman, Philippines
1)Plasma Physics Laboratory, National Institute of Physics
Philippine Coconut fibers (coirs) are organic materials which are naturally hydrophilic in nature.
Low temperature plasma treatment was used to modify the surface and physical properties ofcoirs. Samples were treated with argon and water plasma generated using a Plasma Enhanced
Chemical Vapor Deposition (PECVD) facility. The effect of varied discharge conditions wasinvestigated. Scanning Electron Microscopy (SEM) and Fourier Transform Infrared Spectroscopy(FTIR) were used to determine the physical and chemical changes between untreated and treatedsamples. It was observed that argon plasma treatment lower the hydrophilitic nature of the coirs
and eventually convert it to hydrophobic fibers.
References[1] M. J. John, R. D. Anandjiwala, “Recent developments in chemical modification andcharacterization of natural fiber-reinforced composites,” Polymer Composites Vol. 29(2), pp
187-202, 2008.
[2] P. A. Dos Santos, J. C. Giriolli, J. Amarasekera, G. Moraes,“Natural Fibers Plastic
Composites for Automotive Applications” Sabic Innovative Plastics, 2011.
[3] S. Kalia, B.S. Kaith, I. Kaur, “Pretreatments of natural fibers and their application as
reinforcing material in polymer composites – a review” Polym. Eng. Sci., 49:1253-1272,
2009.
This study is sponsored by the Department of Science and Technology.

173

P2-10Anti-bacterial property of oxygen-ion treated cotton gauze
J. Salamania, M. Villanueva, A. Gines, H. Ramos

The anti-bacterial property of oxygen-ion treated cotton gauze is determined using contact angle
measurements, scanning electron microscopy (SEM), and Escherichia coli (E. coli) adhesion tests.
Cotton gauze samples are irradiated using oxygen low-energy gas discharges using a gas discharge
ion source (GDIS) facility. Ion energies varied by changing the plasma discharge current (Id) from
1mA to 3mA with 1mA intervals. Results show oxygen plasma treatments modify the gauze
surface in morphology. Oxygen treatment shows that as the Id increases, the gauze becomes morehydrophilic. The oxygen treated gauze samples exhibited reduced E. coli attachment on thesamples.
References[1] J.H. Noh, H.K. Baik, I. Noh, J.C. Park, and I.S. Lee, Surf. Coat. Technol. 201, 5097-5101(2007).
[2] T. He, Z.L. Shi, N. Fang, K.G. Neoh, E.T. Wang, and V. Chan, Biomaterials 30, 317-326(2009).
[3] C.I. Pereni, Q. Zhao, Y. Liu, and E. Abel, Colloids Surf. B: Biointerfaces 48, 143-147(2006).
[4] K.L. Mittal (Ed.), Polymer Surface Modification: Relevance to Adhesion, Vol. 4,
VSP/Brill, Leiden (2007).
[5] J.H. Loh, Med. Device Technol. 10 (1), 24-30 (1999).
[6] J.Z. Wu, E.T. Kang, K.G. Neoh, C.Q. Cui, and T.B. Lim, Int. J. Adh. Adhesives 20, 467476 (2008).
[7] H.S. Salapare III, G. Q. Blantocas, V.R. Noguera, and H.J. Ramos, Appl. Surface Sci.
255, 2951-2957 (2008).
[8] H.S. Salapare III, G. Q. Blantocas, V.R. Noguera, and H.J. Ramos, in: Contact Angle,
Wettability, and Adhesion, K.L. Mittal (Ed.), Vol. 6, pp. 207-216, VSP/Brill, Leiden (2009).
[9] H.J. Ramos, J. L. C. Monasterial, and G. Q. Blantocas, Nucl. Instrum. Meth. Phys. Res. B
242, 41-44 (2006).

174P2-11
Sterilization of polyethylene terephthalate (PET) using low-pressure glow discharge H2O2
and O2 plasmas

C. Romero, J. Daseco, K. Pabelina, H. Ramos1)
Plasma Physics Laboratory, National Institute of Physics, University of the Philippines, Diliman,
Quezon City, Philippines
Low-pressure glow discharge hydrogen peroxide(H2O2) and oxygen (O2) plasmas were used tosterilize the Bacillus subtilis and other spores artificially inoculated on polyethylene terephthalate(PET) sheets. A plasma enhanced chemical vapor deposition (PECVD) facility developed at the
Plasma Physics Laboratory, National Institute of Physics (NIP) was used to treat the samples. The
effect of varied discharge conditions on the decimal reduction value (RDV), which is the timerequired to inactivate or destroy 90% of the original population of microorganisms, wasinvestigated. The results showed that the sterilization efficiency was closely related to the plasma
exposure time, gas flow rate and sample temperature. Furthermore, it was found out that placing anegatively biased copper shield around the test substrates, which confined the chemically active
species in the plasma that are believed to be responsible for the inactivation of the bacteria,enhances the effectiveness of the plasma sterilization.
References[1]M. Moisan, J. Barbeau, M. Crevier, J. Pelletier, N, Philip and B. Saoudi, “Plasma
Sterilization: Methods and Mechanisms,” Pure Appl. Chem., 74 (3), 349-358(2002).
[2]M. Laroussi, D. A. Mendis and M. Rosenberg, “Plasma Interaction with Microbes,”
New Journal of Physics, 5, 41.1–41.10 (2003).
[3]A. Soloshenko, V. V. Tsiolko, V. A. Khomich, V. Yu. Bazhenov, A. V. Ryabtsev,
I. A. Schedrin, I. L. Mikhno, “Features of Sterilization using Low-Pressure DCDischarge Hydrogen Peroxide Plasma,” IEEE Trans. Plasma Sci., 30 (4), 14401444 (2002).

175P2-12
Wettability and Antimicrobial Property of Plasma Treated Polyethylene Terephthalate(PET)
K.Pabelina, C.F.Romero, J.A.Daseco and H.Ramos
Plasma Physics Laboratory, National Institute of Physics, University of the Philippines, Diliman,
Quezon City, Philippines
Polyethylene terephthalate (PET) sheets were treated using a plasma enhanced chemical vapor
deposition (PECVD) to enhance the wettability properties and to obtain antibacterial surface
properties. Bacterial adhesion performance of PET materials was tested using contact angle
measurements, scanning electron microscopy (SEM), Fourier transform infrared spectroscopy(FTIR), and adhesion tests. The quantity of Bacillus subtilis adhered onto the different PET sheets
was quantitatively determined using colony forming units (CFU) plate counting in vitro. The
results showed that changes in surface morphology and chemical composition caused by plasma
polymerization and etching mechanism enhanced the bacterial repellence of the polymer surfaces.
References[1]K.Bazakaa, et.al., Acta Biomaterialia, Volume 7, Issue 5, May 2011, Pages 2015-2028
This study is sponsored by the DOST-PCASTRD.

176

P2-13
Ic analysis of Nb3Sn strand cable-in-conduit conductor under the electromagnetic force by
the structural mechanics
T.Yagai1), T.Hamajima2)
1)Sophia University, Chiyoda-ku, Tokyo 102-8554, Japan
2)Tohoku University, Sendai, Miyagi 980-8579, JapanThe Cable-In-Conduit Conductor (CICC) with Nb3Sn superconducting strand is the most popular
conductor for high-field magnet of fusion devices. The CICC is composed of assembled twisted
strands and has over hundreds of strands which get tangled up surrounded by conduit. Although
the mechanical strength of the CICC is high enough against very large electromagnetic force(EMF) during energizing the magnets, the superconducting characteristic such as Ic is sensitive tothe bending distortion and degradation of the performance seems unavoidable. Our approach toanalyze the conductor performance is using structural mechanics based on measured strand traces
inside the conduit. We simulate the strand traces under EMF using beam model which uses
measured traces as initial state. The results indicates the detail information about the degradationof Ic under the large EMF, which is not reported ever before.

177

P2-14
Analysis of contact lengths of strands with Cu blocks in CICC joints
S. Nakazawa, D. Arai, T. Morimura, M. Tsuda, D. Miyagi, T. Hamajima, T. Yagai1),
N. Koizumi2), Y.Nunoya2), K. Takahata3), T.Obana3)
Tohoku University, Sendai, Miyagi 980-8579, Japan
1)Sophia University, Chiyoda-ku, Tokyo 102-8554, Japan
2)Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan3)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
It is observed that measured critical current of Cable-in-Conduit-Conductor (CICC) for ITER TF
coil becomes lower than expected one. This comes from unbalanced current distribution in the
CICC. If a lot of strands carry different currents each other, some strands reach the critical current
faster than others during charging and discharging and can cause degradation of the CICC.
Unbalanced current distribution at steady state is caused through contacting resistance between
strands and Copper (Cu) blocks, saddles or sleeves at CICC joints. In order to evaluate thecontacting resistances, we measure 3 dimensional positions of all strands in the CICC for LHD
OV coil, and then we measure contact parameters such as number and lengths of strands whichappear on surface of the cable contact with the Cu blocks, which are inversely proportional to thecontacting resistance. Then, we developed the numerical code which simulates strand positions in
the CICC, and compare the analyzed contact parameters with measured ones. It is found that theboth results are in good agreement and hence the numerical code is available for evaluating thecontact parameters. We apply the code to various CIC conductor joints, and thereby obtain optimal
joint parameters.

178

P2-15
Measurement Technique of Stress-Strain Properties for Cryogenic Materials by Neutron
Diffraction
Y. Tsuchiya, Y. Ono, H. S. Suzuki, H. Suzuki1), K. Akita1), K. Osamura2)
National Institute for Materials Science, Tsukuba, Ibaraki 305-0047, Japan
1)Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1185, Japan
2)Research Institute of Applied Science, Sakyo-ku Tanaka, Kyoto 606-8202, Japan
Strain of superconductive wire for magnets should be controlled for a good performance of themagnet system because properties of many superconductive materials show deterioration by strain
on wire. Accordingly, strain measurements for superconductive wire are required. As one of themethod of strain estimation on superconductive wire, in-situ neutron diffraction measurement has
been considered by taking an advantage of its possibility to measure the superconductive material
through sheath layer of wire. On the other hand, structural materials for superconductive magnet
are required high strength at cryogenic temperature. Therefore, estimation of stress-strain
properties of the structural materials at cryogenic temperature is of the same importance as that ofsuperconductive wires. In this study, we developed uniaxial tensile machine for cryogenic
temperature test applicable to neutron diffraction measurements in order to low-temperaturetension test for superconductive wire of magnets including low-temperature structural materials[1]. Stress-strain relation of a structural material and a superconductive wire at low-temperature
obtained using the tensile machine will be introduced for a demonstration of the in-situ neutron
diffraction measurement method.
References[1] Y. Tsuchiya, et al., Meas. Sci. Technol. 21 (2010) 025904
This work was partially performed under the Shared Facility Use Program of JAEA (2010A-A38).

179

P2-16Long-Term Monitoring of Hydraulic Characteristics of LHD Poloidal Coils
Kazuya TAKAHATA, Sadatomo MORIUCHI, Kouki OOBA, Toshiyuki MITO,
Shinsaku IMAGAWA
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
We present fourteen-year data summary of hydraulic characteristics of the large helical device(LHD) poloidal coils. The superconductors of the poloidal coils are cable-in-conduit conductors(CICC) cooled by circulated supercritical helium. Because the CICC has narrow cooling channels,
impurities in helium, such as a metal piece, ice of impurity gas and oil from helium compressors,might obstruct the circulated helium flow in the conductor. The flow obstruction results indegradation of stability of the conductor. Therefore, we have continuously monitored a pressure
drop between the inlet and the outlet of each coil since 1998 when the LHD operation started. The
observed pressure drops were converted into dimensionless friction factors to eliminate the effects
of a flow rate and a configuration.
The coils experienced a cool-down, an experimental campaign and a warm-up in every fiscal year.
The mass flow rate of supercritical helium was approximately 50-60 g/s per coil and the Reynolds
number was 2000-3000 during the experimental campaign. The long-term monitoring showeddiscontinuous decreases in friction factor with each experimental campaign over the first sevenyears (the second to the eighth campaigns). That is a change for the better. The friction factors
then remained approximately constant for the next six years (the ninth to the fourteenth
campaigns). The long-term operation of the LHD demonstrates that the initial hydrauliccharacteristics of coils with CICC can be maintained for years. Although the cooling system has
operated smoothly in the most recent campaign (the fifteenth campaign), the observations indicate
an apparent increase in the friction factor.
In the poloidal coils, fine mesh filters were installed at the inlet to trap particles of impurities. The
filters actually trapped impurities during cool-down of the coils, which were confirmed bymonitoring the pressure drop of the filters. A clogged filter were warmed up and evacuated to clear
the impurities. During the past fourteen years, the filter clogged in the first three years and themost recent campaign. Therefore, particles of impurities that passed through the filter during cooldown probably influenced the friction factor in the coil.

180

P2-17
Discussion of heat transfer to liquid helium on surface orientation dependence
A. Iwamoto
National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, JapanFor the stability of pool boiling superconducting magnets, heat transfer characteristics ofconductor surfaces are essential. Heat transfer surface orientation may be varied in a magnet.
Helical coils of the Large Helical Device (LHD) are an example. It is well-known that heattransfer characteristics depend on the orientation.[1] To date, a relatively large copper surface wasprepared, and the dependence of liquid helium heat transfer on surface orientation has been studied
experimentally for the stability analysis of the helical coil.[2] This paper discusses that theexperimental date compares to theory and empirical laws.
References[1] D. N. Lyon, “Boiling heat transfer and peak nucleate boiling fluxes in saturated liquid
helium between the λ and critical temperatures”, Advances in Cryogenic Engineering, Vol.10(1965), pp.371-379.
[2] A. Iwamoto, et al., “Heat transfer from an oxidized large copper surface to liquid helium:
dependence on surface orientation and treatment”, Advances in Cryogenic engineering,
Vol.41 (1996), pp.217-224.

181

P2-18Design Study of a 15 T Test Facility for High-current Superconductors
S. Imagawa, T. Obana, T. Mito
National Institute for Fusion Science
Future fusion power plants need larger scale and higher field superconducting magnets than ITER
magnets, the highest field and conductor current of which are 11.8 T and 68 kA for TF coils and13.0 T and 45 kA for CS coils. It is well known that the sufficient length of testing area isnecessary to examine the superconducting properties of large conductors. Therefore, the existing 9
T test facility in National Institute for Fusion Science [1] is planned to be upgraded to 15 T
solenoid coils with the large cold bore of 0.7 m. The existing 9 T split coils are replaced to a set ofsolenoid coils. A sufficient testing area longer than 1.5 m can be obtained by adopting conductor
samples with coil shape even in the case of one turn coil. The 15 T solenoid coils are designed tobe divided into two parts, which are the inner coil and the outer coil. The larger coil sample can betested at 7 T with only the outer coil.
High coil current density of 70-90 A/mm2 is needed to attain 15 T at sample position, the radius of
0.30 m, under the restriction of the outer diameter less than 1.3 m. The highest field of 7 T issuitable for the outer coil made of NbTi conductors to increase the average current density. As areference design, the magnetic field at sample position is 14.7 T when the center magnetic field
and lengths of inner/outer magnets are 14.0 T and 1.2 m/1.0 m, respectively. The highest field in
the coils is 14.99 T. When only the outer coil is installed, 7.4 T can be generated at the radius of
0.44 m. The total weight of the coils is estimated at 8.0 tons.
Rutherford cables are candidates to attain the high current density. An external protective
resistance is used for the coil protection. In order to suppress the discharge voltage less than 800
V, the operating current is set at 6 kA. The copper current density of the conductor is set at 112
A/mm2 to suppress the hotspot temperature below 200 K at the discharge time constant of 16 s. In
these design criteria, the expected current densities are 65 A/mm2 for the inner coil and 81 A/mm2
for the outer coil, by increasing the load factors, Iop/Ic (the operation current / the critical current)
to 88% and 81%, respectively. These high factors seem marginal for the high field magnets. The
thickness of inside coil cases is 10 mm including the ground insulation, and that of outside case is
30 mm to withstand the strong magnetic force. The detailed mechanical analysis is a future work.
The highest field in the outer coil can be reduced by adopting longer inner coil than the outer coil.
In the case that the length of inner coil is elongated from 1.0 to 1.2 m, the highest field is reduced
from 7.7 T to 7.08 T and the load factor on the loading line of the outer coil is reduced from 87%
to 81%. In this paper the optimization of the coil figure is discussed.
References[1] T. Mito, K. Takahata, N. Yanagi, S. Yamada, M. Sakamoto, A. Nishimura, J. Yamamoto
and O. Motojima, Short sample tests of aluminum-stabilized superconductors for Large
Helical Device, Fusion Engineering and Design 20 (1993) 233-242.

182

P2-19Directory on Recent Information about Operational Experiences on Large Superconducting
System and Cryogenic Facility at Fusion and Particle Accelerator Research Organizations
Satoh, S., Satow, T., Mito, T.1)
Previously National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Large superconducting and cryogenic technologies become dominantly applied in the area of highenergy particle accelerators and magnetic fusion experimental devices. The purpose of the study isto accumulate and construct an extensive and beneficial directory about their operative
experiences from the proceedings of international conferences held during past 20 years. With apreliminary investigation more than 600 papers were collected and each experience as well as
device specifications was tabulated as a data card in the form of Microsoft Excel format whichyielded an extensive directory ( database ) about 91 devices. These are ready for query for key
words with FileMaker Pro. The report talk about the several result obtained so far but the work
must keep continued to expand the database from time to time and expects any collaboration of therelevant researches and engineers to the authors. It is intended to be open before public.
The study is supported under the code no. NIFS11KVXP007 by the 2011-NIFS intercollegiate researchcollaboration promoting program.

183

P2-21
Thermal hydraulic analysis on supercritical helium in cable-in-conduit conductor of helicalcoils for FFHR-d1
S. HAMAGUCHI, T. OBANA, N. YANAGI, S. IMAGAWA, A. SAGARA
National Institute for Fusion Science, Toki, Gifu 509-5292, JapanA design study on the helical-type fusion reactor (FFHR) has been promoted in National Institutefor Fusion Science (NIFS) [1]. In the latest design for the helical DEMO reactor FFHR-d1, themaximum nuclear heating in the helical coils is estimated to be 0.6 mW/cc due to the restriction of
the blanket space. In the case of the continuous helical coils, it has become a major issue from the
aspect of cooling. In the present study, a one-dimensional thermal hydraulic analysis in thelongitudinal direction of the superconductor was conducted in order to evaluate the temperaturerise, and the temperature and pressure of the supercritical helium in the 480-meter-long cable-inconduit conductor (CICC) of the innermost layer of the helical coils with the largest heat load
were obtained under the supposed operational condition.
References[1] A. Sagara et al., Fusion Eng. Des., 83 (2008) 1690

184

P2-22Conceptual design of dc power supplies for FFHR superconducting magnet
H. Chikaraishi, FFHR design group
National Institute for Fusion Science
FFHR is a design study of a fusion power plant, and it will use three sets of superconducting coils
to make the magnetic field to confine the plasma.
The FFHR is not a fusion plasma experimental device, and the requrements to the power system
for superconducting coils are different with the LHD case. For example, the magnet field
configulation will be optimaized and it doesn't require the ability of modification.
At the sametime, the stored energy of FFHR magnet system may be much larger than the LHD
case, so the protection system must be designed to supress the coil terminal voltage or unbulanceof eletro-magnetic force while the current dumping.
Under these requirements, a conceptual design of the power system was studied.
In this design, all of coils are connected in series and only one dc power supply is used for
excitation.
With this circuit configuration, the current ratio between coils is kept constant and the electromagnetic force is balanced eventhou the coils are under the excitation or dumping.

This study is partially supported by NIFS research project(NIFSULAA110,NIFSULAA504)

185

P2-23
Radial build design in the Helical DEMO reactor FFHR-d1

H. Tamura, N. Yanagi, T. Tanaka, T. Goto, J. Miyazawa, S. Imagawa, A. Sagara
and the FFHR Design Group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
FFHR-d1 is a conceptual design of the helical DEMO reactor being developed at the National
Institute for Fusion Science. It has many advantages such as the steady-state nature and lack ofplasma current operation [1]. In the LHD-type configuration, the space between the plasmaboundary and the nearest region of the superconducting helical coil is small especially at the
inboard of the torus. Many components have to be installed in the radial build, and geometrical
positions of the components have to be guaranteed during the operation. They are the first wall,
the tritium breeding blanket, the neutron radiation shield, the vacuum vessel, the thermal shieldwith the vacuum gap, and the coil can for the superconducting helical coil. An estimation of the
distance for the FFHR-d1 is 89 cm which includes the blanket space of 70 cm [2, 3]. Eachcomponent changes its position by the surrounding conditions, such as the mechanical support andelectromagnetic forces, etc. For example, the temperature difference at the superconducting
magnet between the construction phase and the normal operation phase is approximately 300 K,
which causes a thermal contraction of 4 cm. Furthermore, deformation of the magnet by theelectromagnetic force can be 1-2 cm [4]. To guarantee the geometrical position at the normal
operation phase, the following design criteria have been proposed. (1) The breeding blanket andthe radiation shield are divided into several sectors in the poloidal cross-section to prevent hoop
deformation. (2) Thin vacuum vessel sector is attached to the blanket sector and each sector isconnected to each other by using bellows. (3) The blanket is connected to the fixed support nearby
the vacuum vessel so that the displacement does not affects the gap. (4) The thermal shield is
attached to the outer surface of the vacuum vessel. (5) The coil position at the construction phase
is set taking into account the thermal contraction and the electromagnetic hoop deformation. The
detailed geometry and the support method for the radial build components will be introduced in
the presentation.
References[1] A. Sagara, et al., Fusion Eng. Des. 83 (2008) 1690.
[2] T. Goto, et al., in this conference.
[3] T. Tanaka, et al., in this conference.
[4] H. Tamura, et al., presented at EUCAS 2011, The Netherlands.

186

P2-24
Feasibility study on applying cryogenic oscillating heat pipes to fusion magnet
K. Natsume1), T. Mito1)2), N. Yanagi1)2) and H. Tamura2)1)Graduate University for Advanced Studies, 322-6, Oroshi-cho, Toki, Gifu 509-5292
2)National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
Improvement of superconducting magnets for high magnetic field and high heat load is an
important subject to achieve the early realization of nuclear fusion power generation. Applicationof high-temperature superconductors (HTS) to magnets has widely been studied, because HTS
magnets may achieve higher magnetic fields with less operation cost and higher stability against a
coil quench compared to low-temperature superconducting magnets. However, it is difficult toremove the local heat generated in an HTS magnet because the thermal diffusivity decreases as theoperating temperature increases. Therefore, large temperature gradients are easily produced inmagnets, which could cause degradation of superconducting properties and mechanical damagesby thermal stresses.
A new method of including cryogenic oscillating heat pipes (OHPs) in the HTS coil windings as aheat transfer device has been suggested [1]. The OHP is a highly effective two-phase heat transfer
device which can transport several orders of magnitude greater heat flux than the heat conductionof solid metals. In our previous work [2], the performance of cryogenic OHPs were intensively
examined. The typical effective thermal conductivity was observed to be > 10 kWm-1K-1 at 20 K.
The results indicates the possibility of dramatically improving the performance of HTS magnets
by using cryogenic OHPs.
Encouraged by these successful experiments, the feasibility of applying cryogenic OHPs to afusion magnet is examined with a preliminary design of coil windings having a rectangular crosssection (1.8 m wide and 0.9 m high). In the design, HTS superconductors are stacked in layers andcooled by heat conduction from the cooling panels which contain channels for refrigerant andmaintain the temperature at 20 K. Using this structural model, the temperature distribution with
nuclear heating of 1 kWm-3 has been calculated by a finite element thermal model (ANSYS
Release 12.0). In the result, the temperature rise has been limited to 3.1 K when the cooling panel
includes OHPs, which is lower than 6.3 K without OHPs. It is concluded that cryogenic OHPs
enable to eliminate steady-state heating more effectively. Further investigation on thermal analysis
including the transient thermal phenomenon like a magnet quench is also discussed.
References[1] T. Mito et al., Applied Superconductivity IEEE Trans., vol. 20, 2010, pp. 2023-2026.
[2] K. Natsume et al., Cryogenics 51, pp. 309-314 (2011).

187

P2-25
Potential Use of HT-SC Wires in Fusion Reactor Application
Kozo Osamura1), Toshiyuki Mito2), Nagato Yanagi2), Shutaro Machiya3)
and Yoshinori Tsuchiya4)
1) Research Institute for Applied Sciences
2) National Institute for Fusion Science
3) Daido University4) National Institute for Material ScienceAs a next heliotron-type fusion reactor, FFHR-2m2 is now designing under various preconditions,in which a severe condition of maximum magnetic filed of 12 T with stored energy of 160 GJ is
subjected to the magnet technology. In order to realize this large scale fusion reactor, advanced
magnet technology is inevitably required. For maintaining stable and homogeneous current
distribution under operation, a number of twisted SC wires are braided and placed into the metallic
outer envelope, which acts as strengthening member and stabilizer. Thus the conductor
experiences complicated stresses during fabrication process and under operation. Due to achieving
the higher performance, several options are proposed by selecting the SC material, the conductor
type and the cooling method. The use of HT-SC wires generates advantage for designing thefusion reactor. It is possible to operate the magnet with high operation current of 100 kA at a
relatively higher temperature at 20 K.
As candidate of practical HT-SC wires, REBCO coated conductor ( RE = Y, Gd and Sm) and
BSCCO multi-filamentary tape might be considered In the first part of the present paper, it is
attempted to review their general performance on the engineering critical current and its magnetic
field and stress dependences and to discuss the applicability to the high performance fusion
reactor.
The more important performance relates to the mechanical- electromagnetic properties of wire andconductor. In the second part, our recent results on REBCO and BSCCO wires are presented on
the mechanical properties and their temperature dependence, the critical current and their magnetic
field and strain dependences. Further the local strain exerted on the SC element and its external
stress and temperature dependences are reported, which have been directly measured by means ofquantum beam techniques. Based on the experimental results, we discuss the intrinsic strain effect,
the reversible strain limit on critical current under various temperature and magnetic field
conditions and elucidate the influence from the thermal strain and the fabrication condition. As asummary, we point out the key technology to accelerate future progress of their performance andthe potentiality to apply the large scale fusion reactor.

188

P2-26HTS Current Lead Units Prepared by the TFA-MOD Processed YBCO Tapes
Y.Yamada, Y.Ishii, Y.Sakai, H.Motohashi, T.Nakanishi1), Y.Aoki1), Y.Hikichi1),
M.Minowa1), H.Tamura2) and T.Mito2)
Tokai University, Hiratsuka, Kanagawa 259-1292, Japan
1)Showa Cable System Co., LTD., Sagamihara, Kanagawa 229-1133, Japan.
2)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan.
High temperature superconducting current leads have been prepared by the TFA-MOD processed
YBCO tapes. The YBCO tapes are 5 mm in width, about 130 μm in the overall thickness and 190
mm in length cut from the long tape. The YBCO superconducting layer with 1.5 μm in thickness isformed on oxide buffer layers/Hastelloy substrate tape. An Ag layer with 26 μm is deposited on
the YBCO layer to improve the stability and protect the YBCO layer against the moisture of open
air. The critical current Ic of the tapes ranges from 122 A to 152 A at 77 K in self-field, the mean
Ic being 140 A.
A current lead unit is composed of twenty YBCO tapes soldered to Cu caps at both of ends, a
GFRP board and a pair of stainless steel boards as a shunt. The twenty tapes are arrayed in fiverows of stacked four tapes in parallel. The transport current of ten current lead units ranges from
1.6 kA to 2.5 kA at 77 K, the mean current being 2.2 kA. The voltage at Cu end caps is around 0.4
mV at transport current of 2 kA, which corresponds to small resistance of 0.2 μΩ between YBCO
tapes and Cu caps.
The heat leakage of the current lead unit with 150 mm in length between 77 K and 4.2 K is
estimated to be 273 mW. Therefore, the heat load at 2 kA corresponds to 137 W/kA, which isaround one order of magnitude smaller than that of conventional current lead (1.2 W/kA). The
present YBCO current leads with large transport current and small heat load are promising forsuperconducting magnet systems.
References[1]Y. Yamada et al., IEEE Trans. Appl. Supercond., 20(2010) 1714

189

P2-27
Accurate and Stable Numerical Method for Analyzing Shielding Current Density in HighTemperature Superconducting Film Containing Cracks
A. Kamitani, T. Takayama, A. Saitoh1), H. Nakamura2)
Yamagata University, Yonezawa, Yamagata 992-8510, Japan
1)University of Hyogo, Himeji, Hyogo 671-2280, Japan2)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Although several methods have been so far proposed for analyzing the shielding current
density in an high-temperature superconducting (HTS) film, they can be applicable only to an HTS
film containing no cracks [1], [2]. In other words, the method for calculating the shielding current
density has been well-established only for the case where an HTS film does not contain any
cracks. On the other hand, it has remained developing for the case where an HTS film containing
cracks.
In the present study, a numerical method is developed for analyzing the shielding current
density in an HTS film. When an HTS film contains a crack, an additional boundary condition is
imposed on the crack surface and it can be easily incorporated into the weak form. Although the
weak form can be numerically solved with the essential boundary conditions, the resulting solution
does not exactly satisfy Faraday’s law on the crack surface. In order to resolve this problem, thefollowing method is proposed: a virtual voltage is applied around the crack so as to make
Faraday’s law satisfied numerically.
A numerical code for analyzing the shielding current density is developed on the basis of the
proposed method and, by means of the code, the permanent magnet method [3], [4] is investigatednumerically. Especially, the influence of a film edge or a crack on accuracy is assessed. The
results of computations show that, if the magnet is positioned in relation to a film edge or a crack
at a distance equal to three times its radius or less, the accuracy of the permanent magnet method
will be degraded remarkably.
References[1] Y. Yoshida, M. Uesaka, K. Miya, IEEE. Trans. Magn., 30 (1994) 3503.
[2] A. Kamitani, T. Takayama, S. Ikuno, IEEE. Trans. Magn., 44 (2008) 926.
[3] S. Ohshima, K. Takeishi, A. Saito, M. Mukaida, Y. Takano, T. Nakamura, T. Suzuki, M.
Yokoo, IEEE Trans. Appl. Supercond., 15 (2005) 2911.
[4] S. Ohshima, K. Umezu, K. Hattori, H. Yamada, A. Saito, T. Takayama, A. Kamitani, H.
Takano, T. Suzuki, M. Yokoo, S. Ikuno, IEEE Trans. Appl. Supercond., 21 (2011) 3385.
This work was supported in part by Japan Society for the Promotion of Science under a Grant-in-Aid for
Scientific Research (C) No. 21560287. A part of this work was also performed with the support and under the
auspices of the NIFS Collaboration Research Program (NIFS10KNXN178, NIFS11KNTS011).

190

P2-28Microstructure of V3Ga Superconducting Wire Using Cu/V with High Ga Contents
Kenji Matsuda1), Satoshi Murakami2), Tokimasa Kawabata1),
Katsuhiko Nishimura1), Yoshimitsu Hishinuma3), Susumu Ikeno4)
1) Graduate School of Science and Engineering for Research, University of Toyama, 3190,
Gofuku, Toyama, Japan
2) Graduate School of Science and Engineering for Education, University of Toyama, Japan
3) National Institute for Fusion Science, Japan4) Hokuriku Polytechnic College, Japan
Our co-worker, Hishinuma et. al. has established a new route Powder-In-Tube (PIT) process using
a high Ga content Cu-Ga compound in order to improve the superconducting property of the V3Ga
compound wire. In this study, we investigated microstructure of this high Ga content Cu-Ga/V
composite superconducting wire. The different contrasts of matrix, V-Ga phase and Cu-Ga core
were observed by SEM observation in cross section of 19 multifilamentary wire. And V-Ga phase
was confirmed by SEM mapping. The area fraction of V-Ga phase increased when Ga content
increased from 30% to 50%. Thin film sample with V-Ga phase for TEM was fabricated by FIB
and observed by TEM in detail. Selected area diffraction pattern was obtained for V matrix, V-Ga
phase and Cu-Ga core. The ratio of V to Ga for V-Ga phase was probably V3Ga according to the
EDS result. There was a linear interface between V matrix and V-Ga phase, while the interface
between Cu-Ga core and V-Ga phase was not linear. On the other hand, there were some granulargrains observed in V-Ga phase wear Cu-Ga core.

191

P2-29Microstructure and Superconductive Property of MgB2/Al Based Composite Materials
Tokimasa Kawabata1), Daisuke Tokai2), Katsuhiko Nishimura1),
Yoshimitsu Hishinuma3), Shigeki AOYAMA4), Susumu Ikeno5), Kenji Matsuda1)
1) Graduate School of Science and Engineering for Research, University of Toyama, 3190,
Gofuku, Toyama, Japan
2) Graduate School of Science and Engineering for Education, University of Toyama, Japan3)National Institute for Fusion Science, Japan4) Nikkei Niigata, Japan5) Hokuriku Polytechnic College, Japan
MgB2 has the higher critical temperature of superconducting transition (TC : 39K) among theintermetallic compound superconductive materials, however, MgB2 is hard for practical use
because of its unworkable and lower critical current density (JC) in a high magnetic field than Nbbased superconductive materials. We have developed the original method of three-dimensional
penetration casting (3DPC) to fabricate the MgB2/Al composite materials. In the composite
material we made, MgB2 particles dispersed to the matrix uniformly. Thus, these composite
materials can be processed by machining, extrusion and rolling. The TC was determined byelectrical resistivity and magnetization to be about 37～39K. In this work, we made composite
material with ground MgB2 particle with the purpose of extruding thinner wires of composite
material, successfully produced 1mm wire and changed the matrix from pure Al to Al-In alloy. JCof composite materials with the matrix of Al-In alloy was calculated from the width of themagnetic hysteresis based on the extended Bean model. The result was better than that of MgB2/Al
composite material without Indium.

192

P2-30Fundamental evaluation of joint resistance in mechanical butt joint of a stacked GdBCO
conductor
T.Ohinata, S.Ito, H.Hashizume
Tohoku University, Sendai, Miyagi 980-8579, Japan
We proposed a remountable high-temperature superconducting magnet where segments of themagnet can be mounted and demounted repeatedly with mechanical joints, since fusion reactors
usually needs superconducting magnet with complex geometry as seen in the helical system [1, 2].
In our latest study [3], we achieved 1.05 ×10-12 Ω ·m of joint resistance for mechanical butt jointof a stacked BSCCO 2223 conductor. In this study, as a next stage, we experimentally evaluatedjoint resistance of a stacked GdBCO conductor in a strong magnetic field. The GdBCO conductor
consisted of 4 GdBCO tapes and copper jacket. For comparison, we carried out the testing of
direct joint and indium film compliant layer joint using an indium film of 10 μm thick. The
experimental result showed that joint resistance decreases with an increase in joint stress, and theminimum joint resistances were 1.5 μΩ and 1.0 μΩ for the direct joint and the indium film
compliant layer joint. However, the joint resistance of the stacked GdBCO conductor was
significantly larger than that of the stacked BSCCO 2223 conductor obtained in the previous study[3]. The reason could be explained by smallness of cross-sectional area of conductive metal layer
in the GdBCO in comparison with that in BSCCO 2223 tapes since the current may flow only
through the metal layer at the joint region. The effect of the thickness of the metal layer on the
joint resistance was evaluated by current distribution analysis for a butt joint model of the GdBCO
tape. In the model, the tape consists of GdBCO layer, silver layer and copper layer. In addition, avirtual material region is introduced in the joint section. The electrical resistivity of the virtualmaterial was decided based on the normalized joint resistance of the stacked BSCCO 2223
conductor. The current distribution analysis was performed using ANSYS ver. 8.1 and the joint
resistance was evaluated by joule loss of the model. The thickness of the copper layer is chosen as
a parameter in the analysis. The calculated result showed that the joint resistance decreases with anincrease in the thickness of the copper layer from 0 μm to 200 μm, and the joint resistance can be
reduced to 20 % of the joint resistance without the copper layer. Based on the above discussion,
we plan to carry out experiments with the lager conductor.
References[1] H. Hashizume, S. Kitajima, S. Ito, Y. Usui, K. Kitago, A. Sagara, J. Plasma Fusion Res.
SERIES, 5 (2002) 532.
[2] S. Ito, H. Hashizume, Fusion Eng. Des. 81 (2006) 2527.
[3] S. Ito, H. Hashizume, IEEE Trans. Applied Superconductivity, 21 (2011) 1995.

193

P2-31
Measurement of joint resistance of large-current YBCO conductors
Y. Terazaki, N. Yanagi1), S. Tomida2), H. Noguchi1), K. Natsume, T. Mito1), S. Ito3),
H. Hashizume3) and A. Sagara1)
Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki 509-5292, Japan
1)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
2)Kagawa National College of Technology, 355 Chokushi-cho, Takamatsu 761-8058, Japan
3)Tohoku University, 6-6-01-2 Aoba-ku, Sendai 980-8579, Japan
Feasibility studies on applying high-temperature superconductors (HTS) to the heliotron-type
fusion energy reactor FFHR are being carried out [1]. Because the HTS conductor has highcryogenic stability at elevated temperature operations (e.g. 20 K), we consider that refrigeration
power of FFHR can accept Joule heating dissipation in segment-fabricated HTS coils of FFHR.
To examine this scenario, we are carrying out R&D tests with 10 kA-class conductors using
YBCO tapes. In order to connect two conductors, we used two methods: soldered lap joint andmechanical lap joint. The soldered lap joint has a merit of assuring low joint resistance. The
mechanical lap joint of conductors may further ease the construction process of FFHR.
In both methods, we first made two separate conductors having seven layers and two lines ofstacked YBCO tapes (4.3 mm width and 0.19 mm thickness) imbedded in a copper jacket (7 mm ×
10 mm) using solder. We applied a current to the conductor and measured the joint resistance.
From the measured result, the joint resistance is evaluated. For the soldered lap joint, it isdetermined to be 19.1 nΩ and the contact resistivity is 103.5 nΩ-cm2. Because the joint resistanceof the single YBCO tapes that we measured beforehand was 42.5 nΩ-cm2, the present result with a
14-tape conductor is about 2.5 times larger than the expected value.
For the mechanical lap joint, we presently applied clamps to make a connection. The joint
resistance was measured by changing the compressive load. We observed that the joint resistancedecreased as the load increased, however, the minimum value so far was 406 nΩ (4.9 μΩ-cm2),
which is about 45 times larger than that measured for a soldered lap joint. From these results, we
presently consider that the soldered lap joint is more feasible than the mechanical lap joint.
References[1] N. Yanagi et al, Fusion Science and Technology, 60 (2011) 648

194

P2-32
Discussion of structural design issues on a remountable high-temperature superconducting
magnetS. Ito, H. Hashizume, H. Tamura1), N. Yanagi1), S. Imagawa1), A. Sagara1)
Tohoku University, Sendai, Miyagi, 980-8579, Japan
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
A remountable (demountable) high-temperature superconducting (HTS) magnet was proposed fora helical reactor having huge and complex superconducting magnet in the long term [1] and for acomponent-testing machine of tokamak which require good access in the near and intermediate
term [2]. The proposed magnet consists of some segments that can be assembled and disassembled
repeatedly with mechanical joints (electrical joints). The superconducting material of the magnet is
HTS having high critical current and high heat capacity at relatively high operating temperature(>20 K), which could afford mechanical joints. Refrigeration power for the remountable HTS
magnet could be lower than conventional low-temperature superconducting magnet due to thehigh operating temperature though there exists resistive loss at joint section. The design cancontribute to simplify the fabrication of a complex and huge magnet like helical coils and to repair
or replace damaged components such as by neutron irradiation. This concept is also included in
the HTS coil option for the LHD-type fusion DEMO reactor FFHR designed by National Institutefor Fusion Science [3, 4].
In previous studies [5, 6], R&D of a mechanical joint of a HTS conductor have been performed
successfully. Conceptual design of a remountable HTS magnet using structural analysis is also
another important issue. In an actual magnet, electromagnetic forces and thermal strain can
influence joint condition. In this study, we performed structural analysis using a model of aremountable circular coil as a first step of the design. In the analysis, we evaluated contact
condition at the joint section with a change in radial electromagnetic force for several shapes ofjoint section. Based on the results, we discuss about structural component of the remountable
magnet. Then design issues are considered based on the discussion.
References[1] H. Hashizume, S. Kitajima, S. Ito, Y. Usui, K. Kitago, A. Sagara, J. Plasma Fusion Res.
SERIES, 5 (2002) 532.
[2] L. Bromberg, H. Hashizume, S. Ito, J.V. Minervini, N. Yanagi, Fusion Sci. Tec., 60(2011) 635.
[3] G. Bansal, N. Yanagi, T. Hemmi, K. Takahata, T. Mito, A. Sagara, Plasma and Fusion
Res. 3 (2008) S1049.
[4] N. Yanagi, K. Nishimura, G. Bansal, A. Sagara, O. Motojima, Plasma and Fusion Res. 5(2010) S1026.
[5] S. Ito, H. Hashizume, Fusion Eng. Des. 81 (2006) 2527.
[6] S. Ito, H. Hashizume, IEEE Trans. Appl. Supercon., 21 (2011) 1995.

195

P2-33Beam-ion losses due to magnetic field ripple under various plasma parameter ranges on the
Large Helical Device
Kunihiro Ogawa1), Mitsutaka Isobe, Kazuo Toi, LHD experiment group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)Nagoya University, Nagoya, Aichi 464-8603, Japan
Magnetically confined fusion devices such as tokamaks and helical devices/stellarators have beensignificantly developed, approaching the conditions of nuclear fusion reactor. Compared withtokamak plasmas, helical/stellarator plasmas have intrinsically a great advantage in its easy
controllability. One of the important issues to achieve the high plasma parameter is how well fastions would be confined. Obviously, good confinement of D-T produced alphas is essential tosustain the fusion plasma. In addition, auxiliary heating produced by fast ions such as a neutralbeam injection (NBI) is also needed to ignite the fusion plasmas. If they are not confined well,then their energy is lost from the plasma, meaning that the efficiency of heating is lowered. Also,
concentrated fast-ion losses might cause localized damage of plasma facing components. For thesereasons, it is indispensable to understand the confinement/loss characteristics of fast ions.
Beam-ion losses have been measured with scintillator-based lost-fast ion probe installed on theoutboard side of the Large Helical Device. The loss flux reaching the probe with information of
their pitch angle (χ and the energy (E) distributions were investigated in various plasma
parameters. In relatively low-field experiments (<1 T), not only the losses having a transition orbit(the range of E and χ ?are about up to 180 keV and 55 degrees), but also the losses having apassing orbit (the region of E and χ ??are about up to 180 keV and 30 to 40 degrees) are regularly
observed because of the large deviation of the orbit from the flux surfaces. Those loss fluxes are
significantly suppressed as the toroidal field magnetic strength (Bt) is increased. Losses having thetransition orbit are still remained on relatively-high-field experiments (Bt/Rax=1.375 T (CCW)/3.85m) with no strong instability. The loss flux is proportionate to fast-ion density ~ PNB×τse, where
PNB, τse, Raxindicate the deposition power of NBIs, a slowing-down time of beam ions byelectrons, and a magnetic axis position in vacuum, respectively. It is clearly shown that collisionalloss coming to the probe depends on the fast-ion density.
This work is supported in part by the Grant-in-Aid for Scientific Research from MEXT, No. 16082209 and from
JSPS No. 21360457, No. 21340175, and No. 22-7912, and the LHD project budget (NIFS09ULHH508).

196

P2-34
Evaluation of the non-axisymmetry of the central cell plasma by using a segmented limiteron GAMMA 10
T. Iwai, M. Ichimura, M. Hirata, R. Ikezoe, T. Yokoyama, T. Sato, Y. Ugajin, T. Iimura, Y. Saito,
Y. Nakashima, K. Hosoi, T. Imai
Plasma Research Center, University of Tsukuba, 1-1-1 Tennodai, Tsukuba 305-8577, Japan      In the GAMMA10 tandem mirror, electromagnetic waves of two different frequencies in theion-cyclotron range of frequency (ICRF) are used for plasma production and ion heating. For themagneto-hydrodynamic (MHD) stabilization of the whole plasma, ICRF waves are also used at theanchor cells, which are composed of minimum-B field and located at both ends of the central cell.
Waves in the electron-cyclotron resonance (ECR) frequency are used for the electron heating andthe formation of the plasma potential along the magnetic field line. A segmented limiter which is
divided into 8 equal parts in the azimuthal direction is installed near the midplane of the central
cell. The diameter of the limiter is 36 cm. The segmented elements are electrically-insulated fromthe ground. Thus, the floating potentials and its azimuthal distribution in the peripheral region canbe measured. By connecting all segmented elements electrically, we can change the boundary
conditions in the peripheral region. Two types of low-frequency fluctuations which degrade theplasma parameters are detected in GAMMA10. These are drift-type and flute-type fluctuations. It
is clearly observed that the amplitude of the drift-type fluctuation with low-azimuthal mode
numbers becomes small with the connection.
     The azimuthal distribution of the floating potential is a good indication of non-axisymmetry of
the plasmas. Non-axisymmetry of the plasmas in a simple mirror configuration will cause
diffusion in the radial direction and degrade plasma parameters. In this paper, non-axisymmetry of
the plasmas in the central cell of GAMMA 10 is evaluated with the azimuthal distribution of thefloating potentials on each element of the segmented limiter. When ICRF and ECR are used forthe experiments, the azimuthal distribution of the floating potentials depends on their injecting
powers. We measure the influence of ICRF and ECR on non-axisymmetry of the plasma. The
mode numbers of the floating potential in the azimuthal direction are analyzed. Drive mechanisms
of non-axisymmetry of the GAMMA 10 plasmas are investigated.

197

P2-35Magnetic Diagnostics for Magnetic Islands
Y. Suzuki, S. Sakakibara, K.Y. Watanabe, Y. Narushima and LHD Experiment Group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Generating and keeping clear flux surfaces are an aim of magnetic confinement researches,
because magnetic islands and stochasticity of magnetic field leads the degradation of theconfinement connecting and overlapping filed lines. In tokamaks, the degradation of theconfinement due to generating islands like the locked mode and neoclassical tearing mode (NTM)
were observed and studied. In addition, same degradation is also observed in helical system. Thus,
understanding and controlling of island dynamics are urgent and critical issues to aim the fusion
reactor.
A method to identify magnetic islands is the magnetic diagnostics. Since the magnetic diagnostics
detects the change of magnetic flux directly but the diagnostics must be installed appropriately todetect perturbed field of islands. In this study, we design the magnetic diagnostics to study island
dynamics in LHD plasmas. To design the magnetic diagnostics, we apply 3D MHD modelingcodes. A special attention is the identification of magnetic islands with m=3 and 4 modes.

198

P2-36Time-dependence of Balmer series emission intensity during pulse plasma flowin divertor simulator TPD-SheetIV
N.Mine, T.Iijima, H.Komukai, A.Tonegawa, K.Kawamura, K.Sato1)
Department of Physics, School of Science, Tokai University, 411 Kitakaname, Hiratsuka,
Kanagawa, 259-1292, Japan
1)Chubu Electric Power, 1, Higashishinmachi, Higashi-ku, Nagoya, 461-8680, JapanThe research on dynamic behavior of plasma during pulse plasma flow is a topic in space plasma,
plasma processes and nuclear fusion. In particular, transient behavior from recombination toionization processes during pulse plasma flow has become important for divertor region of nuclear
fusion device. The transient behavior during pulse plasma flow has been studied by observing theshort double minimum (negative) spike in Dα emission intensity from the plasma[1]. This responsein Dα emission is explained by electron temperature increase associated with bursts of heat andparticles along the magnetic field on pulse plasma flow. In order to understanding of effects ontransient behavior during pulse plasma flow, however, it is necessary to take into account of highenergy electrons in plasma flow.
We have carried out the experimental observation and modeling of time-dependence of Balmer
series emission intensity in hydrogen recombination plasma in a linear plasma device, TPDSheetIV[2]. Pulse plasma flow generated by controlling the TPD-type plasma source wereintroduced into hydrogen recombination plasma in the experimental region. The pulse plasma flow
was generated by the switching circuit controlled the electric potential of the next floating
electrode of the anode in plasma source. The duration of the pulse plasma was 0.3ms and thefrequency of cycle of the plasma was 50Hz. The time-dependence of electron density ne, electron
temperature Te, electron energy distribution function fe(E) were measured using Langmuir probe.
The ionization and recombination events are discussed by Collisional-Radiative model, taking into
account of high energy electrons.
After the pulse plasma flow in the recombination plasma, the high energy electron was generated
in the electron energy distribution function, immediately. Also, electron temperature Te andelectronic density ne have increased. At the same time, double minimum (negative) spike wereobserved in Hα emission intensity from the plasma.
References[1]A.Loarte, et al., Nucl Fusion 38(1998)331[2]A.Tonegawa, J.Nucl.Mater., 313-316(2003)1046

199

P2-37
Calculation of geometry matrices for IRVBs for application to 3D tomography of radiative
phenomena in LHD
B.J.Peterson, M.Kobayashi
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
InfraRed imaging Video Bolometers (IRVBs) can provide hundreds of channels of bolometric data
forming an image of the plasma radiation [1]. By calculating the geometry matrix (or response
matrix) of the detector field of view (FoV) with respect to a predefined three dimensional (3D)
plasma grid of plasma voxels these geometry matrices can be used in two ways. The first of theseis to compare the 3D carbon radiation results of an impurity transport code such as EMC3EIRENE [2,3], which are translated via the geometry matrix into the field of view of the IRVB,with the experimental results of the IRVB. This type of analysis is called a synthetic instrument (orsynthetic diagnostic). The second technique is to divide the LHD plasma into half field periods(18° in toroidal angle) and assume that the plasma reproduces itself every half field period. This
assumption is called periodic helical symmetry. Then by combining the FOV of multiple imaging
bolometers with different views of the plasma (top, bottom, tangential, semi-tangential, radial,etc.), one large geometry matrix can be derived relating thousands of IRVB channels to thousandsof plasma voxels. Then by applying tomographic techniques the geometry matrix can be inverted
and applied to the measured data to yield the 3D radiation distribution, which can be directlycompared with the 3D carbon radiation distribution from the EMC3-EIRENE code. The advantageof the synthetic diagnostic technique is that it does not require the assumption of periodic helicalsymmetry and can be applied to nonsymmetric phenomena such as those involving the m/n =1/1
magnetic island.
References[1] B. J. Peterson et al, Plasma Fusion Res. 5 (2010) S2095.
[2] Y. Feng et al., Contrib. Plasma Phys. 44 (2004) 57.
[3] D. Reiter et al., Fusion Sci. Technol. 47 (2005) 172.

200

P2-38Design considerations for an Infrared Imaging Video Bolometer for observation of 3D
radiation structures of detached LHD plasmas
S.N.Pandya 1), B.J.Peterson 2)
1) Institute for Plasma Research, Nr. Indira Bridge, Village Bhat, Gandhinagar – 382428, India
2) National Institute for Fusion Science,322-6 Oroshi-cho, Toki, 509-5292, Japan
Plasma detachment is an important operational regime for future diverted magnetic fusion devices
where the power load on the plasma facing components ( PFCs ) is expected to be tremendously
high. The detachment of the plasma will help by temporarily reducing the peak power loads on the
PFCs and thus will facilitate in maintaining the power loads within the engineering design limits.
Plasma detachment in LHD is aided by the external addition of an m/n=1/1 magnetic perturbation[1]. During detachment the radiation structures are found to be localized near the X- points of themagnetic islands [1,2]. Infrared Imaging Video Bolometers ( IRVB ) are successfully being used
to study the 2D impurity radiation profiles from the LHD plasma [2,3]. IRVBs can serve as apromising diagnostic for studying the radiation structures of detached plasmas in LHD and hence acomparison can be established with theoretical models. A new IRVB system is being designed forthe bottom port 6.5-L for better access to the magnetic X – points and thus studies of the 3D
radiation structures. This presentation gives a design overview of this new IRVB system for thebottom view. The design includes magnetic and neutron shielding for the IR camera, IR opticaldesign, spatial resolution and field of view of the IRVB, sensitivity and signal to noise estimates.
The design configurations and parameters will be reviewed and discussed.
References[1]M. Kobayashi et al 2010 Phys. Plasmas 17 056111[2]E.A. Drapiko et al 2011 Nucl. Fusion 51 073005[3]B. J. Peterson et al 2001 Rev. Sci. Instrum. 72 p923

201

P2-39Design and Development of Infrared Imaging Video Bolometer for ADITYA Tokamak
Santosh P. Pandya, Shwetang N. Pandya, J. Govindarajan and ADITYA Team
Institute for Plasma Research, Bhat, Gandhinagar – 382 428 (INDIA)
Infrared Imaging Video Bolometer (IRVB) is one of the modern plasma diagnostic which provides
the measurement of the temporally resolved two & three dimensional (2-D / 3-D) power profile
radiated from the plasma devices [1-4]. It utilizes free standing ultra thin large area metal foil,
which absorbs radiated power from high temperature magnetically confined plasma throughpinhole camera geometry in wide spectral range (SXR to UV). This absorbed power alters thetemperature profile on the foil area which is optically (by infrared radiation) imaged and measured
by the Infrared Camera seating out side the vacuum vessel through Infrared Transmitting vacuum
view port. Using suitable analysis of 2-D temperature profile, total radiated power profile fromplasma can be determined. Since the IRVB is optically transmits the signals (Infrared Radiation) it
is immune to EM noise and also can work under reactor relevant conditions, hence it is moresuitable for the steady state high temperature plasma devices and future fusion devices [3-5].
The techniques is being first time utilize for the medium size tokamak Aditya (R=75cm, a=25cm,
Ip=75 kA, Te(0)~350 eV, BT= 0.7T), where the plasma shot duration is ~100ms and total radiatedpower 40kW to 80kW typically during current flattop[6]. IRVB system was design, developed andinstalled on the Aditya tokamak having 8 mm diameter pinhole, 2.5 micron thick Platinum foil
(Size: 64mmx64mm), MWIR-camera (320x240 pixels, 200Hz frame rate, NETD~20mK), 8x8Bolometer pixel array, 10 ms optimal time resolution. This paper describes the design, calibration
experiments and initial total radiated power measurement results from the Aditya Tokamak.
References[1] B. J. Peterson, Rev. Sci. Instrum., Vo.71, No.10, p3696 (2000)[2] B. J. Peterson, et. al. Rev. Sci. Instrum., Vo.72, No.1, p923 (2001)[3] Byron J. Peterson, et. al. Rev. Sci. Instrum. 79, 10E301 (2008)[4] N. Ashikawa, et. al. Journal of Nuclear Material, 313-316 (2003),pp1103-1108.
[5] B. J. Peterson, et. al. Journal of Nuclear Material, 363-365 (2007),pp412-415.
[6] K. Tahiliani, et.al. Plasma Phy. Control. Fusion. Vol.51, 085004 (2009)202

P2-40Calibration of IR imaging bolometer foil by laser irradiation
R. Sano, B. J. Peterson1), E. A. Drapiko1), D. C. Seo2), Y. Yamauchi, T. Hino
Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo, 060-8628, Japan
1) National Institute for Fusion Science,322-6 Oroshi-cho, Toki, 509-5292, Japan
2) National Fusion Research Institute,Daejeon305-806, Rep. of Korea
An IR imaging video bolometer (IRVB) [1] is a measurement instrument for plasma radiation, anduseful for the measurement of both radiation intensity and spatial distribution. A bolometer foilwith a carbon coating is employed in the IRVB system. The temperature distribution on the foil
formed by the incident radiation is measured, and then the radiation profile is determinedfrom thetemperature distribution by using the calibration data for the foil.
Thus, the thermal response of the foil to the incident power has to be examined and the radiation
power has to be calibrated with the temperature rise. The temperature distribution depends on theproperties of the foil such as thickness and emissivity. The foil properties are often non-uniform
across the foil. This non-uniformity introduces errorin the radiation measurement. The carbon
coating also influencesthe temperature distribution.
In the present study, the bolometer foil was irradiated with a He-Ne laser while changing theirradiation position by moving the foil in two dimensions [2]. The spatial distribution of thetemperature was measured by using an IR camera. The temperature distribution measured wascompared with that calculated by a Finite Element Method (FEM) analysis. In the FEM analysis,
the thickness [3] and the emissivity at each position were repeatedly changed until the FEM profile
converged to the measured profile. The FEM profile converged to the temperature
profilemeasuredafter only two iterations. In the present study, it is found that the FEM profile
agreed wellwith the temperature profile measured, so that the calibration between the radiation
power and the temperature profile can be suitably conducted by using the FEM analysis.
References[1] B.J. Peterson et al., Rev. Sci. Instrum. 74(3) (2003) 2040.
[2] E. Drapiko et al., Rev. Sci. Instrum. 81 (2010) 10E116.
[3] H. Parchamy et al., Rev. Sci. Instrum. 77 (2006) 10E515.

203

P2-41Development of an Array System of Ultra-Soft X-ray Detectors with Large Sensitive Area on
the Large Helical Device
X.D. Du1), S. Ohdachi1),2), K. Toi2) and LHD Experiment Group2)
1) Department of Fusion Science, the Graduate University for Advanced Studies
2) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi 509-5292, JapanA new 17-channel soft X-ray diagnostic system was developed for study ofmagnetohydrodynamics (MHD) fluctuations and installed on the Large Helical Device (LHD).
The Absolute X-ray Ultraviolet Photodiodes (AXUV diode made by IRD) with the large
sensitivity area 10 mm &x 10mm was adopted as the detectors. The sightlines were designed tocover the whole plasma with 3.8 cm space separation and the expected radial resolution was 10cm.
The toroidal elongated pin hole (12mm &x 36mm) was designed to increase the signal to noise
ratio and a Be foil with 15 μm thickness was used to shut the visible light. The detector array wasplaced inside the LHD vacuum vessel, being shielded by an aluminum box. Electromagnetic
noises were effectively reduced by this arrangement.
In the experimental campaign of LHD, this fiscal year 2011, various kinds of MHD fluctuations
excited in core and edge plasma regions have clearly been detected by this newly installed
diagnostic system. The internal disruption, which would be induced by resistive interchange modewith m=2/n=1 (m, n: poloidal and toroidal mode numbers, respectively), was observed. The
sawtooth–like crash occurs around the q=2 rational surface.

204P2-42
Tangential and vertical Soft-X ray imaging for three-dimensional structural studies in the
RFP
A. Sanpei, K. Oki, D. Fukahori, K. Deguchi, S. Nakaki, H. Himura, S. Masamune, S. Ohdachi1),
N. Nishino2), T. Onchi3)
Kyoto Institute of Technology, Kyoto 606-8585, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
2)Hiroshima University, Hiroshima 739-8527, Japan3)University of Saskatchewan, Saskatchewan S7N 5E2, Canada
Three-dimensional (3-D) effects on MHD phenomena in axisymmetric systems such as reversed
field pinches (RFPs) have attracted much attention. In the RFP, for example, recent progress haveshown the importance of helically deformed RFP configuration where single helical magnetic axis
state is self-organized. Since the toroidal mode number n of the dominant m=1 mode is muchlarger than 1, development of imaging diagnostics for 3-D structure is required for the detailed
study on dynamics of the self-organizing process. In a low-A RFP machine RELAX (R = 0.51 m/a= 0.25 m (A = 2)), a quasi-periodic transition to quasi-single helicity (QSH) state has been
observed. We have applied a soft-X ray (SXR) pin-hole camera and a ICCD camera to taketangential SXR snapshots of the RFP during the QSH state, identifying characteristic helical SXR
structures which suggest hot or dense helical core.[1] Moreover, we have constructed a fast
successive SXR imaging system where SXR camera and high-speed camera. As a preliminary
experiment, we have taken tangential SXR pin-hole pictures with time resolution of 10 micro sec,
to identify time evolution of helix structure in RELAX plasmas.[2,3] As a next step, we have beendeveloping a SXR imaging diagnostic system, which uses multiple SXR cameras together withhigh-speed cameras to take time evolution of SXR images from tangential and vertical directions
simultaneously for the study of dynamic structures of 3-D SXR emissivity, through which we
expect to discuss 3-D dynamics of MHD instabilities associated with the QSH state. Initial results
will be reported, together with some discussion on 3-D reconstruction.
References[1] T. Onchi et al., Rev. Sci. Instrum., vol.81 (2010) 073502[2] A. Sanpei et al., IEEE Transactions on Plasma Science (2011) TPS.2011.2160369[3] A. Sanpei et al., Plasma and Fusion Res., vol.6 (2011) 2406096
The study is performed with the support of the NIFS Collaborative Research Program (NIFS10KOAP024).

205

P2-43
Identification of spurious modes of high power 77 GHz gyrotron
for collective Thomson scattering in LHD
S. Ogasawara1), S. Kubo1,2), M. Nishiura2), T. Saito3), Y. Tatematsu3), K. Tanaka2),
T.Shimozuma2), Y. Yoshimura2), H. Igami2), H. Takahashi2), S. Ito2), Y. Takita2),
S. Kobayashi2), Y. Mizuno2), K. Okada2), R. Minami4), T. Kariya4) and T. Imai4)
1)Department of Energy and Technology, Nagoya University, Nagoya 464-8463, Japan2)National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, 509-5292, Japan
3)Research Center for Development of FIR, University of Fukui, Fukui 910-8507, Japan
4)Plasma Research Center, University of Tsukuba, Tsukuba 305-8577, Japan
Collective Thomson scattering (CTS) diagnostic in the Large Helical Device (LHD) has beendeveloped as a technique for measuring bulk and fast ion velocity distribution functions in highdensity and high temperature plasmas. For the bulk and tail ion temperature measurements in the
LHD, the CTS diagnostic has been performed using existing electron cyclotron resonance heating(ECRH) system with high power mega-watt gyrotrons with the frequency 77 GHz [1, 2]. In orderto improve the signal to noise ratio for CTS measurement, the power of the CTS probe beam ismodulated to subtract the background electron cyclotron emission (ECE) and system noise fromreceived signals. During the turing on/off phase of the modulation, the modulation of gyrotron
power excites spurious mode radiations. They exist in the measured CTS spectrum. They evenwith small power result in the saturation of the IF amplifiers if the frequency is in the sensitive
band but outside of the notched one.
As the first step, we measured the radiation frequency of a gyrotron with the nominal frequency of
77 GHz using a heterodyne receiver and a fast sampling oscilloscope. We found that spurious
mode frequency was about 74.7 GHz when the anode voltage of the gyrotron started up. The
spurious mode is considered to be the TE17,6 from the resonator structure of the gyrotron. The
starting current calculation code [3] shows that the TE17,6 mode is relatively easy to start
oscillation as compared with the other modes under the measured condition. The method tosuppress or reduce the radiation of the spurious mode will be discussed.
References[1] M. Nishiura, S. Kubo et al., Journal of Physics: Conference Series 227, 012014(2010)[2] S. Kubo, M. Nishiura et al., Rev. Sci. Instrum. 81, 10D535 (2010)[3] Y. Tatematsu, T. Saito, J. Plasma. Fusion. Res. Vol.84, No.12 (2008)206

P2-44
Coaxial multiple laser beam combiner for the LHD Thomson scattering system
R. Yasuhara, I. Yamada, K. Narihara, H. Funaba, H. Hayashi
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Thomson scattering (TS) system is a basic diagnostic of LHD, because the electron temperatureand density profiles can be simultaneously measured. The LHD TS system has been designed tohandle multiple Nd:YAG lasers for improvement of time resolution or accuracy of measurementby packing mirror method [1]. This method works well in the focal point of near center of plasma
but outside this region the laser beams gradually separate. As the result sampled scattering lightfrom separated laser beam line have the different information of plasma volume.
A coaxial multiple laser beam combiner for LHD TS system have been developed to solve thisproblem. Multiple beams from different aperture of laser gain mediums are combined to a coaxial
beam line in time series by a combination of the cubic polarizer and an electro optical (EO) device[2]. One cubic polarizer and a half wave plate are needed for two beam combing. A horizontal
polarized laser beam passes through the cubic polarizer. On the other hand laser beam polarizationfrom the second laser is rotated to vertical by half wave plate and reflected by cubic polarizer.
After the refraction and the transmission of cubic polarizer, two different laser beams are aligned
at coaxial beam line. For additional multiple lasers, EO device is added after the first cubic
polarizer for preparing the polarization for use of the next cubic polarizer.
We have installed this beam combiner for LHD Thomson scattering system from 15th experimentcampaign and confirmed the improvement of data quality. In the presentation, we will discuss thedetailed design and performance of polarization based laser combining.
References[1] T. N. Carlstrom et al., Rev. Sci. Instrum. 63, 4901 (1992).
[2] M. J. Walsh et al., Rev. Sci. Instrum. 75, 3909 (2004).

207

P2-45The Thomson Scattering Diagnostic System on the KSTAR tokamak
J.H. Lee1), S.T. Oh1), H.M. Wi1), W.R. Lee1), K.P.Kim1), W.H. Ko1), K.D.Lee1), J.G. Kwak1),
M. Kwon1), I.Yamada2), K. Narihara2), K. Kawahata2) and KSTAR Team1)1)National Fusion Research Institute,113 Gwahangno, Yeseong-gu, Daejeon, 305-333, Korea
2)National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanThe Thomson scattering system of the KSTAR tokamak is described. The KSTAR Thomson
scattering system has been developed and tested in the KSTAR 3rd campaign, 2010. Through thiscommissioning, we could measure the Thomson scattering signal successfully in the KSTAR 4th
campaign, 2011. Last year we found that the main reflection mirror had an errors near theThomson signal wavelength region thus before the KSTAR 4th campaign we changed main
mirrors in the core, edge collection optic module. In this campaign, we installed the 1000μm
diameter single core silica/silica type optical fibers for transfer the Thomson signals and Charge-to-Digital conversion type VME modules for measuring core and edge region plasma parameters. To
suppress the noise from the signal, we installed an uninterruptible power supply (UPS) system.
From this system we measure the electron temperature and density profiles on the KSTAR plasma.
This paper reports the experimental results obtained from KSTAR Thomson scattering diagnostics
system and discuss the future works.
This work was supported by the Ministry of Education Science and Technology of Republic of Korea.

208

P2-46The comparison of electron temperature and its density calculation methods in Thomson
scattering diagnostics
S. Oh, J. H. Lee, H. Wi
National Fusion Research Institute, National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon 305-333, South Korea
The Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak [1] achieved thefirst plasma on the last day of June 2008 at the National Fusion Research Institute (NFRI),
Daejeon, Korea. Thomson scattering has become an important diagnostic for measuring electron
temperature and its density profiles in most tokamaks [2] [3] [4].
In Thomson scattering diagnostics, electron temperature and its density are indirectly derived
through calculations of scattered light signals.[5] The signal noise is influenced by the laser powerfluctuation, and the final signals get out after couples of signal conversions such as a conversion of
optical signal to electrical signal or internal electrical signal conversions. The noise can cause
great effects on the calculations. In the calculations, χ2 method is mostly used. Here the robustnessof χ2 method to the noise is examined and an alternative method is considered.
References[1] G. S. Lee, J. Kim, S. M. Hwang, C. S. Chang, H. Y. Chang and M. H. Cho et al, Nucl.
Fusion 40 (2000) 575[2] N. J. Peacock, D. C. Robinson, M. J. Forest, P. D. Wilcock and V. V. Sannikov, Nature
224 (1969) 488[3] H. Murmann, S. Gotsch, H. Rohr, H. Salzmann, and K. H. Steuer, Rev. Sci. Instrum. 63(1992) 4941
[4] T. N. Carlstrom, G. L. Campbell, J. C. DeBoo, R. Evanko, J. Evans, C. M. Greenfield, J.
Haskovec, C. L. Hsieh, E. McKee, R. T. Snider, R. E. Stockdale, P. K. Trost, and M. P.
Thomas, Rev. Sci. Instrum. 63(1992) 4901
[5] J. Sheffield, Plasma Scattering of Electromagnetic Radiation, New York: Academic(1975)

209

P2-47
Demonstration of an in-situ relative calibration method for a Thomson scattering diagnostics
on TST-2
H.Tojo, A.Ejiri1), J.Hiratsuka1), T.Yamaguchi1), Y.Takase1), K.Itami
Japan Atomic Energy Agency, 801-1 Mukoyama, Naka 311-0193, Japan
1)Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa 277-8561, Japan
Development of in-situ calibration methods for Thomson scattering diagnostics is an essentialtask. However, intense radiation (i.e., neutron and γ-ray) from high performance plasmas changes
spectral transmissivities of optical components and degrades accuracies of electron temperature(Te). In order to determine Te with unknown or degraded sensitivities (transmissivities), a method
using double pass scattering system, where two scattering signals with different scattering angles(θ) can be measured, has been proposed [1]. Since scattered spectra depend on θ and Te, the ratioof the scattering signals at the same wavelength becomes a function of Te. Here, the efficiencies ofeach scattering in the double pass is assumed to be the same. Therefore, the ratio is independent
from the sensitivities. Once Te is decided, the relative sensitivities among different wavelength
channels can also be calculated, because the spectral waveforms are already fixed by the known Teand θ. To demonstrate the method, a Thomson scattering diagnostic on TST-2 [2] utilizing adouble-pass scattering system was used.
As the first step of the demonstration, Te measured from the ratio and Te from the standard method(i.e., wavelength spectrum) were compared. In TST-2, however, the effective scattering length (Ls)and laser energy (E) were different between the two scattering in the double pass. Thus, themeasured ratio depends not only on Te , but also on the decay of Ls×E. The decay of Ls×E at thesecond pass can be estimated by assuming that electron density does not change between the twopasses. With this assumption, Te (100 – 300 eV) measured from the ratio agrees with Te from thestandard method with the accuracies of about 2 – 15 %. When we have more than two wavelength
channels, we obtain more than two ratios. In this case, we can calculate two (fitting) parameters Te
and the decay of Ls×E. The agreement between Te s was not so clear due to the lack of the numberof spectral channels with high SN ratio. The error evaluations in relative sensitivities determinedfrom Te measured by the ratio will be also discussed.
References[1] H. Tojo et al., Plasma and Fusion Research 6 1302018 (2011).
[2] T. Yamaguchi et al., Plasma and Fusion Research 5 S2092 (2011).
This work was supported by Japan Society for the Promotion of Science (JSPS) Grant-in-Aid for Young
Scientists (B) No. 23760819.

210

P2-48Off-axis temperature anisotropy measurement by a double-pass Thomson scattering
diagnostic system on TST-2
J. Hiratsuka, A. Ejiri, H. Hasegawa1), Y. Nagashima1), Y. Takase, H. Tojo2), T. Yamaguchi,
T. Ambo, H. Furui, T. Hashimoto, H. Kakuda, K. Kato, T. Oosako, T. Sakamoto, R. Shino,
T. Shinya, M. Sonehara, T. Wakatsuki, O. Watanabe
The University of Tokyo, Kashiwa 277-8561, Japan
1)Kyushu University, 6-1 Kasuga-koen, Kasuga, Fukuoka 816-8580, Japan
2)Japan Atomic Energy Agency, 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193, Japan
Multi-pass Thomson scattering scheme is attractive from the viewpoints of low density plasmameasurements and electron temperature anisotropy measurements. Especially, direct measurement
scheme of temperature anisotropy has not been established, while temperature anisotropy has
important effects on plasma equilibrium. As a pilot experiment for the temperature anisotropy
measurement, a double-pass Thomson scattering system was constructed on the Tokyo Spherical
Tokamak 2 (TST-2) device. In this system, a laser pulse makes a round trip through the plasma.
Combining the double-pass configuration with a fast detection system, the backward scatteringand the forward scattering can be measured almost simultaneously. In the double-passconfiguration, electron temperature perpendicular and parallel to the toroidal field can bemeasured from the backward scattering and the forward scattering, respectively. Using thissystem, electron temperature and electron density have been measured around the center ofplasma, in which region no temperature anisotropy was detected within error bars.
A multi-point Thomson scattering system has been also prepared on TST-2. In this system,
Thomson scattering measurement can be performed at six points in the plasma simultaneously.
When the plasma is in a banana region, Te⊥ can be higher than Te// in the high field side, and Te//can be higher than Te⊥ in the low field side. Therefore, temperature anisotropy profile is expectedin this case. Test experiments have been performed to search for the off-axis temperature
anisotropy by the double-pass and the multi-point Thomson scattering system. Preliminary resultsof the measurements will be presented.

211

P2-49The method to measure of poloidal plasma rotation by Heavy Ion Beam Probe
L. Eliseev, A. Melnikov, S. Perfilov, C. Hidalgo1)
Institute of Tokamak Physics, NRC “Kurchatov Institute”, 123182, Moscow, Russia.
1) Association EURATOM-CIEMAT, 28040, Madrid, Spain.
The paper proposes a method to measure poloidal rotation velocity of toroidal plasma using Heavy
Ion Beam Probe (HIBP) with a multi-slit energy analyser. The method is based on calculation ofphase shift between broadband density turbulence measured simultaneously in two poloidally
separated sample volumes. Oscillatory component of HIBP beam current is used as a density
turbulence characteristic. HIBP is capable to provide the temporal evolution of the vpol in a fixedradial position and also the evolution of the vpol profile by periodic radial scan. Method wasverified in real plasma experiment in ECRH discharges on the TJ-II stellarator. Result shows thatin low density discharges (ne ≈ 0.3-0.5×1019m-3) absolute values of local vpol is about 6-4 km/s.
The plasma rotation velocity is oriented in the ion diamagnetic drift direction. When HIBP
operates for radial scans, it is conventionally measuring the plasma potential profile, and so
provide the radial electric field Er and velocity of ExB drift (vExB) at the same time as plasma
rotation. Experimental data shows that in low density ECRH plasma the rotation velocity
coincides with ExB velocity within achieved experimental accuracy. When the density is
increasing, both vExB and vpol tends to decrease and then change the sign at ne ≈ 0.6-0.7×1019m3.With this new proposed technique HIBP becomes the new effective tool to study plasma rotationand turbulence characteristics in toroidal plasmas. **The work was supported by RFBR grants 10-02-01383 and 11-02-00067.

212

P2-50Development of a multichannel far-infrared interferometer for the Experimental Advanced
Superconducting Tokamak
N. Shi, X. Gao, Y. X. Jie, S. B. Zhang, Z. Long, Z. X. Liu and X. C. Wei
Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126, Hefei 230031, Anhui,
China
A new vertical five-channel far-infrared (FIR) deuterium cyanide (DCN) laser interferometer has
been developed successfully and will be operational in the next Experimental Advanced
Superconducting Tokamak (EAST) campaign. The system is based on the three-channel FIR HCN
laser interferometer. The FIR beams with a wavelength of 195 μm, at which the high-power source
availability makes this wavelength ideal for EAST plasma, is produced by DCN laser. The laser isdesigned to have a full remote control capability using the Programmable Logic Controller (PLC).
It can improve the stabilization of laser power in long-term operation automatically. The technicaldetails of the new FIR system will be described.
The first high-confinement mode (H-mode) with type-III edge localized modes has been obtained
during last EAST campaign. The electron density limit and effects of triangularity on confinementof H-modes from the old FIR laser interferometer have been studied. The ‘H-mode’ limit is found
at 0.65 ne/nG (the ordering parameter to describe the confinement with respect to the ITER scaling,nG is the Greenwald density). It is also observed that the higher triangularity discharges have better
confinement for a given density, and meanwhile, the confinement degrades with density for agiven triangularity.
References[1]N.Shi,Plasma Science and Technology,13(2011)[2]G.S.Xu,Nuclear Fusion,51(2011)213

P2-51On Plasma Diagnostics Required for Heliotron-type DEMO Reactor
M. Isobe, T. Akiyama, T. Tokuzawa, T. Goto, J. Miyazawa, T. Tanaka, A. Sagara
and the FFHR design group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Conceptual design studies of heliotron-type DEMO reactor FFHR-d1 have been initiated lately in
National Institute for Fusion Science [1]. In existing fusion devices, many superior plasma
diagnostics are routinely operated for extensive physics studies. Note that necessary and/or
feasible diagnostics depend on the stage of experiment. Diagnostic instruments and/or methods
have been greatly developed as plasma parameter rises according to practicability and/or
requirement. Because existing critical physics issues will be solved until the start of DEMO
operation, measurements in DEMO are thought necessary only for burning control andperformance optimization. Reliable long-term operation of diagnostics in DEMO must be
challenging because instruments have to function in hostile thermal and radiation environments.
The implementation of diagnostics is also challenging since the diagnostic port and space that canbe utilized must be fairly limited since the reactor plasma has to be surrounded by a massive
blanket. This line of thinking will reach the minimum diagnostic set and the least implementation
to the DEMO machine. In this paper, required plasma parameters that must be diagnosed tosupport stable, steady-state operation of heliotron-type DEMO are going to be discussed. In
addition, realistic issues and possible solution of diagnostic set will be also discussed.
References[1] A. Sagara et al., ISFNT-10, 11-16 September 2011, Portrand, US. O38.

214

P2-52Conceptual design of electron density measurement system for DEMO-relevant helicalplasmasT. Akiyama, R. Yasuhara, T. Tokuzawa, M. Isobe, K. Kawahata, K. Nakayama1), S. Okajima1) andthe FFHR Design Group
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)Chubu University, Matsumoto-cho, Kasugai-shi, Aichi 487-8501, Japan
Electron density measurement is still indispensable to control fueling on a DEMO reactor. Since
operation of the DEMO reactor will be steady-state, the density measurement should be highly
reliable and accurate. On the other hand, interferometers used for the continuous density
measurement on present fusion devices have to suppress the measurement error due to mechanicalvibrations and sometimes suffer from "fringe jump errors". Two-color measurement, which uses
two different wavelengths, can compensate the phase component due to the mechanical vibrations
and a short-wavelength laser can reduce risk of the fringe jump errors. However, introduction ofdifferent schemes in principle is necessary to establish highly-reliable density measurement
enough for steady-state operation.
A dispersion interferometer [1] is less sensitive to the mechanical vibrations. A mixed beam of thefundamental and the second harmonic components is used as a probe beam. The second harmonics
is generated with a nonlinear crystal. After passing through a plasma, another second harmoniccomponent is generated from the fundamental again. The fundamental is cut with a filter and thenan interference signal between two second harmonic components is detected. While the phaseshifts due to mechanical vibrations are the same between the second harmonic components, thosedue to the plasma are different because of the dispersion. Hence the phase shifts due to thevibrations are canceled optically and those due to the plasma only remain. It also becomes freefrom fringe jump errors principally by selecting a short enough wavelength, whose phase shift due
to a plasma is smaller than 1 fringe. Hence it is suitable for the density measurement on steadystate fusion reactors. Dispersion interferometers with a 10.6 µm laser source [2,3] are operated atpresent. A wavelength around 1 µm is one of candidates for DEMO-relevant plasmas fromviewpoints of a variety of optical components, efficiency of the frequency doubling and the phaseshift due to plasmas. High power and stable continuous-wave lasers such as a YAG laser with awavelength of 1.064 µm are commercially available. Periodically poled crystals such as
PPMgSLT and PPKTP will generate enough power of the second harmonics enough for detection.
In case that the phase shift is difficult to be smaller than 1 fringe, compensation with a polarimeter
which measures the Faraday rotation will be effective. The appropriate wavelength will be severalµm to avoid coupling with the Cotton-Mouton effect which causes difficulty of conversion to theelectron density from the measured rotation angle. Although the polarimeter is less accurate than
the interferometer because of the small rotation angle and weight by the magnetic field, the fringe
shift of the dispersion interferometer can be corrected easily since the fringe shift is only 3-4
fringes at most. The conceptual design of the density measurement system will be presented,
taking FFHR-d1 as an example. Some challenges particular to DEMO reactors (e.g., degradationof first mirrors, availability of optical fibers) are also discussed.
References[1] V. P. Drachev et. al., Rev. Sci. Instrum. 64, 1010 (1993).
[2] H. Dreier et. al., Rev. Sci. Instrum. 82, 063509 (2011).
[3] T. Akiyama et. al., Plasma and Fusion Research 5, 047 (2010).

215

P2-53Developments of Pulsed Terahertz Wave Diagnostics for Fusion Plasma
T. Tokuzawa, Y. Kadoya1), M. Hangyo2), M. Tani 3), K. Tanaka, I. Yamada, and K. Kawahata
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
1)Department of Quantum Matter, Hiroshima University, 1-3-1 Kagamiyama, Higashihiroshima,739-8530, Japan
2)Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871,
Japan3)Research Center for Development of Far-Infrared Region, University of Fukui, Bunkyo 3-9-1,
Fukui 910-8507, Japan
In the future burning plasma experiment, the electron density will be quite high and up to the orderof 1022 m-3. In such high density plasmas, the conventional microwave diagnostic technique will
be still expected to measure the density profile, its fluctuation, etc. However, the demanding andutilizing frequency is getting into terahertz regime (0.1 THz-10 THz). Recently, the terahertz wave
technologies have been much interest in world wide. There are several techniques for own
applications. Among them, THz Time-domain Spectroscopy (THz-TDS) on which many of theresearches have focused has several features, because it has a capability of getting broadband
spectral information covering from sub 100 GHz to a few THz. At present, however, the highpower source and high resolution detector in terahertz regime have not been enough to apply
plasma measurements.
Currently, we heve been constructing a typical THz-TDS system. The pumping femto-second fiber
laser light with 780 nm is focused to the photoconductive antenna. THz optical pulse of which thewidth is less than 1 pico-second is generated by exciting a bow-tie type of photoconductive
antenna. THz beam is collected and focused on a detector which is activated only for a sub-picosecond period by the optical pulse branched from the excitation one. By scanning the timedifference between the excitation and detection pulses, the time-domain trace is recorded as insampling oscilloscopes. The amplitude and phase spectra were obtained by the Fouriertransformation of the time domain trace.
At the conference, we will present the system performance and discuss the future plan ofdevelopments.
This work was partially supported in part by KAKENHI (Nos. 22360394, and 22017007). Also it was in part bya budgetary Grant-in-Aid of NIFS LHD project and the Joint Research of ILE Osaka University.

216

P2-55
Assessment of multi pulse laser damage threshold of metallic mirrors for laser diagnostics infusion devices
M.Sato, S.Kajita1), N.Ohno2), R.Yasuhara3), M.Tokitani3), Y.Tawara1)
Nagoya University, Nagoya, Aichi 464-8603, Japan
1)ETSI Nagoya University, Nagoya, Aichi 464-8603, Japan
2)Nagoya University, Nagoya, Aichi 464-8603, Japan3)National Institute for Fusion Science, Toki, Gifu 509-5292, JapanIn ITER, laser transmission mirrors will be used for Thomson Scattering Diagnostics. For the lasertransmission mirrors, which transmit laser pulses to plasma, high reflectivity at 1064 nm, highmulti pulse LIDT(Laser induced damage threshold) and high durability for sputtering by plasma
are needed. Furthermore, for laser transmission mirrors it is required to maintain optical properties
even after 107 laser pulses irradiation. If the surface of mirror is damaged by laser irradiation,reflectivity of the mirror decreases seriously. However, most of the experimantal data about multipulse LIDT of copper mirror is about 102-103 pulses and few investigation has been done about
silver mirror. Therefore, in this study, the effects of the multi pulse LIDT on OFHC(Oxygen - freehigh thermal conductivity)-copper(KUGLER Co.) are experimentally investigated up to 7.3×104
pulses. The copper mirror was irradiated with pulsed Nd:YAG laser (1064 nm, 5 ns). Single shot
LIDT was 4.8 J/cm2 and multi pulse LIDT was 1.5 J/cm2 for 7.3×104 pulses. Experimental results
perdicted that the expected multi pulse LIDT for 107 pulse is ~0.5 J/cm2. This value was about
twice as high as the predicted value in ref [1].The surface roughness of the KUGLER mirror was
better than 3nm and the good surface property could have raised the LIDT.
References[1]S.Kajita et al., Plasma and Fusion Reserch Volume 3,032(2008)217

P2-56
Confinement Analysis of Spherical Tokamak-Stellarator Hybrid Configurations with Simple
Shaped Coils
H.Ozeki1), K.Yamazaki1), H.Arimoto1), T.Oishi1), T.Shoji1), M.I.Mikhailov2)
1)Nagoya University, Nagoya, Aichi 464-8603, Japan
2)Russian Research Center 'Kurchatov Institute', 123182 Moscow, Russia
One of the critical issues for the achievement of attractive fusion reactors is to reduce reactor
construction cost and to get good economic efficiency. To solve this problem, it is desired that
innovative magnetic configurations with compactness and high capability of confining fusion
plasmas are researched. In our laboratory a new concept of magnetic configuration called
TOKASTAR[1] was proposed. This configuration is a tokamak-stellarator hybrid with a few
spherical helical coils. In the previous study of TOKASTAR one type of coils shape was analyzed.
Analyses of different types of coils are needed to get deep knowledge of TOKASTAR
configurations and improved capability of plasma confinement.
In this study, we proposed 5 new types of TOKASTAR configurations in addition to previous N=2
TOKASTAR; N=1, N=4, N=2 with central helix, N=2 with crescent cross-sectional configuration,
and N=2 cylindrical coil system. Here N is the toroidal mode number of spherical helical coils. We
carried out comparative computational analysis among these configurations to improve
TOKASTAR’s capability of confinement.
We studied the magnetic flux surface, the single particle orbit, the equilibrium, and the Mercier
mode stability of TOKASTAR configurations. The HSD code was used for the analysis ofmagnetic flux surface and single particle orbit, and the fixed boundary VMEC code was utilized
for the analysis of equilibrium and Mercier mode stability.
In the magnetic flux surface analysis, the aspect ratio of magnetic flux surface, rotationaltransform and magnetic well were analyzed for 6 types of TOKASTAR configuration. These
configurations produce low values of aspect ratio from 1.1 to 3.1, low values of last-flux-surfaceaveraged rotational transform from 0.01 to 0.04, deep well depth from 30% to 50%. Especially,
N=2 center helical configuration produces relatively high value of rotational transform about 0.04
in comparison with other TOKASTAR and deep well of 50%. In this configuration Mercier modecan be stabilized due to the effect of well. In the analysis of equilibrium, equilibrium beta limit isanalyzed. The results indicated that high equilibrium beta is achieved in the configurations withhigh ellipticity, triangularity.
Separately from computational analysis, we are going to make experiments of TOKASTAR. We
also mention about present status of experiments in the conference.
References[1] K.Yamazaki and Y.Abe: Nagoya Research Report IPPJ-718 (1985)

218

P2-57
Experiments of tokamak discharge and simulation of magnetic field configuration in
tokamak-helical hybrid device TOKASTAR-2
M.Hasegawa, K.Yamazaki, H.Arimoto, T.Oishi, R.Nishimura, T.Shoji
Nagoya University, Nagoya, Aichi 464-8603, JapanTOKASTAR configuration [1] is one of compact tokamak-helical hybrid confinement systems.
The small device named “TOKASTAR-2” (the major radius is ~0.1m and the magnetic filed
strength is ~1kG) was designed and constructed to generate both tokamk and stellarator
configurations independently. One of main purposes of TOKASTAR-2 is to evaluate the effect ofhelical field application on tokamak plasmas.
We produced pre-ionized plasma using toroidal field (TF) coil system and 2.45GHz RF wave
injection, and induced plasma current through pre-ionized plasma using ohmic heating (OH) andvertical field (VF) coil systems. About 4V one-turn voltage was applied to the pre-ionized plasma.
However, we could not obtain more than 0.1kA plasma current within the limitation of DC power
supply for VF coil system and induced eddy current on the vacuum chamber. Photographs taken
by fast camera showed that the plasma moved vertically upward when one-turn voltage wasapplied. It was considered this movement is one of factors preventing the increase in plasma
current.
Then, we installed the conducting shell in TOKASTAR-2 and plasma current of shell-plasma was
about 40A larger than that of shell-less plasma. Although plasma current increased, it was unclearthat the conducting shell restrains vertical movement of plasma. Therefore, we plan to investigatethat effect using the fast camera. Additionally we have been making magnetic probes to measuremagnetic field configuration of tokamak plasma and its movement.
Other purpose of TOKASTAR-2 experiments is to generate current-less helical configuration. By
magnetic field line tracing analysis, it was revealed the vacuum magnetic surface can not be
formed because present TOKASTAR-2 has only outer helical field (HF) coil system. Installing
additional helical field (AHF) coil system is able to generate the vacuum surface. For example,
magnetic field tracing analysis clarified the formation of last closed surface in the case that coilcurrents of TF (ITF), VF(IVF), HF(IHF) and AHF (IAHF) are ITF=6122AT, IVF=17596AT,
IHF=6250AT and IAHF=7187.5AT respectively.
In the poster, the detailed experimental results of tokamak plasma and the simulation analysis ofhelical configurations will be presented.
References[1]K.Yamazaki and Y.Abe, Nagoya Reseach Report IPPJ-718 (1985)

219

P2-58
Improved Data Acquisition Methods for Uninterruptible Signal Monitoring and Ultra Fast
Plasma Diagnostics in LHD
H. Nakanishi, S. Imazu, M. Ohsuna, M. Kojima, M. Nonomura, M. Shoji, M. Emoto, M. Yoshida,
C. Iwata, H. Miyake, Y. Nagayama, K. Kawahata
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Since the beginning of LHD’s steady state experiments, long pulse plasma diagnostic data havebeen acquired according to a simple concept that partitions a long pulse into a consecutive seriesof 10-second "subshots" [1-2]. It can deal with not only 100 MB/s high-speed real-time (RT) dataacquisitions (DAQs), but also non-RT batch processing ones using conventional CAMAC
digitizers. The newest digitizers applying the high speed PCI-Express technology, however, canoutput GB/s nonstop data stream from each DAQ frontend. They would produce too huge data for
a "subshot" chunk if 10-second rule were just applied.
Uninterruptible monitoring for environments and device healthiness will be more required forsteady state fusion plants. Device healthiness signals and environmental radiological dosage had
been acquired by their embedded engineering computer systems, respectively, for the past ten
years in LHD. They had done their term of service and have been renewed as a part of the standard
DAQ framework of LHD, namely, LABCOM system.
In order to cope with both the uninterruptible monitoring and ultra-fast diagnostics, the subshot
length fixed on 10 seconds has been modified to be able to have variable length. The former
adopts dual subshot lengths of 10 minutes and 1 day. Longer subshots contain 1 Hz sparse datathinned from the original 1 kHz sampling ones. The latter uses shorter variable intervals less than
10 seconds, according to each use case. In this study, the design modification for variable subshot
lengths has been implemented and then verified its practical effectiveness at LHD.
References[1] H. Nakanishi, et al., Fusion Eng. Des, 56-57 (2001) pp.1011-1016[2] H. Nakanishi, et al., Fusion Eng. Des, 66-68 (2003) pp.827-832

220

P2-59Study of Performance Improvement of Real-Time Mapping of Thomson Scattering Data to
Flux Coordinates in LHD
M.Emoto, M.Yoshida, C.Suzuki, Y.Suzuki, K.Ida, Y.Nagayama, T.Akiyama, K.Kawahata,
K.Narihara, T.Tokuzawa, I.Yamada
National Institute for Fusion Science, Toki, Gifu, 509-5292, Japan
More than 100 diagnostic devices are attached to the vacuum vessel of the LHD (Large Helical
Device), and they are measuring the various aspects of the plasma physics. Because the shape of
the LHD plasma is not symmetric, each diagnostics obtains the physical values in different crosssection. For example, Thomson scattering measures the electron temperature profile in thehorizontally elongated cross section and the laser interferometer measures the line integrated
electron density profile in the vertically elongated cross section. In order to analyze the dataobtained by different diagnostics, their measurement positions must be mapped to the unified
coordinate system, or the flux coordinate system. Therefore, the authors have been building thedatabase to map the physical coordinate to the flux coordinate. Using the database, a mapping
system of the electron temperature profile to the flux coordinates, which is called TSMAP, hasbeen developed.
The profiles calculated by TSMAP are fundamental data to analyze the plasma physics during theexperiment. Therefore, they are required as soon as possible. However, the execution of TSMAP
needs the computational power, and the performance of a typical personal computer is not highenough to catch up with the 3 minutes’ plasma discharge cycle. In order to increase theperformance, the authors use multiple personal computers. Because the calculation of each plasma
discharge is independent from each other, they can be executed at the same time. Therefore, theauthors use multiple physical and virtual machines to execute the calculation of several plasma
discharges simultaneously. Using this approach, the calculation manages to catch up with theexperiment cycle.

221

P2-60THE INITIAL DESIGN OF PROCESS CONTROL SYSTEM BASED ON EPICS FOR HL2A & HL-2M
F．Xia, L. Pan, X. Song, X.Y. Li, C. Wang, L.Y. Chen
Southwestern Institute of Physics
A new TOKAMAK named ‘HL-2M’ will be built in next four years at SWIP. The objectives of
HL-2M is to elucidate important research issues such as the mechanism of plasma confinement
and transport, the physics of high energy particles, MHD stability, the interaction between firstwall and plasma and so on in physical and engineering related to ITER and future fusion reactor.
Many conceptual designs for the operation of the new device may be validated on HL-2A which iscurrent running at SWIP.
In addition to the real time plasma control system, the process control is also very important to theoperation of the Tokamak. The process control system will aim at the normal control of plantsystems, the information sharing, the schema of Tokamak operation, the alarm handling and theerror tracing. For those purposes, we select EPICS as the basic platform for the process control
system for its performance, reliability, expansibility and maintainability. Some user interface, data
management and development tools are adopted such as control system studio(CSS), MS SQL
Server, Matlab, LabVIEW, VC++,Java, etc. For the initial design, we will take several subsystems
such as the operator interface (OPI), the data acquisition system, the ECRH system, the fuelingsystem, the power system etc. into the platform and will handle about 2000 process variables (PV).
The paper will focus on the initial conceptual design for the process control system for HL-2A/HL-2M.
This work is supported by the National Nature Science Foundation of China under Grant No. 10835009 and by
the Chinese National Fusion Project for ITER under Grant No. 2009GB103005.

222

P2-61
Operation planning of tritium recovery system based on investigation results of LHD
exhaust system
Y.Asakura, N.Suzuki
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
In order to realize the planned deuterium plasma experiments using the Large Helical Device(LHD), the National Institute for Fusion Science (NIFS) is planning to install a system for tritium
recovery from exhaust gas.
In the LHD, following two types of tritium recovery systems are planned. One system will be used
to recovery tritium generated during the plasma experiments and contained in vacuum pumpinggas (processing capacity: 10Nm3/h) and the other to recovery tritium remained in the vacuum
vessel and contained in purge gas of the vessel during inspection and maintenance of the vessel(processing capacity: 300Nm3/h).
It is necessary to understand the treatment conditions at the connecting point between the LHD
vacuum pumping system and the tritium recovery system, such as, the hydrogen gas concentration,
the impurity concentration, and the gas flow rate to decide fabrication specifications and theoperation method of the two types of tritium recovery systems. Then, the measurement survey in apresent vacuum pumping system was carried out.
Both the hydrogen concentration and the gas flow rate were tried to measure simultaneously byusing gas sampling lines installed before and behind the exhaust blower. Two types of hydrogen
meters (gaseous heat conduction type and flammable gas combustion type) were set up in thesampling line at the blower exit part and the hydrogen concentration was measured. On the other
hand, the pure He gas is added to the sampling line at the blower entrance part in the constant flow
rate. The increment in concentration by the addition of the He gas was measured from thedifference of the measurement values of the flammable gas combustion type hydrogen meter (onlyhydrogen can be detected) and the gaseous thermal conducting hydrogen meter (both hydrogenand helium can be detected). The flowing rate of the exhaust was evaluated from the dilution rateof the pure He gas.
The reconstruction plan of the LHD exhaust system and the best operation plan of the two types oftritium recovery systems for the deuterium plasma experiments will be discussed based on theinvestigation result of the present exhaust system.

223

P2-62
Electromagnetic Fields Measurement and Safety Consideration in Magnetic Confinement
Fusion Test Facilities
T. Uda, M. Tanaka, J. Wang1), O. Fujiwara1)
1) Nagoya Insitute of Technology, Nagoya, Aichi 466-8555, Japan
Occupational electromagnetic fields were measured around the Large Helical Device (LHD),
which is a large magnetic confinement nuclear fusion test facility located in Japan. The LHD
equipped with a superconducting magnet coils system and high-power plasma heating systems like
Neutral Beam Injection (NBI), Electron Cyclotron resonance Heating (ECH) and Ion Cyclotron
Range of Frequencies (ICRF) heating. The leakage of the static magnetic field from LHD was
almost less than 0.1 mT, although it varied according to the superconductive coils system
operation conditions. The extremely low frequency electromagnetic field was measured around thepower supply boxes for the coil system, and higher magnetic strength which exceeds the guideline
levels was observed. Leakage of high frequency electromagnetic fields from the ICRF devices of
25-100 MHz were observed in bursts according to the experimental plasma shots. The measured
leakage was less than the occupational guide line levels. As results, it revealed that theelectromagnetic fields in the magnetic confinement fusion test facility should monitor in more
precisely and in long term from the viewpoint of the occupational safety and management forworkers. Also because the electromagnetic fields were generated in burst like and statisticallyvarying mode, we have developed and established the continuous measurement system.

224

P2-63
Analysis on Tritium Management in FLiBe Blanket for FFHR2
Yong Song, Akio Sagara1), Takeo Muroga1), Qunying Huang, Muyi Ni2), Yican Wu
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031, China
1)Reactor Engineering Research Center, National Institute of Fusion Sciences, Oroshi-cho, Toki
509-5292, Japan
2)School of Nuclear Science and Technology, University of Science and Technology of China,
Hefei, Anhui, 230027, China
In FFHR2 design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity withoxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care
and adequate mitigation measures should be applied for the effectively extract tritium frombreeder and control the tritium release to the environment. In this paper, a tritium analysis modelof the FLiBe blanket system was developed and the preliminary analysis on tritium permeationand extraction for FLiBe blanket system were done. The factors which affected tritium extraction
and permeation were calculated and evaluated, such as the heat exchanger material, tritium
permeation reduction factor (TPRF) in blanket, proportion of FLiBe flow in tritium recover system(TRS) and efficiency of TRS etc. The results of the analysis showed that further R&D efforts wererequired for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example
reasonable quality of tritium permeation barriers on blanket, requirement for the TRS andfabrication technology of the heat exchanger etc..
References[1]A. Sagara, S. Imagawa, O. Mitarai, et al., Improved structure and long-life blanket concepts for heliotron reactors, Nucl.
Fusion 45 (2005) 258-263.
[2]S. Fukada, A. Morisaki, A. Sagara, T. Terai, Control of tritium in FFHR-2 self-cooled Flibe blanket, Fusion Engineeringand Design 81 (2006) 477-483.
[3]G. Gervasini, F. Reiter. Tritium Inventory and Permeation in Separately cooled Liquid Breeder Blankets. J. Nucl. Mater.,
1989, 168(3): 304-311.
[4]W. Farabolini, A. Ciampichetti, F. Dabbene, M.A. Futterer, L. Giancarli, et al., Tritium control modeling for a helium
cooled lithium-lead blanket of a fusion power reactor, Fusion Eng. Des. 81 (2006) 753-762.
[5]S. Fukada, A design for recovery of tritium from Flibe loop in FFHR-2. Fusion Power Plants and Related Advanced
Technologies, Feb. 5, 2007.
[6]A. Sagara, O. Mitarai, T. Tanaka, S. Imagawa, Y. Kozaki, M. Kobayashi, et al., Optimization activities on design studies of
LHD-type reactor FFHR, Fusion Engineering and Design 83 (2008) 1690-1695.
[7]B. Merrill, S. Malang, Tritium Permeation and Extraction Issues for the ARIES-CS, ARIES Meeting, University of
Wisconsin, June 15th, 2005.
[8]Y. Song, Q. Huang, Y. Wang, M. Ni, Analysis on tritium controlling of the dual-cooled lithium lead blanket for fusion
power reactor FDS-II. Fusion Engineering and Design 84 (2009) 1779-1783.
[9]C.P.C. Wong, M. Abdou, J. Blanchard, P. Calderoni, D.P. Carosella, M. Dagher, et al., Design Description Document for
the U.S. Dual Coolant Pb-17Li (DCLL) Test Blanket Module, TBWG Meeting, Nov 15, 2005.
[10]E. Serra, G. Benamati, O. V. Ogorodnikova, Hydrogen isotopes transport parameters in fusion reactor materials, Journalof Nuclear Materials 255 (1998) 105-115.
[11]R. A. Causey, W. R. Eampler, The use of Silicon Carbide as a tritium permeation barrier, Journal of Nuclear Materials
220-222 (1995) 823-826.
[12]P.X. Wang, J.S. Song, Helium in Material and the Permeation of Tritium, National defence industrial publicationcompany of China (2002).
[13]N. Baluc, K. Abe, J.L. Boutard, et al., Status of R&D activities on materials for fusion power reactors, Nucl. Fusion 47(2007) S696-S717.
[14]H. Yoshida, O. Kveton, J. Koonce, D. Holland, R. Haange, Status of the ITER Tritium Plant design, Fusion Eng. Des. 3940 (1998)875-882.

This work was partly supported by the JSPS-CAS Core -University Program in the field of "Plasma and Nuclear
Fusion". We would further like to thank the great help from the members of FDS Team in ASIPP and USTC in
this research.

225

P2-64
Hydrogen Isotopes Recovery from Liquid Li with Y Hot Trap
S. Shigeharu, Y. Hatachi, T. Hiromoto, Y. Edao and S. Fukada
Department of Advanced Energy Engineering Science, Interdisciplinary Graduate School ofEngineering Sciences, Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581,
Japan
Materials to construct fusion reactor chamber are subjected to neutron irradiation. It is necessary toanalyze how it affects the material in International Fusion Materials Irradiated Facility (IFMIF).
The material is irradiated by the neutron beam which is generated by D-Li stripping reaction.
Tritium(T) generated by the reaction need to be recovered from liquid Li for safety. T recovery byan yttrium(Y) hot trap is one of the ways to recover T from liquid Li in IFMIF. Therefore, it isnecessary to analyze the behavior of T in the liquid Li. We’ve already traced the results under theconditions of static liquid Li and dynamic liquid Li. We concluded those hydrogen absorption
rates of Y from liquid Li depended on the condition of surface no matter it was static or dynamic.
In this study, we measured H2 absorption rates in stirred conditions of liquid Li and analyzed theeffect of hydrogen absorption rate of Y by elevating temperature from 250°C to 400°C to establish
the way to recover T from liquid Li with the Y hot trap in IFMIF.
In the experiment, we prepared a Y plate of 99.9% in purity, 0.25mm in thickness and therectangular size of 25mm × 25mm. Oxide films on the Y surface were removed by HF treatment
because the films prevent Y from absorbing H. The Y plate was put in a Mo crucible and Li wasput on the plate. The Li was stirred by a propeller under conditions of the rotation rates of 0100rpm. When a constant concentration of H2/Ar mixture was supplied into the Y-Li system, theH2 effluent concentration was determined by a gas chromatograph. H absorption rates in the Y-Li
system were compared with numerical calculations. We compared those H absorption rates in thetemperature range of 250°C -400°C, various rotation rates and activation conditions of Y. Judging
from the comparison, we considered the rate-determining step was H diffusion in Y. Mass transfer
coefficients were determined by fitting among the results of the calculation and the experiment.
References[1]Kenjiro Matsuhiro, Hirofumi Nakamura, Takumi Hayashi,Hiroo Nakamura, Masayoshi
Sugimoto, "EVALUATION OF TRITIUM PERMEATION FROMLITHIUM LOOP OF IFMIF TARGET SYSTEM", FUSION SCIENCE ANDTECHNOLOGY, 48 (2005) 625-628[2] H. Nakamura et al, "Latest design of liquid lithium target in IFMIF", Fusion
Engineering and Design, 83 (2008) 1007-1014

226

P2-65
Analysis of simultaneous H and D permeation through lithium-lead
M. Okada, H. Okitsu, Y. Edao and S. Fukada
Department of Advanced Energy Engineering Science,Interdisciplinary Graduate School ofEngineering Sciences, Kyushu University
Lithium-lead eutectic alloy (Li-Pb) is a promising liquid blanket of an advanced fusion reactor
because of the potential abilities for less radiation damage, high T breeding ratio, easy
maintenance, and possible use as a coolant. For a fusion reactor to be realized, it is necessary torecover T from a blanket loop highly efficiently and to realize low T leak from fusion reactor. For
establishing a T recovery system and safety confinement, mass-transfer data of permeability,
diffusivity and solubility of hydrogen isotopes in Li-Pb are important issues. We haveexperimentally determined the mass-transfer properties by an unsteady permeation method. When
recovering T from Li-Pb, isotopic exchange reaction is proposed to improve T recovery ratio,
which means that there are multi-component hydrogen isotopes in Li-Pb. We have presented mass-transfer data of a single component of H2 or D2 in Li-Pb. In order to study an influence of
permeation of multi-component hydrogen isotopes in Li-Pb, we have experimented on
simultaneous H and D permeation through Li-Pb in the two component (H+D) system. T data canbe predicted based on isotope effects between H and D. In our experiment Li-Pb was put on a α-Fe
plate in a permeation pot, which side wall was made of SUS304. The following several kinds of H
-D composition ratios were tested ; H2 : D2=100:0, 90:10, 80:20, 66:34, 44:56, 34:66 and 0:100 in
the temperature range from 673K to 973K. Permeability, diffusivity and solubility were obtainedby comparing with analytical equations. We have simulated two-dimensional permeation analysis
through Li-Pb. By this simulation, part of hydrogen leaks in the side direction was estimated. The
effect is quantitatively taken into consideration. As a result, it was found that H and D leaks with alow rate through side wall and permeate independently regardless of the H/D component ratio.
While permeability and diffusivity of H were around 1.4 times larger than that of D, solubility of
H was close to that of D in the experimental temperature. The ratio of the isotope effect isconsidered to be equal to the square root of mass ratio of H and D, so T permeability anddiffusivity can be predicted as around 1.7 times of H in the temperature range from 673K to 973K.
References[1]F.REITER,Fus.Eng.Des.,14(1991)207-211[2]Y.EDAO,H.NOGUCHI,S.FUKADA,Fus.Eng.Des.,85(2010)1225-1228[3]T.SHIRAISHI,M.NISHIKAWA,T.YAMAGICHI,J.Nucl.Mater.,273(1999)60-65[4]A.TAHARA and Y.HAYASHI,Trans.Jap.Inst.Met.,26(1985)869[5]W.A.SWANSIGER and R.BASTASZ,J.Nucl.Mater.,85&86(1979)335-339

227

P2-66
Optical emission and mass spectra observations during hydrogen combustion processes inatmospheric pressure microwave plasma
N.Ezumi, K.Akahane, Y.Tanaka1), M.Tanaka2), T.Uda2), K.Nishimura2)
Nagano National College of Technology, Nagano 381-8550, Japan
1)Kanazawa University, Kanazawa, Ishikawa 920-1192, Japan2)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Recovery of leakage tritium in a nuclear fusion reactor building is one of the important issues. So
far, the tritium recovery system removes tritium from a gas by cracking the tritium-containing
components on a heated precious metal catalyst. The tritium combines with oxygen in the air
stream to form tritiated water. Then the tritiated water contained in the air stream is removed by amolecular sieve bed. Although this system has suitable performance efficiency, there are someproblems for high-pressure drop, the utilization of a large amount of precious metals and heating
efficiency etc., when the processing throughput is quite huge. In order to resolve the issues, wehave proposed a hydrogen isotope oxidation process by atmospheric pressure plasma. This method
have the advantages as follows, low pressure drop, without noble metals such as platinum andpalladium, hydrogen and oxygen radicals are easily generated by high energy electron and ion
impact etc.
Experimental studies on hydrogen isotope oxidation by an atmospheric pressure plasma generated
by 2.45 GHz microwave discharge have been done. Small amount of hydrogen and oxygen weremixed in the operational Argon gas during the discharge. The constituents of combustionprocessed gas were observed by a quadruple mass spectrometer. To clarify the detail of the
combustion process, optical emission measurement has been done. Notice that hydrogen was usedas simulated gas of tritium in this experiment.
The degree of hydrogen oxidation, so called conversion rate, is increased as increasing the input
microwave power. Maximum hydrogen conversion rate was reached to 85.5% in our experimental
condition. During the discharge, optical emission spectra from OH radicals were observed. These
results indicate OH radical might play important role in the hydrogen combustion process.
This work was supported by NIFS Collaborative Research Program (NIFS09KFSS015 and NIFS11KESA005)
and by Research Foundation for the Electrotechnology of Chubu, Grant-in-Aid (R-21244).

228

P2-67
Nonlinear Gyrokinetic Simulation Study on ITG Turbulence and Zonal Flow in LHD
Discharge
M. Nunami1), T.-H. Watanabe1,2), H. Sugama1,2), K. Tanaka1)
1)2)

National Instutite for Fusion Science, Toki, Gifu 509-5292, Japan
Graduate University for Advanced Studies (SOKENDAI), Toki, Gifu 509-5292, JapanFor ion temperature gradient (ITG) turbulent transport and zonal flows in a Large Helical device(LHD) discharge, nonlinear gyrokinetic simulations are applied by means of the GKV-X code [1].
In recent LHD experiment with the high ion temperatures, spatial profiles of the density
fluctuation were measured precisely by two-dimensional phase contrast imaging [2]. The
measured fluctuations show the characteristics of the ITG turbulence, i.e., propagating fluctuation
along the ion diamagnetic direction in the plasma frame, and increase of the amplitude due to thegrowth of the ion temperature gradient. In order to investigate the turbulent transport driven by the
ITG modes, we perform gyrokinetic simulations for the equilibrium field configuration of the
LHD discharge. While the linear ITG modes are most unstable in radial and wavenumber regions
where the fluctuation of the experimental measurement peaks [3], nonlinear simulation results givethe turbulence transport levels which are comparable to the experiment results of the anomaloustransport, and reasonable agreements with the experiment are found in the wavenumber spectra of
the fluctuations. Furthermore, we obtain a clear relationship between the turbulence, zonal flow,
and the transport, including the case with optimized LHD configurations.
References[1] M. Nunami, T.-H. Watanabe, H. Sugama, Plasma Fusion Res. 5 (2010) 016.
[2] K. Tanaka, C. A. Michael, L. N. Vyacheslavov et al, Plasma Fusion Res. 5 (2010) S2053.
[3] M. Nunami, T.-H. Watanabe, H. Sugama, K. Tanaka, Plasma Fusion Res. 6 (2011)
1403001.

229

P2-68
Effects of ion-temperature-gradient driven turbulence on magnetic islands
A. Ishizawa, F. L. Waelbroeck1), R. Fitzpatrick1), W. Horton1), N. Nakajima
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
1)Institute for Fusion Studies, University of Texas, Austin, Texas 78712, USA
Effects of ion-temperature-gradient driven turbulence on magnetic islands are investigated bymeans of numerical simulations of a reduced set of two-fluid equations which include not only
electron diamagnetism but the ion diamagnetism in slab geometry. Simulations are carried out inthe island fixed frame, where the width and poloidal location of magnetic island do not change.
Uniform ExB flow is applied to the island, and the drive as well as the drag force acting on theisland are calculated as a function of the externally applied flow velocity. The turbulent
fluctuations enhance the momentum exchange across the sepratrix of island and thus enhance thedrag force acting on the island. The zonal flow produced by the turbulence makes the island
propagation velocity deviate strongly from the one without the turbulence.

230P2-69
A two-scale model for zonal flow and turbulence in non-axisymmetric systems
T.-H. Watanabe, H. Sugama, M. Nunami
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Interactions of zonal flows and turbulence and resultant turbulent transport reduction have long
been an important issue in theoretical, numerical, and experimental studies of the anomaloustransport in magnetically-confined plasmas. In numerical simulations, zonal flows are supposed tobe n=0 modes with low m mode numbers because of the long scale-length in poloidal and toroidal
directions, while the micro-turbulence, such as ITG mode, is regarded as high n modes with small
spatial scales. The toroidal scale separation has naturally been assumed in the flux tube or in thewedge torus models of toroidal plasma confinement. In non-axisymmetric systems such as LHD,
however, the conventional scale-separation is not trivial, as a n=0 mode perturbation may couple
with the non-axisymmetric component of the confinement field with the field-line-label
dependence. The toroidal coupling of zonal flows and the non-axisymmetry is closely associatedwith the zonal flow enhancement in case with the equilibrium-scale radial electric field (that is, amacro-scale poloidal ExB flow) [1-3]. Therefore, we consider a two-scale approach to this issue.
The new theoretical model is applied to simulation of the zonal flow response enhancement in the
LHD type configuration with the macro-scale poloidal ExB flow.
References[1] Sugama H and Watanabe T H 2009 Phys. Plasmas 16 056101[2] Sugama H, Watanabe T H and Nunami M 2010 Contrib. Plasma Phys. 50, 571-575[3] Watanabe T H, Sugama H and Nunami M 2010 “23rd IAEA Fusion Energy
Conference” (Daejon Korea Rep. of)231

P2-70
Neoclassical Tearing Mode Analysis in Tokamak Plasmas
D.Kurita, K.Yamazaki, H.Arimoto, T.Oishi, T.Shoji
Nagoya Univ.
For the achievement of high beta value in tokamak fusion reactors, it is important to control
magnetic islands produced by the neoclassical tearing mode (NTM) and to suppress incidental
plasma confinement degradation [1].
In this work, plasma parameter change due to NTM is analyzed and the NTM island control by theelectron cyclotron current drive (ECCD) was studied using 1.5-dimensional transport code
TOTAL. The time variation of magnetic island is described by the modified Rutherford equation.
The NTM control by ECCD is affected by the stabilization efficiency of the EC current. The
stabilization efficiency of the EC current localization changes by EC injection phase, position andwidth. However, how much the efficiency of the EC current change by these values is notclarified.
The EC control efficiency ηec decreases by the lag in the EC injection phase from O-point ofmagnetic island Δαc. In the case of narrow EC injection width case, a large value ηec can beachieved when Δα;c is nearly equal to zero. However, ηec decreases more suddenly by increasing
Δαc than that in the wide poloidal width case.
When EC current Iec is small, ηec for complete stabilization of NTM becomes a larger value. So
NTM cannot be completely stabilized when Δαc is larger.
The change of the EC control efficiency by the lag in the EC injection phase from O-point ofmagnetic island and the temporal evolution analysis of magnetic island width will also be reported
in the conference.
References[1] Y. Takahashi, K. Yamazaki et al., Journal of Physics: Conference Series 123 (2008)
012036232

P2-71
Nonlinear Simulation of Energetic Particle Modes in High-Beta Tokamak Plasma
A. Bierwage, N. Aiba, Y. Todo1), G. Matsunaga, K. Shinohara, M. Ishikawa, M. Yagi
Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japan
1) National Institute for Fusion Science, Toki, Gifu 509-5292, JapanThe global hybrid code MEGA [1,2] is used to study the dynamics of energetic particle modes(EPM) [3,4] in parameter regimes close to the onset of ideal magnetohydrodynamic (MHD)
ballooning instability. This work is motivated by observations of energetic-particle-driven modesin the wall-stabilized regime (so called EWM) in JT-60U [5,6], the physics of which have not been
fully understood yet. The present simulations exclude several important ingredients required for acomplete description of EWM, such as a free plasma boundary and kinetic thermal ion
compression, and focuses only on the effect of high bulk plasma beta, βbulk, on the stability andnonlinear evolution of EPM. In the linear regime, it is found that the EPM growth rate increases asβbulk increases, which may readily be explained in terms of a reduction in ideal MHD potential
energy δWf. Simulations are currently underway to analyze the effect of high βbulk on the nonlinear
saturation level of the mode and the energetic ion transport. New results will be reported as they
become available.
References[1] Y. Todo et al., Phys. Plasmas 12 (2005) 012503[2] Y. Todo, Phys. Plasmas 13 (2006) 082503[3] L. Chen et al., Phys Rev. Lett. 52 (1984) 1122
[4] L. Chen, Phys. Plasmas 1 (1994) 1519
[5] G. Matsunaga et al., Phys. Rev. Lett. 103 (2009) 045001[6] G. Matsunaga et al., Nucl. Fusion 50 (2010) 084003
This research is partly supported by Grant-in-Aid for Scientific Research (Start-Up) from Japanese Ministry of
Education, Culture, Sports, Science and Technology, Grant No. 22860081, and by collaboration programs
between JAEA and universities.

233

P2-72
Simulation Study of Energetic Particle Driven Geodesic Acoustic Mode in LHD Plasma
H.WANG1), Y.TODO1,2)
1) The Graduate University for Advanced Studies, Toki, Gifu 509-5292, Japan
2) National Institute for Fusion Science, Toki, Gifu 509-5292, JapanThe energetic particle driven geodesic acoustic modes (GAMs) in LHD plasma were simulatedusing a hybrid simulation code for magnetohydrodynamics (MHD) and energetic particles[1]. Theenergetic particle distribution employed in the simulation is anisotropic in velocity space. The
GAM frequencies in the simulation results are 38kHz for 5.0keV electron temperature and 20kHz
for 1.3keV electron temperature. The frequencies agree reasonably with both the experimental
observation[2] and the prediction from MHD theory. The poloidal velocity vθ with mode number
m=0 and perturbed pressure and density with mode number m=1 are observed. The GAM is
excited around the normalized minor radius 0.2, and it propagates in radial direction. Both inward
propagation and outward propagation are observed in the simulation. This is the first simulation of
the inward propagation of the energetic particle induced GAM. Furthermore, the mode intensity
burst is also simulated. In addition, the spatial profiles of the perturbed pressure and density rotate
in poloidal direction. After the saturation of the instability, redistribution of energetic particles
takes place in pitch-angle space both for the lower pitch-angle part and for the higher pitch-angel
part.
References[1]Y. Todo, Phys. Plasmas, 082503 (2006) 13[2]T. Ido et al, Nucl. Fusion, 073046 (2011) 51
This work was supported by NIFS Collaborative Research Program (NIFS11KNST018).

234

P2-73
Effect of Nuclear Plus Interference Scattering on Fast-Ion Slowing-Down Distribution
Functions in Thermonuclear Plasmas
H.Matsuura, D.Sawada, Y.Nakao
Kyushu University, 744 Motooka, Fukuoka 819-0395, Japan
It is well known that fast ions slow down via Coulomb and/or non-Coulombic, i.e. nuclear plus
interference (NI)[1], scattering and create knock-on tails in fuel-ion distribution functions. Effectof the knock-on tail formation on plasma burning characteristics [2] and its application to plasma
diagnostics [3] have been examined. The NI scattering accelerates the slowing-down process offast ions [4], and it reduces the fraction of energetic component to total distribution function. The
reduction of energetic-ion population in steady-state plasma may influence the fast-ion loss andplasma heating efficiency [5]. It is important to quantitatively evaluate the effect of the NI
scattering on the fast-ion slowing-down distribution.
In this paper, on the basis of the Boltzmann-Fokker-Planck (BFP) simulation [2], effect of the NI
scattering on fast-ion slowing-down distribution function is evaluated. As a fast ion, alpha-particle
produced by the T(d,n)4He reaction, proton by the 3He(d,p)4He reaction or deuterium beam
injected into plasma is considered. A noticeable reduction in the fast ion population in confinedplasma is shown. The effect of the reduction on fast-ion loss process is discussed.
References[1] J.J. Devaney and M.L.Stein, Nucl. Sci. Eng. 46 (1971) 323.
[2] H. Matsuura, et al., Phys. Plasmas 13 (2006) 062507.
[3] R.K.Fishier, et al., Nucl. Fusion 34 (1994) 1291.
[4] Y. Nakao, et al., Nucl. Fusion 28 (1988) 1029.
[5] H.Matsuura, et al., Plasma Fusion Res. 6 (2011) 2405086.

235

P2-75
Simulation study of NBI beam ion distributions and heat depositions in the timedevelopment plasma of LHD
H. Yamaguchi and S. Murakami
Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan
High ion temperature plasma is obtained after the rapid decay of plasma density due to the pellet
injection in the LHD plasma. The increase of the plasma density by the pellet injection enhances
the ionization of neutral beams, and following rapid density decay increases the heating power per
particle. In order to analyze the transport property of this time developing plasma we have to use
the beam deposition analysis code including the effect of the plasma time development. In thispaper we improve the GNET code[1] taking into account the time development of the plasma
density and temperature. We perform the simulation in the time development plasma of LHD andestimate the time development of the beam ion distribution and heat depositions. Also we compare
the results with that of the previous version of GNET.
References[1] S. Murakami, et al, Nucl. Fusion 46 (2006) S425

236

P2-76
Study of Plasma Blob Dynamics with Particle Simulation
Hiroki Hasegawa1), Seiji Ishiguro1, 2)
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan2) The Graduate University for Advanced Studies (Sokendai), Toki, Gifu 509-5292, Japan
Recently, it is observed that coherent structures along magnetic field lines in scrape-off layer(SOL) of magnetic confinement fusion devices propagate from the edge of core plasma to the firstwall [1]. These structures, blobs, are thought to transport a plasma into the far SOL acrossmagnetic field lines. Then many theoretical and numerical investigations of blob dynamics havebeen performed on the basis of two-dimensional reduced fluid models [1]. However, in this kindof fluid model, kinetic effects, such as sheath formation between plasma and divertor plate andvelocity difference between electrons and ions are treated under some assumptions andparameterization. In this work, for the purpose of investigating blob dynamics including such
kinetic (microscopic) effects, we have developed a three dimensional electrostatic particle
simulation code with particle absorbing boundaries. Results of preliminary simulations show thatblobs move to the first wall across the magnetic field lines. In the case of periodic boundary
condition in toroidal direction (that is, the effect of sheath is neglected), it is observed that
propagating blobs are collapsed by some instabilities. This feature is consistent with the
expectation with the fluid theory. Furthermore, the effect of sheath on blob propagation is
discussed.
References[1] S. I. Krasheninnikov et al., J. Plasma Phys., 74 (2008) 679 and references therein.
This work is performed with the support and under the auspices of the NIFS Collaboration Research programs(NIFS11KNSS021 and NIFS11KNXN224) and supported by a Grant-in-Aid for Scientific Research from Japan
Society for the Promotion of Science (KAKENHI 23740411).

237

P2-77
Analysis of ECRH Preionization for Plasma Startup in JT-60SA
K. Hada1), K. Nagasaki2), K. Masuda2), R. Kinjo1), S. Ide3), A. Isayama3)1)Graduate School of Energy Science, Kyoto University, Uji, Kyoto, Japan
2)Institute of Advanced Energy, Kyoto University, Uji, Kyoto, Japan
3)Japan Atomic Energy Agency, Naka, Ibaraki, Japan
Tokamaks using superconducting coils have been developed towards a steady state operation. In a
superconducting tokamak, the generated loop voltage for breakdown is generally lower than thatof normal conducting tokamaks due to limitation in the coil voltage. This may cause a difficulty ofplasma startup especially when wall conditions are not ideal. To solve this problem, preionization
using Electron Cyclotron Resonance Heating (ECRH) has been proposed. ECRH preionization has
been experimentally investigated at low loop voltage in normal conducting tokamaks such as JT60U and DIII-D. In the Korean Superconducting Tokamak Advanced Research (KSTAR)
tokamak, it was found that ECRH assist is helpful for a plasma production. Since the JT-60SA
superconducting tokamak (R0 = 2.97 m, a = 1.18 m, BT = 2.25 T), which is now under constructionin JAEA [1], has a limitation of the toroidal electric field up to 0.5 Vm-1, we need to clarify
conditions for robust plasma initiation. In the initial JT-60SA operation phase, application of thesecond harmonic 110 GHz ECRH is under consideration for reliable plasma startup. In this paper,
we study the plasma startup assisted by ECRH in JT-60SA using a zero-dimensional (0-D) model
developed by Lloyd [2] and Bae [3]. The main purposes are to examine the conditions for reliable
startup and to clarify the dominant physics process. Five temporal equations are solved for a
spatially uniform plasma, that is, the electron and neutral density conservation equation, theelectron and ion energy density equations, and the electric circuit equation. We evaluate the ECRH
power required to produce plasma and determine the effects of the neutral density and the carbon
and oxygen impurity ions. It is found that the ECRH power of about 200 kW is required to start up
the plasma under the assumptions that the injected ECRH power is fully absorbed and the initial
neutral density is n0 = 0.3×1019 m-3. In addition, as the initial neutral density and impurities
density increase, the more ECRH power is required. The carbon impurity effect is a little larger
than that of oxygen. When the carbon impurity is more than 0.5% of the electron density, therequired ECRH power is more than 400 kW.
References[1] S. Ishida et al., Proc. 23rd IAEA FEC (Daejeon, Korea, 2010) OV/P-4[2] B. Lloyd et al., Plasma Phys. Control. Fusion, 38 (1996) 1627
[3] Y.S. Bae et al., IEEE Trans. Plasma Sci., 31 (2003) 745

238

P2-78Nonlinear collision effect on energetic particle confinement in LHD plasmas
Y.Masaoka, S.Murakami
Department of Nuclear Engineering, Kyoto University, Kyoto, 606-8501, Japan
Confinement of energetic particles is one of the important issues in thefusion reactor research.
Energetic α-particles are produced by D-T fusion reactions and theenergy of α-particle is indispensable to sustain a high temperature fusion plasma.
Also the lost energetic α-particles might damage the first wall.
Therefore,
it is important to confine the energetic particles until the energy
slow-down to thermal energy.
Particularly,
in helical systems, energetic particle trajectory is complicated in athree dimensional magnetic configuration.
Thus the confinement of energetic particles is one of the criticalissues in designing helical reactor.
Additionally,
collisions between the energetic particles could enhance the pitch angle scattering[1]and deteriorate the confinement.
Thus the analysis including the both complicated orbit and nonlinear
collision effects are necessary to make clear the energetic particle
confinement in LHD plasmas.
In this paper,
we study the energetic particle confinement including the collisions
with various plasma species such as electrons, ions, andenergetic particle.
We have improved the GNET (Global NEeoclassical Transport) code[2] totake into account the nonlinear collision effect in LHD.
We analyze the real and velocity space distributions and the energy andparticle loss rate changing the background plasma parameters,and verify the effects of collision between energetic particles on theenergetic particle confinement.
References
K. Okano, Nucl. Fusion, 31 (1991) 1349.
S. Murakami, et al. , Fusion Sci. Technol. 46 (2004)

239

P2-79Plasma domains and development of operation scenarios in JT-60SA
S. Ide, N. Hayashi, M. Honda, H. Urano, T. Suzuki, M. Miyata, N. Aiba, J. Shiraishi, G. Kurita,
K. Shimizu, H. Kawashima, T. Fujita and the JT-60SA teamJapan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki, 311-0193 Japan
JT-60SA (Super Advanced) is a superconducting (all the toroidal and poloidal coils) tokamak
updated from JT-60U [1]. The objective of the project is to support researches on ITER anddevelop physics and engineering basis towards DEMO reactor. One of the most important
objectives is to establish a plasma operational scenario of high normalized beta (βN) and highbootstrap current fraction (fBS) towards realization of a steady-state tokamak DEMO reactor. For
this establishment integrated achievement and sustainment of high confinement, high βN stability,
fully non-inductive current drive and heat and particle control are required. In other words, highintegration of control in wide extent is indispensable. In JT-60SA, various kinds of control
actuators/schemes are prepared such as, both passive (with the stabilization conducting plate) andactive (by the saddle coils) control for the resistive wall mode (RWM), NB lines with a variety ofboth acceleration energy (up to 500keV) and injection angles for non-inductive current drive andtoroidal rotation control, and an ITER-like divertor with gas fueling, impurity seeding andpumping for heat and particle control. Towards ITER and DEMO relevant regime, operation at ahigher current is desired. JT-60SA is designed to operate up to 5.5 MA with with q95~3 to study
plasmas at the lower normalized collisionality and the lower normalized poloidal gyroradius
domain. Achievable performance in these plasma domains and required operational scenarios havebeen investigated using various codes. And an advanced plasma of high fBS of >60%, a high βN of~4.3 is expected to be sustained fully non-inductively at 2.3 MA, for example. For theachievement of these plasmas, operation scenario and control at the current ramp phase isimportant. These are evaluated also utilizing various codes. In this presentation, expected plasma
domain and assessment of plasma performance in JT-60SA with emphasis on development ofoperational scenarios, including the ramp-up phase, will be discussed.
References[1] S. Ishida et al., Fusion Eng. and Design, 85, 2070 (2010)240

P2-80Current profile control for high bootstrap current operation in ITER
J. Mano, K. Yamazaki, T. Oishi, H. Arimoto, T. Shoji
Nagoya University, Nagoya, Aichi 464-8603, JapanFor the achievement of steady-state reactor, non-inductive current-drive plasma operation should
be maintained in tokamak fusion reactor. Total non-inductive current is a summation of abootstrap current proportional to the pressure gradient of plasma and an externally driven current,such as neutral beam current-drive. But in order to establish a commercial reactor, it is necessary
to reduce the amount of external current drive and to cover the majority of the plasma current withbootstrap current. Burning plasma has high autonomy, so we need to consider the current density
profile with little disturbance including changes in particle and heat transports.
In this study we conducted an analysis of time evolution of the current density profile for burning
plasmas in the ITER machine by using 2.0-dimensional equilibrium, 1.5-dimensional-transport
code (TOTAL code [1]). Here we used current-diffusive ballooning mode model as a heattransport model.
First we performed iteration considering the difference between trial value and target value of thetotal plasma current, such that the current profile becomes consistent with the given profile of thesafety factor. After this plasma equilibrium iteration analysis, we evaluate the bootstrap current
ratio and the total current density profile during steady state which is obtained from profiles of theexternal driving current and the bootstrap current given in the end.
We performed the simulation driving a current at only the center of plasma and both the center andthe normalized minor radius r/a=0.5 of plasma.
In the case of driving a current only at the plasma center, the ratio of the bootstrap current I bs andthe total plasma current Ip, Ibs/Ip is 0.981. However, electron density cannot be maintained andnuclear fusion reactions become lower. In the case of driving a current at both the center and thehalf normalized minor radius, the ratio Ibs/Ip is 0.901 and the profiles of temperature and electron
density can be maintained throughout more than 200 second.
From this we concluded that to maintain the steady-state temperature and electron density profile
of ITER in high bootstrap current ratio in the present model, external current-drive is required atboth the center and periphery of the plasma.
References[1] K. Yamazaki and T. Amano, Nucl. Fusion 32, 633 (1992)

241

P2-81
MHD simulation on pellet injection in torus plasmas
R.Ishizaki, N.Nakajima
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Injecting small pellets of frozen hydrogen into torus plasmas is a proven method of fueling [1].
Since a plasmoid induced by pellet ablation drifts to the lower field side, pellet fueling to make theplasmoid approach the core plasma is successful when a pellet is injected from the higher fieldside in a tokamak. On the other hand, such good performance has not been obtained yet inexperiments in the planar axis heliotron; the Large Helical Device (LHD), even when a pellet was
injected from the higher field side [2]. The purpose of this study is to explain the difference
between the motion of a plasmoid in tokamak and helical plasmas, and make a suggestion to
obtain the good performance on fueling in the LHD. To investigate plasmoid motion, a threedimensional MHD code has been developed by extending the pellet ablation code (CAP) [3]. In
the LHD, the drift motion depends on the initial location of the plasmoid, whereas in a tokamak,
the plasmoid always drifts in the direction opposite to that of the magnetic curvature vector. The
plasmoid motion is mainly determined by 1/R force due to toroidal field and the force due to a
dipole field induced by the plasmoid. The former force implies a drift in the direction opposite to
the magnetic curvature vector, which is dominant in a tokamak. The latter force implies theoscillation due to the field tension, which is dominant in the case that the plasmoid is locatedinside the torus in the LHD. Therefore, the plasmoid motions in the LHD are differ from those in atokamak. It is also verified that the connection length determines the force that dominates theplasmoid motion. In addition, the behavior of the plasmoid induced by ablation from a moving
pellet has been investigated. The pellet velocity dependence on the plasmoid behavior will be
discussed.
References[1] Y. W. Muller et al. Nucl. Fusion, 42 (2002) 301.
[2] R. Sakamoto et al. in proceedings of 29th EPS conference on Plasma Phys. and Control.
Fusion (2002).
[3] R Ishizaki and N Nakajima, Plasma Phys. Control. Fusion 53 (2011) 054009.

242

P2-82Modeling of SMBI experiments based on Monte-Carlo simulation in GAMMA 10
K. Hosoi, Y. Nakashima, S. Kobayshi1), N. Nishino2), T. Mizuuchi1), T. Ishii,
K. Ichimura, H. Takeda, and T. Imai
Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577, Japan
1)Institute of Advanced Energy, Kyoto University, Gokasyo, Uji 611-0011, Japan2)Graduate school of Engineering, Hiroshima University, Hiroshima 739-8527, Japan
Gas fueling control is one of the most important issues to obtain good performance plasmas.
Fueling control enables the profile control of the core plasma density and reduction of neutral
particles in the peripheral area. Supersonic molecular beam injection (SMBI) technique, which has
been developed by L.Yao et al.[1, 2], is a new method of gas fueling. SMBI can inject neutral
particle deeper into the core plasma compared to the gas puffing.
SMBI system is installed in GAMMA 10.The first results of SMBI showed that SMBI achieved
higher density plasmas at the core region than the conventional gas puffing case. A fast camera has
been installed at the central-cell in order to observe plasma behavior. The camera system has twolines of sight in the horizontal and vertical direction of the cross-section by using dual branch
optical fiber bundles. It is observed that the penetration depth of SMBI determined from theemission brightness was longer than that of gas puffing.
In order to interpret above observation results, DEGAS three-dimensional Monte-Carlo code
[3, 4] for neutral transport simulation has been applied to GAMMA 10 [5]. In a whole area of thecentral-cell vacuum vessel, a detailed 3-D mesh structure for the simulation was constructed.
Transport of neutral particles from SMBI can be simulated by using the DEGAS code. In thispaper, we will describe the detail of the simulation geometry for SMBI and the simulation results
carried out under conditions of various energies of neutral particles and of various positions of
SMBI. Effective operation method of SMBI is also discussed based on the 3-D simulation.
References[1] L. Yao, in “New Developments in Nuclear Fusion Research” (Nova Sci. Pub, pp. 61-87
2006).
[2] L. Yao, et al., Nucl. Fusion 47, 1399 (2007).
[3] D. Heifetz, D. Post, M. Petravic et al., J. Comput. Phys. 46, 309 (1982).
[4] Y. Nakashima, et al., J. Plasma Fusion Res. SEREIS 6, 546 (2004)[5] Y. Nakashima, et al., Contrib. Plasma Phys. 48, No. 1-3, 141-146 (2008)243P2-83Self-consistent Modeling of Plasma Production with Radio-frequency Heating
V.E. Moiseenko, Yu.S. Stadnik, A.I. Lyssoivan1)
Institute of Plasma Physics, National Science Center “Kharkiv Institute of Physics andTechnology”, 61108 Kharkiv, Ukraine
1)Laboratory for Plasma Physics - ERM/KMS, Association EURATOM - BELGIAN STATE,
1000 Brussels, Belgium
In stellarator type machines, besides the electron-cyclotron method, the plasma production in theion-cyclotron range of frequencies is practiced. The self-consistent model of the radio-frequency(RF) plasma production in stellarators is described in this work. With this model of plasma
production, one can perform calculations for different antenna systems. The self-consistent model
includes the system of the particle and energy balance equations and the boundary problem for the
Maxwell’s equations. The balance of the electron energy includes the RF heating, the energy
losses for the excitation and the ionization of atoms by the electron impact, energy exchange withions vie Coulomb collisions and the losses caused by the heat transport. The balance of thecharged particles includes the particle supply owing to ionization and the diffusion particle losses.
In the model, it is assumed that the neutral gas is uniformly distributed throughout the vacuum
chamber volume, including the plasma column. Besides plasma build-up inside the confinement
volume, the RF field produces plasma outside it. The losses of the charged particles in this zone
have a direct character because the particles of plasma escape to the wall along lines of force ofthe magnetic field. This process is accounted in the model in tau-approximation. The RF power
density is calculated from the solution of the boundary problem for the Maxwell’s equations. The
collisional and Landau wave damping are accounted as mechanisms of the RF field dumping. The
Maxwell’s equations are solved each time moment for the current plasma density and temperature
distributions. The calculated value of the local RF power, deposited to the electron component ofplasma, is used in the energy balance equation. This value influences on the electron temperature
and, in this way, on the ionization rate which determines the evolution of plasma density. The
model for the stellarator plasma column is the plasma cylinder with identical ends. The plasma isassumed to be azimuthally symmetrical and uniformly distributed along plasma column. The
Crank-Nicholson method is used for solving the system of the balance equations. The Maxwell’sequations are solved in 1D using the Fourier series in the azimuthal and the longitudinalcoordinates. Using this self-consistent model, the plasma density ramp-up with four-strap π-phased
antenna is modelled. Antenna is fed with the frequency below ion-cyclotron and, plasma
production in the Alfven resonance heating regime is realized.

244

P2-84
Wavelet-based analysis of Lower Hybrid full-wave fields
O. Meneghini, S. Shiraiwa, R. Parker
Massachusetts Institute of Technology - Plasma Science and Fusion Center, 77 Massachusetts
Avenue, Cambridge 02138, Massachusetts, USA
In this paper, we introduce the use of Continuous-Wavelet-Transform (CWT) to postprocess fullwave fields [1] of Lower Hybrid waves which have been generated using the LHEAF code.
Compared to the Fourier transform, the CWT has the appealing property of yielding information
as to the spatial location of spectral modes. Using a complex-Morlet CWT, the complicated fullwave field pattern is decomposed into its spectral components parallel to the static magnetic field(n||). In general, the CWT of the LHEAF full-wave fields shows that the local wave spectrum
broadens as the waves propagate through the plasma and after reflection off the low density cutoff
or the vacuum vessel walls. According to beam tracing theory, this result is to be attributed todiffraction effects taking place close to caustic surfaces [2]. The eventual goal of this analysis is toprovide a tool that can be used to assess the range of validity of ray-tracing calculations. Also, theimportance of full-wave effects on the transformation of the wave n|| spectrum and on the LH
power absorption and driven currents can be studied. Concerning this subject, particular attention
is given to LHEAF simulations of Alcator C-Mod LHCD discharges at high densities, in a regime
where an anomalous loss of LHCD has been reported [3]. LHEAF simulations using a realistic 2D
SOL model were found to reproduce qualitatively well the experimental drop of HXR emissionthat is observed for increasing plasma density. This result has been attributed to the combined
effect of an n|| upshift and radial diffusion of fast electrons [4]. The full-wave origin of such
spectral broadening needs to be confirmed by the CWT analysis hereby introduced.
References[1] D. D'Ippolito, et al., AIP Conference Proceedings, 694 463-466 (2003)[2] G. Pereverzev, Nuclear Fusion, 32 1091 (1992)[3] G. Wallace et al., Nuclear Fusion, 51 083032 (2011)[4] O. Meneghini, et al., 38th EPS Conference on Plasma Physics (2011)
Work supported by US DoE awards DE-FC02-99ER54512 and DE-AC02-76CH03073

245

P2-85Reconstruction of a toroidal flow profile of a field-reversed configuration
Y.Fujikawa, T.Asai, M.Gouda, Ts.Takahashi1), L.C.Steinhauer2), To.Takahashi3)1)College of Science and Technology, Nihon University, Tokyo, Japan2)Department of Aeronautics & Astronautics, University of Washington, Seattle, WA, USA
3)Graduate school of Engineering, Gunma University, Kiryu, Gunma, Japan
Strong toroidal flow is generated in a field-reversed configuration (FRC) plasma. It is
continuously accelerated during and after the formation phase. The centrifugal force resulting fromthe toroidal flow causes deformation of the toroidal cross section with toroidal mode number n =2. The deformation eventually brings a contact between the plasma surface and the chamber walland then the configuration is terminated.
The radial profile of this self-generated toroidal flow has been recognized as rigid rotor (RR) i.e.
uniform angle velocity profile. However, the toroidal flow profile measured in the field-reversal
theta pinch FRC on NUCTE (Nihon University Compact Torus Experiment) -Ⅲ facility has notagreed with the RR profile. The toroidal flow profile measured by ion Doppler spectroscopy (IDS)
technique has a peak at the vicinity of the separatrix and falls gradually in the scrape-off region.
The employed IDS is basically line-integrated measurement. Therefore, in this work, measured
distribution of line emission on wavelength and radial position has been reconstructed accurately
by applying Abel transform. The reconstructed profile has been discussed comparing with two
analytical equilibrium profile model i.e. widespread RR profile and newly presented two point
equilibrium (2PE) model [2].
References[1] R.L.Morse and J.P.Freidberg et al., Phys. fluids Vol. 13, 0531 (1969).
[2] L.C.Steinhauer and T.P. Intrator, Phys. Plasmas Vol. 16, 072501 (2009).

246

P2-86
Dependence of the quasi linear heating term model on the ECCD in helical plasmas
Y.Moriya, S.Murakami, K.Nagasaki1)
Department of Nuclear Engineering, Kyoto University, Kyoto, Kyoto 606-8501, Japan
1)Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011, Japan
Electron cyclotron current drive (ECCD) is one of the reliable methods of driving a plasma
current. ECCD can control current profile locally and has been applied for toroidal devices to keep
the current profile, to stabilize MHD instabilities and to cancel the bootstrap current in helicalsystems. In order to study the ECCD physics related to trapped electrons, ECCD experiment has
been performed in Heliotron-J and the ECCD dependence on magnetic configurations has been
investigated[1]. It was found that EC current depended on the magnetic configuration and thecurrent direction was reversed in high bumpiness configuration compared with the other twomagnetic configurations.
In this paper, we study the ECCD on helical plasmas by GNET code[2] in order to make clear therole of trapped electrons in ECCD. GNET can evaluate a steady state solution of distributionfunction in 5D phase-space using the Monte Carlo method. GNET has been developed for the
study of high energy electron transport in helical systems and is applied to ECCD analysis.
We analyze ECCD assuming magnetic configurations, plasma parameters and heating conditions
similar to the experimental ones. Here, electron cyclotron heating is taken into account throughquasi linear heating term. In the previous study[3], we approximated the heating term as a point
heating term using the delta function in the velocity space for simplicity. Simulation results withthis heating term showed that the current direction reversed between high and low bumpiness
configurations and these directions are determined by the balance between the Fisch-Boozer effectand the Ohkawa effect. These tendencies agrees well with the experimental ones. However, theobtained values of ECCD current is several times larger than that of the experimental ones. We
modify the heating term to the realistic one, which has the distribution in the velocity space.
In GNET simulation, we follows the test particle orbits to solve the drift kinetic equation using thelinear Monte Carlo collision operator (Boozer and Kuo-Petravic model[4]). This operator does notconserve the momentum and energy between test particles and field particles. We develop the
momentum conserved collision operator for GNET in order to simulate the current drive. We
present the development process of this operator and simulation results.
References[1] G. Motojima et al., Nucl. Fusion 47 (2007) 1045.
[2] S. Murakami et al., Nucl. Fusion 40 (2000) 693.
[3] Y. Moriya et al., Plasma Fusion Res. 6 in press.
[4] A. Boozer and G. Kuo-Petravic, Phys. Fluids 24 (1981) 851.

247

P2-88
Effect of Radiation Power Loss due to Ne Impurity Gas Puff for Ergodic Layer
I. Murakami, G. Kawamura, Y. Tomita, S. Masuzaki
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Reducing high heat load on a divertor plate is an issue to prevent serious damage to divertor plates
in future fusion devices such as the ITER, a DEMO, and a helical-type reactor (FFHR) [1]. Gas
injection to a divertor plasma is one of possible ideas to reduce heat load on divertor plates, since
impurity gas causes radiative power loss and decreases electron temperature, expected to result inplasma detachment. The effect of impurity radiation loss was examined analytically for the ITER
divertor [2] and neon was found to be the optimum candidate. A helical-type device has different
characteristics of edge magnetic field line structure from a tokamak. Plasma in the Large Helical
Device (LHD) at National Institute for Fusion Science has ergodic layer surrounding a coreplasma and the field connection lengths are distributed from orders of 1 m up to 105 m [3].
Charged particle flux onto divertor plates in LHD is not uniformly distributed and the heat load
becomes very high at a small region of the divertor plates. High particle flux points correspond topoints with longer connection lengths. Some plasma experiments with Ne gas puff into the edge
region in the LHD were performed during the 14th experimental campaign in 2010-2011. It was
found that the Ne gas puff reduced electron temperature in the outer region and the particle fluxes
onto the divertor plate were decreased much more at higher flux points. These results imply thatthe radiation power loss due to Ne gas plays an important roll at the ergodic layer. In order toexamine the effect of Ne gas radiation loss for the ergodic layer, we study a simplified 1
dimensional fluid model for magnetic flux tube with different lengths. In a one-zone model wefound the Ne gas radiation loss can reduce electron temperature less than 1 eV and we expect
plasma detachment [4]. In this work we examine how the Ne radiation power loss affects theplasma with spatial distribution along a flux tube and condition for plasma detachment.
References[1] A. Sagara et al., Fusion Engineering and Design 83 (2008) 1690.
[2] D. E. Post, J. Nucl. Materials 220-222 (1995) 143.
[3] M. Kobayashi et al., Contrib. Plasma Phys. 48 (2008) 255.
[4] I. Murakami et al.,Plasma Fusion Res. 6 (2011) 2403029.

248

P2-90A Pilot Plant as the Next Step toward an MFE DemoG. H. Neilson
Princeton Plasma Physics Laboratory, Princeton, NJ 08534 U.S.A.
With ITER now well underway, the next steps toward commercial magnetic fusion energy (MFE)
are under consideration worldwide. Substantial science and technology development is required
beyond ITER, for example in energy and tritium extraction, rapid replacement of internal
components, and steady-state plasma control with minimal recirculating power. An intermediate
integration device between ITER and the first prototype commercial power plant, or Demo, may
be required to reduce the risks of developing a reliable fusion system with increased availability.
This motivated an examination of possible pilot plants: facilities that would substantially narrow
the technical gap to a prototype commercial system in a next step. The Pilot Plant mission is to: 1)
test internal components and tritium breeding in a steady-state fusion environment, 2) prototype amaintainable design and maintenance scheme for a power plant, and 3) generate net electricity.
Such a facility would integrate the science and technology of a fusion power plant, demonstrateoverall system efficiency, and convincingly demonstrate fusion’s potential as an energy source.
Preconceptual designs based on the advanced tokamak (AT), spherical tokamak (ST), and compact
stellarator (CS) have been developed in order to compare their relative merits as fusion systems.
[1] Any of them would take a large step toward Demo in key performance metrics, e.g.
engineering gain QENG (≥1), neutron wall load (> 1 MW/m2), tritium breeding ratio (> 1), pulse
length (106 - 107 s), blanket lifetime fluence (≥ 3 MW yr./m2), plant lifetime (6 20 MW-yr./m2),
and availability (10-30%). Pilot plant facility designs differ from traditional physics researchdevices in important ways. Magnet systems must have large openings to accommodate rapid
replacement of large blanket and shield modules for high availability. First wall openings fordiagnostics and heating systems needed for plasma control must be minimized since these competewith blanket coverage needed for tritium breeding. Plant systems, especially those for plasma
heating and thermal conversion, must achieve a high level of energy efficiency in order to produce
net electricity.
A description of pilot plant design characteristics and an assessment of pilot plants against Demoprerequisites will be presented.
References[1] J.E. Menard, et al., “Prospects for pilot plants based on the tokamak, spherical tokamak
and stellarator,” Nucl. Fusion 51 (2011) 103014.
Research supported by the U.S. DOE under Contract No. DE-AC02-09CH11466 with Princeton University.

249

P2-91Comparative study of cost models for tokamak DEMO fusion reactors
T. Oishi, K. Yamazaki, H. Arimoto, K. Ban, T. Kondo, K. Tobita1), T. Goto2)
Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603, Japan
1)Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan2)National Institute for Fusion Science, Toki, Gifu 509-5292, JapanTo prove the feasibility of the nuclear fusion power plant, various designs of tokamak-typed
demonstration reactor DEMO have been proposed [1, 2]. One of the tasks of DEMO includes
establishment of the economical perspective after realization of fusion power plants. Therefore, anattractive design of DEMO requires the assessment of the economical evaluation. Especially,
appropriate choice of the cost model in the studies of the system codes for the optimization of thereactor design is necessary to show the economical validity of the DEMO or commercial reactors.
In the present study, we investigate the economical evaluation methods for previously proposed
DEMO reactor designs. The concept of both design and cost evaluation will be compared among
each design. The amount of materials used in the fusion island is employed as fundamental data of
cost evaluation. The cost of the balance of plant will be considered corresponding to the thermal
output power. We construct the database of the amount of materials using account number
classification for each design concept.
In the international comparison of cost models, it is important to consider the temporal variation ofprice of commodities or rate of exchange from the time when each cost evaluation was performed.
The methodology for the development of more generalized international cost comparison should
be discussed based on the past examples of cost comparison.
As a future study, development of a scaling law for the cost calculation is planned to clarify theeffect of each design parameter on the cost and to calculate the cost conveniently using
representative design parameters. It will be useful for the parameter survey study to optimize thecost using the system code for the reactor design [3].
References[1] M. Kikuchi, Nuclear Fusion, 30 (1990) 265[2] K. Tobita et al., Nuclear Fusion, 47 (2007) 892[3] K. Yamazaki et al., Nuclear Fusion, 49 (2009) 055017
This work was supported in part by the collaborative research program associated with the design activities forthe demonstration reactor in the Broader Approach.

250

P2-92
Economic evaluation of D-T, D-3He, and catalyzed D-D fusion reactors
T.Kondo, K.Yamazaki, T.Oishi, H.Arimoto, T.Shoji
Nagoya University, Nagoya, Aichi 464-8603, Japan
D-T fusion reactor system can be realized at low plasma temperatures in the first; however it
generates 14MeV neutrons that cause engineering problems and it needs tritium breeding system.
Since D-3He reaction generates no neutron and D-D reaction has abundant fuel resource, thesereactions are expected as advanced fuel fusion reactions. Not only the engineering problems but
also the economics are important in order to achieve the advanced fuel fusion reactor as acommercial plant. Therefore, we estimate and compare the cost of electricity (COE) of D-T, D3
He, and catalyzed D-D fusion reactors using the Physics-Engineering-Cost (PEC) system code
[1].
Since the reaction rates of D-3He reaction and D-D reaction are small, D-3He and D-D reactor
need to have high efficient confinement properties and be operated at high temperature.
Furthermore, power density of D-3He and D-D reactor are smaller than that of D-T reactor, so that
D-3He and D-D reactors require large plasma volume. Consequently, COE increases owing tolarge size of these reactors.
Because of neutron damage to the first walls, D-T reactor requires several replacements of blanket
within plant lifetime. In contrast, D-3He reactor has low neutron wall lord, and no need to replace
blanket. Thereby, D-3He reactor has no replacement cost. In addition, D-3He reactor can omit
tritium breeding system, except 3He gas should be explored in the moon.
Assuming high ion temperature (=70keV) and high normalized beta value (=8), COE of D-3He
reactor is expected to be similar to that of D-T reactor. According to analysis, pure ignited D-Dreactor has no feasibility as commercial plant. In terms of cost, catalyzed D-D might bedisadvantage compared with D-3He and D-T reactor.
References[1] K.Yamazaki, T. J. Dolan, Fusion Engineering and Design 81(2006)1145-1149

251

P2-93
Economical and Life-Cycle Energy Assessment of Magnetic Fusion Power Reactors
K.Ban, K.Yamazaki, H.Arimoto, T.Oishi, T.Shoji
Nagoya University, Nagoya, Aichi 464-8603, Japan
Fusion reactors might require enormous amount of construction costs and rare and valuable
materials. To search for an optimized power plant, we analyzed several types of fusion reactors(tokamak TR, spherical tokamak ST, and helical HR reactors) using physics, engineering and cost(PEC) code [1], which evaluates economical and life-cycle energy amount. We compared the costof electricity (COE) and the energy payback ratio (EPR) of each fusion reactor to the fission
reactor [2]. In addition, we compared several blanket and shield designs used rare and valuable
materials; such as silicon carbide (SiC), vanadium alloy (V), and ferritic steel (FS).
Reference fusion reactors in this study using PEC code have 1GW electric output, 75% availability
factor, and 30 year operating period, which corresponds to the reference of the fission reactor. The
EPR is defined as a ratio of output /input total system energy. As for electric power generation
plants, input energy is defined by the total energy relevant to fuel production, fusion power plant
construction, management and operation (M&O), replacement and decommission of reactor
equipment. Output energy is net total electricity from the fusion reactor during the plant life
period.
In the present study we found that the COE of TR is the lowest than those of other fusion reactors,
since the capital cost of the tokamak reactor TR is lower than those of HR and ST. However, theCOE of TR is higher than that of the fission reactor. The EPR of TR with the SiC blanket ishighest among several fusion reactor designs because of high thermal efficiency, and higher thanthat of the fission reactor.
In the conference, we will also present the assessment results of the inertial confinement fusion
reactor IR.
References[1] K. Yamazaki, T. J. Dolan; Fusion Engineering and Design, 81 (2006) 1145-1149.
[2] O. Amano et al.; Nuclear Technology Research Laboratory Rep.No.L09002 (2009) (inJapanese)252

P2-94
Design Window Analyses for the Helical DEMO Reactor FFHR-d1
T. Goto, J. Miyazawa, H. Tamura, T. Tanaka, N. Yanagi, A. Sagara and the FFHR Design Group
National Institute for Fusion Science, Toki, Gifu 509-5292, JapanA conceptual design activity for the helical DEMO reactor FFHR-d1 was started at the beginningof the last fiscal year. As the first step, design window analyses using a system design code arecarried out to fix main design parameters (major radius, magnetic field strength and fusion output).
A candidate design point is narrowed and examined by considering consistency of engineering andphysics constraints.
According to the steady-going progress in the LHD experiment [1], an LHD-type helical fusion
reactor is now foreseeable. In past design studies, a design with a commercially-attractive concept
has been considered and several design concepts, represented by the latest design FFHR-2m2 [2],were proposed. From the last fiscal year, a new conceptual design for a helical DEMO reactor,
FFHR-d1, was started by utilizing the knowledge of the past design studies. As the next step
reactor, FFHR-d1 aims at an early demonstration of maintainability, tritium self-sufficiency andnet electric power production. To determine main design parameters of FFHR-d1, design window
analyses are carried out using the system design code HELIOSCOPE [3]. In the evaluation of thecore plasma performance, the direct profile extrapolation (DPE) method [4], which directly
extrapolates the radial profiles observed in the LHD experiments to the fusion reactor condition
assuming a gyro-Bohm type parameter dependence, is used to enhance the reliability.
It was found that a reduction in minimum blanket thickness can moderate the requirement on theconfinement improvement with keeping the stored magnetic energy of the coil system, which isone of the key parameters in the engineering design. In the past design study, the stored magnetic
energy of 160 GJ has been considered as an achievable maximum value with the extension of the
ITER technical basis. If the minimum blanket thickness of ~ 70 cm is accepted, a steady-state, self
-ignition plasma can be designed with the core plasma parameters which can be reasonably
extrapolated from the present experimental results under the constraint of the stored magnetic
energy of ≤ 160 GJ. To examine engineering design feasibility of the thinned blanket system, anevaluation of a tritium breeding ratio (TBR) and fast neutron shielding performance of the blanket
system is carried out by neutronics calculations which reflect a consideration on the radial-build of
the helical coil, vacuum vessel and the blanket system.
The details of the result of the design window analyses and the selected main design parameters
will be described in the presentation.
References[1] A. Komori et al., Fusion Sci. Technol., 58, (2010) 1.
[2] A. Sagara et al., Fusion Eng. Des., 83 (2008) 1690 .
[3] T. Goto et al., Nucl. Fusion, 51 (2011) 083045 (6pp).
[4] J. Miyazawa et al., to be published in Fusion Eng. Des. (2011).

253

P2-95
Neutronics Investigations for Helical DEMO Reactor FFHR-d1
T. Tanaka, A. Sagara , T. Goto, N. Yanagi, S. Masuzaki, , H. Tamura, J. Miyazawa, T. Muroga,and the FFHR Design Group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japa
The conceptual design activity for the helical DEMO reactor FFHR-d1 has been started in
National Institute for Fusion Science. For an early demonstration of a steady state helical reactor, adesign window analysis has been conducted by balancing three parameters of (1) a confinement
improvement factor of a core plasma, (2) a stored magnetic energy of a superconducting magnet
system and (3) a blanket space for installation of a tritium breeding blanket and a radiation shield.
The decision of the minimum required blanket space which has a direct impact to the reactor sizeis the most major neutronics issue at this first stage.
Since a position of a magnetic axis shifting to the inboard side of the torus has been selected, theavailable blanket space will be smallest at the inboard side. For the average neutron wall loading2of 1.3 MW/m tentatively assumed for FFHR-d1, the minimum blanket space to provide anadequate radiation shielding performance for protection of a superconducting magnet has been
investigated by neutron transport calculations using MCNP5 code with a simple torus model. The
investigation showed that a combination of a 15 cm thick Flibe+Be/ferritic steel blanket and 55 cm
thick WC radiation shield followed by a 5 cm thick vacuum vessel could suppress radiations in asuperconducting magnet to acceptable conditions, i.e., a fast neutron flux of <~3 x 1010 n/cm2/s
and a nuclear heating of <~0.5 mW/cm3 . The minimum blanket space has been set to 70 cm.
The ratio of the inboard region where the blanket space is limited to 70 cm (15 cm thick breedingblanket and 55 cm thick radiation shield) was estimated to be only ~15 % of the torus from a
calculated magnetic field distribution. In contrast to the inboard region, a sufficiently thickbreeding blanket and radiation shield could be installed with the total thickness of >150 cm in theoutboard region. Fully covered tritium breeding ratios (TBRs) considering no neutron leakage
have been evaluated with a torus calculation model simulating a 15 cm thick breeding blanket atthe inboard side and a 60 cm thick breeding blanket at the outboard side. The fully covered TBRof >1.3 was obtained by the Flibe+Be/ferritic steel blanket (6Li enrichment: 90 %). The result
indicates that the tritium self-sufficiency could be expected in the FFHR-d1 concept evenconsidering neutron leakage through divertor ports and injection ports for plasma heating.
The neutronics investigation for FFHR-d1 has proceeded into three dimensional transport
calculations by simulating detailed non-axisymmetric helical blanket shapes. Results of the threedimensional evaluation of the neutronics performances will be discussed.

254

P2-96
Conceptual Design of Heating Devices for Heliotron-type DEMO Reactor
K. Tsumori, S. Kubo, H. Igami, Y. Yoshimura, H. Kasahara, K. Saito, T. Seki, T. Goto,
J. Miyazawa, T. Tanaka, A. Sagara and the FFHR design group
National Institute for Fusion Science, Toki, Gifu, 509-5292, JapanThe FFHR-d1 has been proposed as a conceptual heliotron-type DEMO reactor in National
Institute for Fusion Science [1]. The design study has started on the FFHR-d1 heating system
including ECH, NBI, and ICRH, whose total injection power is 80 MW. The FFHR-d1 plasma issurrounded with the twisted blanket and divertor, and arrangement of those heating devices is oneof the key issues to reduce the neutrons heat the FFHR-d1 plasma efficiently. The ECH isapplicable in low-density operation with normal electromagnetic modes, and even in high-density
operation with the electron Bernstein wave (EBW). The ECH system with EBW has moreaccessibility to the core plasma than the other heating devices. The waveguide outlet can beshielded from the neutron flux by applying a tilt-angle mechanism for microwave direction, andthe mechanism is also applied to adjust the heating position of the core plasma. The NBI beamline
will be installed to the FFHR-d1 tangentially, and the maximum beam-energy is set at 1.5 – 2.0
MeV for high-density operations. The beam is not required to sustain the plasma current as intokomak reactors, and the injection port can be closed after achieving self-ignition. The R&D of
the plasma- or photo-neutralizer is required to improve the efficiency of the port-through power tothe electric input. Present technology of ICRH is applicable for the DEMO-class reactor. A key
issue of the ICRH is the damages of the conventional ICRH antenna or the launcher exposed withhigh fluence of high-energy neutrons and alpha particles. A blanket imbedded antenna isinvestigated for an ICRF heating interface.
In this article, we report the specifications of the heating devices required for the FFHR-d1 reactor.
The arrangements of the devices reducing the neutron flux with the shield and blanket is also
going to be discussed, and the advantage of each device is compared.
References[1] A. Sagara et al., ISFNT-10, 11-16 September 2011, Portrand, US. O38.

255

P2-97
Fuel Particle Balance Study in FFHR DEMO Reactor
M. Tanaka, T. Goto, Y. Kozaki, A. Sagara and FFHR design group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanThe establishment of the fuel cycle system for a fusion DEMO reactor is one of key issues. The
first step to consider the fuel cycle system would be estimated the fuel particle balance and flow in
the fusion reactor system. The scenario of steady state fuel particle balance is as follows: the fuelparticles are supplied into vacuum vessel via the fueling system such as pellet injection, gas puff,
etc. A part of supplied fuel particles is ionized and burned in the core plasma, and the rest of fuelparticles transports to the edge plasma region. Then, although the fuel particles and the helium ash
are exhausted from the vacuum vessel, various interactions between plasma, neutral particle andmaterial are occurred in the edge and the diverter plasma region. As the results, a small amount of
the fuel particles is trapped in the wall by the retention and it is the loss of fuel particles. The
exhausted fuel particles and helium ash are separated and purified by fuel processing system. On
the other hand, tritium from blanket system and external deuterium are supplied for fuel cycle
system to replenish burned fuel particles. Then, the fuel particles from both the fuel processing
system and the blanket system are temporarily stored in the fuel storage bed and supplied to fuelinjection system according to demand. In the fuel cycle loop, however, a part of the fuel particles
are loss from the wall by permeation. These fuel particle losses are considered to be affected on
the fuel balance. As an inevitable consequence, tritium loss causes the requirement of higher
tritium breeding ratio in the blanket system.
To consider the fuel balance in fusion reactor system, hence, the fuelling rate or burning rate incore plasma, the tritium breeding ratio (TBR) and the tritium loss rates both the permeation fromthe wall and the retention in the vacuum vessel would be critical parameter. In previous studies,
the analytical model of fuel balance have been reported with regard to the efficiency of a tritium
processing system, the fueling scenario, the effect of tritium loss, and TBR, etc [1-3]. In this study,
we focus on the particle balance of tritium and deuterium in FFHR DEMO reactor with regard to
fueling efficiency into core plasma. Then, it is found that the fueling efficiency is one of criticalparameters to realize the fuel particle balance. The detail analytical model and calculation results
will be presented in the conference.
References[1] S. Kuroki, et al., J. Nucl. Sci. Technol., 19 (1982) 186.
[2] M. Nishikawa, Fusion Sci. Technol., 57 (2010) 120.
[3] H. Takenaga, et al., Fusion Sci. Technol., 57 (2010) 94.

256

P2-98Studies on Investigation of Behavior of Traveling Wave Direct Energy Converter Using a
Simulator Installed on GAMMA 10 Tandem Mirror with One Side Plugging

Shuuhei HARADA1), Shota MIYAZAKI1), Masahiro KUME1), Hiromasa TAKENO1),
Yasuyoshi YASAKA1), Kazuya ICHIMURA2), Yousuke NAKASHIMA2)
1)Kobe University, 2)University of Tsukuba
The large part of produced energy in D-3He fusion is released as the kinetic energy of protons. Theenergy is directly converted into electricity by a Traveling Wave Direct Energy Converter(TWDEC). Experimental verification has not been studied well, although the energy broadening offlux influences significantly on the conversion efficiency. The authors have investigated behavior
of TWDEC for flux with wide energy spread by application of a TWDEC simulator to end-loss
flux of GAMMA 10 tandem mirror [1]. As one method for those studies, end-loss flux of one sideplugging operation of GAMMA 10 has been employed for higher energy flux. This report willtreat that method.
The TWDEC simulator was installed on one of the end cells of GAMMA 10. The TWDEC
simulator consists of a modulator and a decelerator. The RF voltages, the frequency of which is
7MHz, are supplied to these electrodes. Two synchronized RF power is supplied to the modulator
and the decelerator, respectively, and relative phase difference between them (Δφ)can be set up
arbitrarily.
During one side plugging operation, end-loss flux of the plugging side is significantly reduced due
to confinement of plasma, while reduction of flux of no plugging side is not so large. To examine
the dependence on relative phase difference, the variation of energy distribution function due to
TWDEC operation has been investigated. The peak of function appears in the lower energy side in
the case of Δφ=π, while it appears in the higher energy side in the case of Δφ=0. Those variations
correspond to effects of deceleration and acceleration, respectively. In both cases, incident energy
components remain around peak of incident energy. To investigate in more detail, one dimensional
orbit calculation has been performed. In the calculation with change of Δφ, the effects of both
deceleration and acceleration have been found, and the similar results with experiments have beenobtained. Although some incident energy components have remained, the shapes of the peak are
more gradual compared with experiments. To investigate the cause of this difference, thenumerical calculation with varied parameter is planned.
The experiments using one side plugging have been continued to increase the number of data by
improvement of measurement system. More experimental and numerical results will be shown inthe conference.
References[1]H. Takeno, et al., ”Simulation Experiments of Traveling Wave Direct Energy Converter
on GAMMA 10 Tandem Mirror,” Transactions of Fusion Science and Technology, in Press.

257

P2-99Studies on Application of Cusp-Type Particle Separation
and Two-Stage Deceleration to Dense Plasma for Direct Energy Conversion
Y.Munakata, T.Kawaguchi, N.Sotani, H.Takeno, Y.Yasaka
K.Ichimura1), Y.Nakashima1)
Electrical and Electronic Engineering, Kobe University1)Plasma Research Center, University of Tsukuba
3

In the D- He fusion power generation, high plant efficiency can be expected by applying the directenergy conversion because the majority of the produced energy is kinetic energies of charged
particles. Moreover, harmful high energy neutrons are hardly generated, so the D-3He fusion
power generation is expected as a new excellent power generation method in the economical andenvironmental aspects. For efficient recovery, particles escaping from the reactor should be
separated from each other in energy and electrical polarity. For this purpose, use of a cusp-type
direct energy converter (Cusp DEC) is expected. Because of the difference of Larmor radius, ions
go to point cusp and are separated from electrons going to line cusp.
Working characteristics of Cusp DEC on plasma density is an important subject. Not only charge
separation, but also deflection of particle’s orbit due to space potential becomes a problem. Whenthe plasma density increases, ions stagnating ahead of the ion collector create the high potential
area there. This is a barrier for the low energy ions, and they are reflected before it reaches the ioncollector. To improve this problem, an additional energy conversion scheme with setting thesecond ion collector can be applied. The two-stage direct energy conversion has already beenreported[1], but its experimental verification is not enough. This paper treats those subjects on
working characteristics of Cusp DEC for dense plasma.
Using a multi-hole extraction electrode for the plasma source, 3-4 times increment of plasma
density has been achieved. Further increment of density is expected by enhancement of magnetic
field of plasma source or applying much further input power. As for two-stage direct energy
conversion, systematic experiments on the ion collector's location and size have been performedby simulating the ion space potential due to bias voltage of the point cusp ion collector. As aresult, it has been shown in the experiment that the collected low-energy ion increases byappropriately setting of the location of the point cusp ion collector. Moreover, it also has beenshown in the experiment that the electric power obtained by the second ion collector increases byenlarging the size of the collector. Details of the behavior of the particles due to the change ofthese ion collectors have been understood by using a numerical analysis.
Not only two-stage direct energy conversion, but also separation experiments for higher density
plasma are in progress and the detail will be presented in the conference.
References[1] Y. Yasaka et al., Nucl. Fusion, Vol.49, 075009, (2009).

258

P2-100
Internal Strain Measurement in Superconducting Composites by Neutron Diffraction
S. Harjo, K. Aizawa, T. Ito1), J. Abe, K. Osamura2), T. Hemmi3), X. Jin4), T. Nakamoto4),
S. Awaji5), K. Takahashi5)
J-PARC Center, Japan Atomic Energy Agency, Tokai-mura, Japan
1)Research Center for Neutron Science and Technology, CROSS, Tokai-mura, Japan
2)Research Institute of Applied Science, Kyoto, Japan3)Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka, Japan
4)High Energy Accelerator Research Organization, Tsukuba, Japan
5)Institute of Materials Research, Tohoku University, Sendai, Japan
It is well known that superconducting materials are very brittle and therefore they are needed to becomposites with generally metals to increase their strengths and/or ductilities. These facts lead tothe prediction that phase strains are always generated in the superconducting composites, after the
manufacturing and also during their uses, due to differences on thermal expansion anddeformability among the constituent phases. In contrast, it has been reported elsewhere thatinternal strain state has high influences on superconducting properties, e.g. the presence of internalstrain degrades critical current densities. The internal strain states in superconducting composites
depend on compositions of composites, thermo-mechanical treatment during manufacturing,superconducting conditions during uses. Clarifying internal strains behaviors accurately at various
conditions will help us to develop superconducting composites having optimum superconducting
performances during uses.
However, the internal strains in superconducting composites have not been measured so far
because of the composites configurations. Moreover, in the case of Cable-In-Conduit Conductors
for ITER cables, superconducting phases are located in a thick jacket. Neutrons that have highpenetration ability into metals may be a powerful tool to measure internal strains in the
superconducting composites by the diffraction technique. Here, we introduce internal strain
measurement techniques, our engineering materials diffractometer at J-PARC facility and ourtrials on measuring internal strains in several kinds of superconducting strands and tapes withvarious conditions, and a CICC for ITER TF coils.

259

P2-101
Numerical Investigation on Accuracy and Resolution of Contactless Methods for Measuring
jC in High-Temperature Superconducting Film: Inductive Method and Permanent Magnet
MethodT. Takayama, A. Kamitani, A. Saitoh1), H. Nakamura2)
Yamagata University, 4-3-16 Johnan, Yonezawa, Yamagata 992-8510, Japan
1) Graduate School of Science and Engineering, Yamagata University, Japan
2) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
A critical current density jC is one of the most important parameters for engineering
applications of high-temperature superconductors (HTSs). The standard four-probe method has
been generally employed for measuring jC. However, it may lead to the destruction of a sample
surface or to the degradation of superconducting characteristics. For this reason, contactless
methods have been so far desired for measuring jC.
For the purpose of contactlessly measuring jC in an HTS film, Claassen et al. have proposed
the inductive method for measuring the critical current density jC [1]. By applying an ac current toa small coil placed just above an HTS film, they monitored a harmonic voltage induced in the coil.
They found that, only when a coil current exceeds a threshold current, the third-harmonic voltage
develops suddenly. They conclude that jC can be evaluated from the threshold current.
Ohshima et al. proposed another contactless method for measuring jC [2]. In the method,while moving a permanent magnet above an HTS film, the electromagnetic force acting on thefilm is measured. As a result, they found that the maximum repulsive force FM is roughly
proportional to jC. This means that jC is estimated from the measured value of FM. This is called
the permanent magnet method.
In order to simulate a numerical code was developed for analyzing the time evolution of ashielding current density in a non-axisymmetric HTS film [3]. By using the code, the inductive
method/ permanent magnet method was reproduced. The results of computations showed that, theaccuracy of the two contactless methods degrade remarkably when the coil or the magnet is placednear the film edge.
The purpose of the present study is to simulate two contactless methods and to investigate theaccuracy and the resolution of the methods numerically.
References[1]J. H. Claassen, M. E. Reeves, and R. J. Soulen, Jr., Rev. Sci. Instrum., 62 (1991) 1004.
[2]S. Ohshima, K. Takeishi, A. Saito, M. Mukaida, Y. Takano, T. Nakamura, I. Suzuki, andM. Yokoo, IEEE Trans. Appl. Supercond., 15, (2005) 2914.
[3]T. Takayama, A. Kamitani, K. Hattori, A. Saito, and S. Ohshima, IEEE Trans. Appl.
Supercond., 21, (2011) 3360.
This work was supported in part by Japan Society for the Promotion of Science under a Grant-in-Aid for
Scientific Research (B) No. 22360042. A part of this work was also carried out under the Collaboration
Research Program (NIFS10KNXN178, NIFS11KNTS011) at National Institute for Fusion Science (NIFS),Japan. In addition, numerical computations were carried out on Hitachi SR16000 XM1 at the LHD Numerical
Analysis Server of NIFS.

260

P2-102
Adsorption Behavior of Lithium from Seawater using Manganses Oxide Adsorbent
T. Wajima, K. Munakata, T. Uda1)
Akita University, Akita, 010-8502, Japan
1) National Institute for Fusion Science, Toki, Gifu 509-5292, JapanThe deuterium-tritium (D-T) fusion reactor is expected to be a system to provide the mainelectricity in the future without any serous release of hazardous products such as a radioisotope oftritium, and it is the easiest fusion reaction to achieve. Lithium will be required in amounts
dependent on the reactor design concept. When liquid lithium is used as a tritium breeder and acoolant, lithium inventories are large [1]. Lithium is now recovered from the mines and from saltlakes which contains about 14 millions tons of lithium. Although the amount of lithium in thoseresources is quite insufficient at this point, alternative resources should be found to satisfy lithium
inventories for nuclear fusion plants and the increasing demand for battery and so on in the near
future. Seawater, which contains 2300 hundred million tons of lithium in total, has thus recently
become an attractive source of this element and the separation and recovery of lithium fromseawater by co-precipitation, solvent extraction, adsorption, etc. have been investigated. Amongthese techniques, the adsorption method is suitable for recovery of lithium from seawater because
certain inorganic ion-exchange materials show extremely high selectively for lithium ion only.
Among the inorganic adsorbents, spinel-type manganese oxides are interesting materials becauseof their extremely high affinity toward lithium ions only.
In this study, we prepared the lithium adsorbent by elution of spinel-type lithium di-manganesetetra-oxide (LiMn2O4), and the kinetics of the adsorbent for lithium ions in seawater was examinedusing pseudo-second-order kinetic model.
The adsorbent,which can be recover lithium ion from seawater, can be prepared from LiOH·H2O
and Mn3O4. With increasing temperature of solution, the kinetics of adsorption becomes faster andthe amount of lithium adsorbed on the adsorbent increases. The thermodynamic values, ΔG0, ΔH0
and ΔS0, indicated that adsorption was an endothermic and spontaneous process.
References[1] J.N.Hartley, Energy, 3 (1978) 337

261

P2-103
Application of Virtual Reality System to Fusion Science
H.Ohtani1,2), Y.Tamura3), A.Kageyama4), M.Nunami1), S.Ishiguro1,2), M.Shoji1)1)National Institute for Fusion Science, Toki, 509-5292, Japan
2)The Graduate University for Advanced Studies (SOKENDAI), Toki, 509-5929, Japan3)Konan University, Kobe, 658-8501, Japan4)Kobe University, Kobe, 657-8501, Japan
Virtual reality (VR) technology can give you a deep absorption into the VR world, and enables
you to analyze complex structures in a really three-dimensional space. This scientific visualization
is a powerful and useful tool in an analysis of simulation data and development of experimental
devices. National Institute for Fusion Science (NIFS), Japan installed VR System “CompleXcope”
based on CAVE system [1] in 1997 as a scientific analysis instrument for the simulation results.
The system produces a stereo, immersive and interactive view. In this paper, we report recent
development of an application of VR system to Fusion Science.
In the previous Toki Conference, we reported an integrated scientific visualization of plasma
simulation data and experimental device data in one VR world [2,3]. In the integrated
visualization, an isosurface of plasma pressure, a stream line of a magnetic-field line and a
trajectory of drift particle calculated from the MHD equilibrium simulation data [4,5] wereinteractively visualized in the Large Helical Device (LHD) vessel data with an objective
description (that is, a realistic device mechanical vessel) in the VR space [6,7]. Recently, we
modified the software for visualizing the simulation results to visualize the trajectories of multiple
particles simultaneously. Trapped and un-trapped particles can be investigated at the same time.
This success has opened up a lot of attractive possibilities for intuitively understanding the physics
of plasma, for aiding in the design and arrangement of the devices, and for confirming the field ofvision from the observation port in VR space.
References[1] Cruz-Neira, C. et al: ACM SIGGRAPH 93 (1988) 65.
[2] H.Ohtani, Y.Tamura, A.Kageyama and S.Ishiguro: IEEE Transactions on plasma, PP(99),(2011), 1.
[3] H.Ohtani, A.Kageyama, Y.Tamura, S.Ishiguro and M.Shohji: Plasma Fusion Research, 6,(2011), 2406027.
[4] K.Harafuji, T.Hayashi and T.Sato: J.Comput. Phys., 81, (1989), 169.
[5] A.Kageyama, T.Hayashi, R.Horiuchi, K.Watanabe and T.Sato: Proc.ICNSP, (1998), 138.
[6] H.Miyachi et al:IEEE Computer Society, (2005), 530.
[7] H.Miyachi et al:IEEE Computer Society , (2007), 536.

262

P2-104
Localization of Mobile Client using Wireless LAN and Natural Features of Camera-Image
Sequences for Indoor Structure
T. Umetani, T. Yamashita, S. Yamane, Y. Takabatake, Y. Tamura
Konan University, 8-9-1 Okamoto, Higashinada, Kobe 658-8501, Japan
Most nuclear fusion reactor is generally large and its buildings are also large and multi-storied.
Data on the physical position of mobile agents, for example, workers with personal electronic
devices or mobile robots, are important for achieving the surveillance and maintenance task in a
nuclear fusion reactor system. By integrating the data and stored information on objects andplaces, workers and robots can be provided with information on the location of the devices. Manykinds of localization services have been presented for mobile agents.
We introduce a method for localization of mobile client using wireless LAN and natural featuresof the camera-image sequences for the indoor structure. A Global Positioning System (GPS) is
widely used for localization [1]; the GPS is not suitable for localization in the indoor environment
and the area that is surrounded by many buildings. We use the stochastic localization system for
wireless mobile client using the signal strength from wireless LAN access points based on a sparse
Bayesian learning [2][3]. The estimated location of the client is used for searching the database ofnatural feature points obtained by the camera-image sequences [4]. The sensory data fusion
improves the localization accuracy. The presentation consists of two parts; (1) construction of theestimation system based on the signal strength and the database of the natural feature points, (2)localization and acquisition of the information based on the sensing data at the other time.
Our goal is to realize the localization method for surveillance or maintenance tasks in the real
nuclear fusion reactor system. The experimental testbed is a laboratory building; the building isthree-storied, and there are several stairwells and open-ceiling spaces in the building.
Experimental results show that the proposed method is feasible.
References[1] P. Enge, P. Misra, Proc. IEEE, 87 (1999) 3.
[2] T. Umetani, T. Yamashita, Y. Tamura, J. Robotics and Mechatronics, 23 (2011) 475.
[3] M. E. Tipping, J. Mach. Learn. Res. 1 (2001) 211.
[4] T. Sato, H. Koshizawa, N. Yokoya, Int. J. Virtual Reality. 9 (2010) 37.
This work was supported in part by the Hirao Taro Foundation of the Konan University Association for
Academic Research, Japan.

263

P2-106
Behavior of core plasma potential and fluctuation in end-diverter simulation experiments on
the tandem mirror GAMMA10
M.Mizuguchi1), M.Sakamoto1), M.Yoshikawa1), Y.Miyata2),
M.Aoyama1), K.Furutachi1), T.Iwai1), Y.Akabane1), K.Jitsukawa1), S.Matsuyama1),
Y.Nakashima1), M.Ichimura1), T.Imai1)
1)Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-0006, Japan
2)Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan
Recently, diverter simulation experiments have been performed by using a mirror end of thetandem mirror GAMMA10, where the plasma is initiated by plasma guns and sustained using ion
cyclotron heating (ICH). The axial confinement potentials are produced using electron cyclotron
heating (ECH) at both end mirrors of GAMMA10. In this study, we have measured and analyzed
potential, electric field and those fluctuations of the core plasma at the central cell using a gold
neutral beam probe (GNBP) to investigate influence of the insert of the diverter plate to the end
region on the core plasma. GNBP can measure both potential and density fluctuation
simultaneously. Then it is useful for analysis of the cross spectra between them and the particle
fluxes induced by them. Recently, GNBP has been developed to obtain local electric field in asingle plasma shot with simultaneous multipoint plasma potential measurement[1]. The error of
the potential measurement is about ±10 V. The time and spatial resolutions are estimated to beabout 3 μs and 10 mm (i.e. a beam diameter), respectively. In the plasma with ICH and ECH, thepotential fluctuation of the core plasma occurred in the low frequency region (1-20kHz). When thediverter plate was inserted to the mirror end in the (ICH+ECH) plasma, the fluctuation greatly
increased and the mode number of the fluctuation changed from m=+2 (the direction of electronic
diamagnetic drift) to m=-3. The potential also increased due to the insert of the diverter plate. The
increase in the potential and its fluctuation occurred in the whole radial region of the plasma,
which was beyond the region at the central cell projected along the magnetic field from thediverter plate. On the other hand, in the plasma with ICH only, the fluctuation was not observed in
the low frequency region, and the potential did not change even though the diverter plate wasinserted.
References[1] Y. Miyata et al., Plasma Fusion Res. 6, 1202090 (2011).

264P2-107
Measurement of LHD divertor heat flux response of plasma detachment with the Hydride
Directional Langmuire Probe
H.Matsuura, K.Nagaoka1), T.Morisaki1), S.Masuzaki2), M.Osakabe2)
Osaka Prefecture University, Sakai Naka-ku, Osaka 599-8570, Japan
1)National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
In the design of fusion reactors like International Tokamak Experimental Reactor (ITER), vast
heat flux ( > 10 [MW/m2]) is expected to flow onto divertor target plates. In order to reduce thisheat load, enhancement of plasma detachment with cooling gas puff has been proposed. The
concept of thermal probe method has been proposed to measure heat flux directly in the reactive
plasma and divertor plasma.[1-3] Since present fusion plasma devices have shorter discharge
duration than heat diffusion time of the probe tip, however, heat flux response of plasma conditionchange such as detachment has not been reported yet.
The Hydride Directional Langmuire Probe (HDLP) recently equipped in Large Helical Device(LHD)[4] has the heat diffusion time with the same order as plasma discharge. Heat flux must be
determined with the solution of heat conduction problem in probe tips and the data of Thermo
Couple (TC) embedded in the tip. In order to deal with the change of plasma heat flux, timedependent heat flux is modeled as the summation of step-like heat flux with both positive andnegative amplitude. The size of each step is determined so as that the summation of theirtemperature response reproduces the observed temperature variation data. If a generalmathematical method such as singular value decomposition or Marquart method is applied to thisfitting procedure, reconstructed heat flux shows unreasonable large positive and negative value.
We consider the casualties of the heat conduction problem and develop a new iterative
optimization method to determine each component step-like flux amplitude.
Present modeling was applied for the first time to the calibration experiment of HDLP with
Nd:YAG laser and reasonable heat flux variation could be reproduced. Plasma heat flux analysis
for many discharges is also done successfully. By using Neon Gas puffing, plasma detachment canbe controlled in LHD plasma. TC data obtained by HDLP set on the divertor leg shows theincrease after plasma discharge. If plasma detachment occurs, this increment becomes small. By
using present modeling, the onset of heat flux decrease was confirmed to agree well with those ofion saturation current.
References[1]E.Stamate et al., Appl. Phys. Lett., 80 (2002) 3066.
[2]H.Matsuura and T. Jida, Conrtib. Plasma Phys., 46 (2006) 406.
[3]H.Matsuura et al., Plasma Fus. Res., 5 (2010) S1045.
[4]K.Nagaoka et al., Rev. Sci. Instr., 79 (2008) 10E523.
This work is partially performed with the support and under the auspices of the NIFS Collaborative Research
Program. (NIFS10KLPR001)

265

Plenary / Invited / Oral
Sessions(Wednesday 30th November)PL-5
Challenges on the Path to DEMO and Activities
in the Power Plant Physics and Technology Department under EFDA
Gianfranco Federici
EFDA CSU Garching, Boltzmannstr.2, Garching 85748 (Germany)
The technical basis for designing a next-step DT burning plasma experiment has greatly expanded
during the last two decades thanks mainly to remarkable improvements in plasma performance andcontrol in today’s machines and advances in various areas of physics and engineering. Integratingand extending these advances toward long pulsed or steady state burning plasmas is now the focus
of international tokamak research, which is proceeding with the construction of ITER in the southof France and JT60-SA in Japan.
However, beyond ITER there are still several physics and technology issues, which must be
addressed and resolved for a Demonstration Power Fusion Reactor (DEMO). The most important
technology problems to be solved include the qualification of resilient materials for in-vessel
components, the development of sound technological solutions for the divertor and of optimised
remote maintenance schemes for high machine availability, the achievement of adequate thermal
efficiency and tritium breeding, and the reliability and efficiency of heating and current drive
systems. Among the physics questions, the divertor power exhaust, the definition of a reliable
modes operation, the need to guarantee plasma performance at high density, the avoidance andmitigation of disruptions and ELMs, which can damage the in-vessel components, are the mostimportant.
A Power Plant Physics and Technology Department (PPPT) has been recently established under
EFDA, with the purpose to begin a coordinated effort in Europe (building on efforts done in thepast) to quantify the key physics and technology prerequisites for DEMO and to address theremaining outstanding physics and engineering problems that need to be solved to confirm ourability to design a device that meet the requirements. In particular, it shall (i) define a set oftechnical characteristics for DEMO and, subsequently, carry out the design work necessary toestablish its conceptual design; (ii) define future research and development needs and to draw up
cost, and schedule estimates; and (iii) carry out, in a coordinated manner, specific validating R&D
supportive of the conceptual design activities;
This talk will describe the main technical challenges on the path to DEMO and the organization of
the EFDA PPPT activities with emphasis on the elements of the work programme 2011 and thoseplanned for 2012.

269

I-11
Progress of ITER Test Blanket Module Development in Japan
M. Enoeda, H. Tanigawa, T. Hirose, S. Suzuki, K. Ochiai, C. Konno, Y. Kawamura,T. Yamanishi, T. Hoshino, M. Nakamichi, H. Tanigawa, K. Ezato, Y. Seki, A. Yoshikawa,
D. Tsuru
Japan Atomic Energy Agency, Naka, Ibaraki, 311-0193 Japan
ITER Test Blanket Module (TBM) program is the first module scale breeding blanket performance
test in the real DT fusion reaction environment which will be realized in ITER. Therefore, it isregarded as one of the most important development step of breeding blanket for DEMO.
For the TBM program, Japan has the plan to test Water Cooled Ceramic Breeder (WCCB) TBM as
the primary candidate breeding blanket. Japan also has intention to participate and perform liquid
breeder TBM testing liquid breeder TBM testing, based on the firm achievements and activities
which Japan has on the development of liquid breeder blanket technologies. ITER construction
activity is showing progress toward the first plasma. Keeping consistency with ITER construction
activities, TBM design and development activities are also being performed. This paper overviews
the TBM testing program and the recent achievements of the development of the module
fabrication technology and tritium production technology of the WCCB TBM in Japan.
Fabrication technology development of the module structure is one of key technologies. In Japan,
fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of thefirst wall and side walls have succeeded, based on the elementary technology development forvarious joining technologies, such as welding and Hot Isostatic Pressing (HIP) joining. Recently,
the real scale partial mockup of the back wall was fabricated clarify the fabrication procedure of
the back wall, whose thickness is up to 90 mm. Important key technologies are almost clarified forthe fabrication of the real scale TBM module mockup. From the view point of testing andevaluation of solid breeder TBM tritium production performance in fusion environment,
development of the technology of the blanket tritium recovery, development of advanced breeder
and multiplier pebbles and the development of the blanket neutronics measurement technology arealso performed. Also, tritium production and recovery test using D-T neutron in the Fusion
Neutronics Source (FNS) facility has been started as the verification test of tritium production
performance.
A part of research on tritium performance of TBM was supported by KAKENHI (Grant-in-Aid for Scientific
Research) on Priority Areas "Tritium for fusion" from the Ministry of Education, Culture, Sports, Science andTechnology of Japan.

270

O-13
Effect of lithium purity on in-situ formation of Er2O3 oxide layer on V-4Ti-4Cr alloy

V. Tsisar1), O. Yeliseyeva1), T. Muroga2), T. Nagasaka2)
1) Physical Mechanical Institute of National Academy of Science of Ukraine, 5, Naukova Street,
Lviv 79601, Ukraine2) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
The in-situ formation of Er2O3 oxide layers on the surface of V-alloys pre-charged with oxygen in
the liquid Li doped by Er has been recognized as a promising technology to form the viable andfunctional insulator coatings for V-Li blanket concept. Most of the works were carried out in"pure" Li. However in the real system the contamination of the liquid lithium by non-metallic
impurities is inevitable. The aim of this work is to determine the influence of N dissolved in Li on
the formation of Er2O3 oxide coating on the surface of V-4Ti-4Cr alloy.
The samples of V-4Ti-4Cr alloy (NIFS HEAT) were pre-charged by oxygen in Ar-7%O2
atmosphere (oxidation at 700°С for 3 h followed by the vacuum annealing at 800°С for 34 h).
Weight gain of samples averaged 0.35 mg/сm2, while the depth of the hardened zone (~ 680
kg/mm2) reached 100 μm. The formation of Ti–O net structure in the hardened zone wasconfirmed by TEM. Then samples were exposed to the N-contaminated (CN[Li]≤0.5 wt%) liquid Li
doped with Er at 700°C for 100 h. The post-test examinations (weight change, hardness, XRD,
SEM/EDX, OM) showed that samples lost the oxygen while the nitride phases (TiN, VN0.2) wereformed on the surface at the same time. The weight of the Er sample does not change, i.e. Er was
not dissolved in Li. It was supposed that formation of ErN on the surface of Er sample prevented it
dissolution since from the thermodynamic point of view the affinity of Er to N is higher than thatof Li. As a result the Er-oxide coating was not formed on the surface of V-alloy.
The next experiment was carried out in Li pre-cleaned by getter material (Zr, Nb-Ti alloy, V) at
700°С for 24 h. Then pure Li was transferred into the reaction zone in which samples of V alloys
and Er were placed. The test was carried out at 650°С for 100 h. Contrary to the N-containing Li
the Er2O3 oxide film was formed on the surface of the V-alloy successfylly. Samples of V-alloy
showed weight gain while Er sample demonstrated weight loss indicating dissolution of Er in Li.
It was concluded that in-situ formation of Er2O3 oxide coating on the surface of V-alloys depends
strongly on the purity level of Li with regard to the N. The cleaning of Li by means of getters is asine qua non of successful mass transfer in Li[Er]-V,Ti[O] system.

271

O-14
Silicon Carbide Nanopillar Fabrication on Silicon Carbide Substrate by Direct CF4 Etching
Using the Gas Discharge Ion Source
A.R. Gines, J. Salamania, M. Villanueva, H. Salapare, H. Ramos
Plasma Physics Laboratory, National Institute of Physics, University of the Philippines, Diliman,
Quezon City, Philippines
This paper reports the fabrication of silicon carbide nanopillars on a silicon carbide substrate. The
samples were etched with low energy CF4 ion shower produced by a gas discharge ion source. The
flow rate and discharge current were varied to determine the effects on the growth of the siliconnanostructures. SEM images of the treated samples were investigated. Furthermore, thetopography of the sample surfaces were analysed through AFM.
References[1]H.Salapare III, G. Blantocas, V.Noguera, H.Ramos, Low-energy Hydrogen ion shower
treatment of PTFE material, Applied Surface Science, Vol. 255, Issue 5, Part 2, Dec. 2008[2]C.Welch, A.Goodyear, T.Wahlbrink, M.Lemme, T.Mollenhauer,Silicon etch process
options for micro- and nanotechnology using inductively coupled plasmas, Microelectronic
Engineering, Vol. 83, Issues 4-9, Sept. 2006

272

I-12
Assessment and comparison of pulsed and steady-state tokamak power plants
Farrokh Najmabadi
Dept. Electrical & Computer Eng., Center for Energy Research, UC San Diego
The current visions for steady-state tokamak plants are mostly based on operation with thereversed-shear mode in order to achieve a high bootstrap current fraction at a high plasma β. Thismode of operation has been demonstrated in current experiments only in transient fashion mainly
due to the lack of sufficient mid-plasma current drive power. Furthermore, it appears that reversed-shear mode require stringent control on plasma current profile and may require stabilization ofresistive wall modes. As such, there is a wide-spread belief that steady-state tokamak power plants
require major extrapolations from current data base. However, the first vision of steady-state
tokamak power plant, introduced around 1990 simultaneously in US (ARIES-I) and in Japan(SSTR), did not utilize revered-shear mode and operated with a monotonic q profile. In fact,
ARIES-I proposed plasma profiles are very similar to those of the "hybrid" mode in tokamak
experiments.
In this paper, we compare the operation and needs of such a "conservative" steady-state tokamak
plant with a pulsed-plasma version. Plasma physics constraints on both systems are identified andcompared, the major difference is the need for ~100-150 MW of current-drive power for steadystate system and the limited control on "optimizing" plasma performance in a pulsed-plasma
device. Most of the paper, however, is devoted to the impact of pulsed-plasma operation oncomponents: thermal energy storage, issues related to superconducting magnets (cyclic fatigue,larger PF coils, rapid PF ramp rates, large and expensive PF power supplies, large joule losses incryogenic structures), reduced performance in fusion power technologies (first wall, blanket, anddivertor), and reliability of complex components under cyclic operation. Overall, we find that theadditional costs associated with the pulsed-plasma operation far exceed savings due to theelimination of the current-drive power.

273

I-13
Design and R&D activities for the LHD-type heliotron DEMO
N. Yanagi, T. Goto, J. Miyazawa, H. Tamura, T. Tanaka, K. Tsumori, M. Tanaka, R. Sakamoto,
M. Yokoyama, O. Mitarai 1), S. Imagawa, T. Muroga, A. Sagara and the FFHR Design Group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
1)Tokai University, 9-1-1 Toroku, Kumamoto 862-8652, Japan
Conceptual design studies on the LHD-type heliotron DEMO reactor FFHR-d1 are being
conducted at NIFS through domestic and international collaborations. Intensive R&D activities arebeing simultaneously carried out for various components. The key requirements are the safety, net
electric power production, tritium self-sufficiency and maintainability. The design parameters are
examined by a system design code incorporating the confinement scaling deduced from the LHD
experiments. The 3D transport of neutrons and gamma rays shows that the distance between thehelical coils and the plasma at the inboard side of the torus should be >0.8 m with advanced
shielding materials used in the blanket. The tritium breeding ratio must be supported by thicker
breeder blankets at the outboard side. The present design gives the major radius 15.6 m, toroidal
magnetic field 4.7 T and stored magnetic energy 160 GJ. The low neutron wall loading assureslong lifetime blankets. The structural integrity of the in-vessel components including the built-in
helical divertors is examined. The start-up scenario with external heating (NBI, ECH, ICRF) is
discussed in accordance with the fueling scheme.
The superconducting magnet system is being designed along with development of 100 kA, 13 T
conductors. Cable-in-conduit conductors are supposed to be the primary option as an extension of
the ITER technology using low-Tc superconductors (LTS). To reduce the degradation of criticalcurrents due to strains, Niobium Aluminum is considered for strands. To simplify the winding
structure, an indirectly-cooled magnet concept seems attractive and a solid LTS conductor with analuminum-alloy jacket is being developed. The winding process of the continuous helical coils
using LTS conductors is being investigated in detail. High-Tc superconductors (HTS) using
YBCO tapes could be a counter option that proposes an innovative winding method by joining half
-pitch conductors on site. The supporting structure is being designed using a 3D FEM to withstandthe large electromagnetic stress while supplying large maintenance ports.
The liquid breeder blanket using molten salt or liquid metal is being designed with emphasis on
safety. A circulation testing facility using Flinak and liquid lithium has been constructed andcorrosion tests are being carried out. Suppression of tritium permeation and reduction of MHD
effects in liquid blankets are important subjects. Large-area ceramic coatings with Erbium oxideare being tested. The operation temperature of blankets can be enhanced by suppressing thermal
creep deformation of reduced activation ferritic/martensitic steels and vanadium alloys by oxide
dispersion strengthening.

274

O-15
Critical Issues of Burning Plasma, Engineering, Economic and Environmental Assessments
in Steady-State Fusion Reactors
K. YAMAZAKI, T. OISHI and H. ARIMOTO
Nagoya Univ., Nagoya, Aichi 464-8603, Japan
In order to search for attractive steady-state tokamak fusion reactors, burning plasma studies andengineering system design analyses were carried out focusing on advanced plasma operations withinternal transport barrier (ITB) and high bootstrap current (BSC) fraction, high field and highneutron wall loading limits, high economic efficiency without frequent blanket exchanges andenvironmental assessments related to global warming gas emission.
Assessments have been done using integrated toroidal transport linkage analysis code (TOTAL)
with ITB [1], high BSC [2], impurity and NTM [3] effects, and global Physics-Engineering-Cost
system code (PEC) [1] focusing on steady-state operations of D-T tokamak fusion reactors.
Analyses are extended to spherical tokamak, helical, inertial and advanced fuel reactors [4].
Plasma simulation models clarified the requirement of deep penetration of pellet fueling to realize
steady-state advanced burning plasma operation and appropriate non-inductive current drive
profile adjustment for ITB with high BSC fraction ( >90%) [2] in steady-state tokamak reactors.
Engineering assessment shows the relationship among the maximum magnetic field strength,
neutron wall load, blanket thickness, thermal efficiency, operational period and reactor system
availability. Economic and environmental system analysis has been carried out [5-6], andcomparative studies among other conventional electric power plants are done, such as oil, coal,water, solar, wind, fission power plants, with respect to cost of electricity (COE), CO2 emission
amounts and energy payback ratio (EPR). The assessment shows that fusion power plants have an
advantage in energy payback ratio (EPR), but have disadvantages in COE and CO2 emission, incomparison with fission reactors.
References[1] K. Yamazaki et al , Nucl. Fusion 49 (2009) 055017[2] J. Mano et al., “Current profile control for high bootstrap current operation in ITER”, thisconference.
[3] D. Kurita et al., “Stabilization Analysis of Neoclassical Tearing Mode in Fusion
Plasmas”, this conference.
[4] T. Kondo et al., “Economic evaluation of D-T, D-3He, and catalyzed D-D fusion
reactors”, this conference.
[5] Oishi et al., “Comparative study of cost models for tokamak DEMO fusion reactors”, thisconference.
[6] K. Ban et al., “Life Cycle Assessment for Energy Payback of Spherical Tokamak
Reactors”, this conference.

275

O-16
Assessment of the physics and technology requirements for a fusion DEMO
R.C. Wolf1), J. Aktaa2), S. Antusch2), W. Biel3), L.V. Boccaccini2), B. Bornschein2), S. Brezinsek3),
F. Cismondi2), C. Day2), D. Demange2), U. Fantz1), W. Fietz2), P. Franzen1), G. Gantenbein2),
B.E. Ghidersa2), Th. Giegerich2), S. Gunter1), V. Hauer2), R. Heller2), K. Hesch2), Y. Igitkhanov2),
A. Kallenbach1), S. Kern2), R. Knitter2), R. Konig1), J. Konys2), H. R. Koslowski3), K. Lackner1),
I. Landman2), M. Lehnen3), R. Lietzow2), J. Linke3), C. Linsmeier1), H. Meister1), P. Mertens3),
M. Mittwollen2), A. Moslang2), O. Neubauer3), M. Noe2), P. Norajitra2), J. M. Noterdaeme1),
G. Pautasso1), V. Phillips3), D. Reiter3), M. Rieth2), J. Roth1), G. Tardini1), W. Treutterer1),
F. Schauer1), S. Schlachter2), D. Strauß2), B. Unterberg3), K. Weiss2), J-H. You2), H. Zohm1)
1)Max-Planck-Institute for Plasma Physics, IPP-EURATOM Association, Germany2)Karlruhe Institute for Technology, KIT-EURATOM Association, Germany3)Forschungszentrum Jülich, Institute for Energy Research, FZJ-EURATOM Ass., Germany
In 2010 the German fusion laboratories formed a closer collaboration to assess the physics andtechnology requirements of a fusion power plant. The activity is organized in six physics topics(long pulse / steady state / high beta operation, high density operation, plasma wall interface,disruptions, plasma diagnostics and integrated control, combination of first principles and reactor
codes) and seven technology topics (heating and current drive, structural and functional materials,
in-vessel components design and integration, fuel cycle, magnet design, maintenance and remotehandling, safety and licensing). In a first step the interdependencies between the topics have been
analyzed and critical areas have been identified. Examples are (i) the heat flux limits to thedivertor targets, assuming tungsten as a plasma facing material and He-cooling, (ii) current drive
efficiencies to achieve steady state operation in tokamaks, (iii) the understanding of the density
limits in tokamaks and stellarators to maximize the fusion reaction rate and ease the requirementsfor pumping neutral gas in the divertor chamber. In a next step these requirements are
implemented in three essentially 0-D power plant models which are presently under development:(1) A pulsed tokamak which does not rely on a large current drive fraction. Such a tokamak will
have a pulse length of 10-15 hours, but with the possibility of reduced confinement (HH ~ 0,85).
(2) An advanced tokamak aiming at steady state operation, decreasing the size, but requiring
improved confinement and significant power for current drive, and reducing the wall life time. (3)An optimized stellarator which is basically a further development of the Wendelstein 7-X concept.
Recent engineering studies suggest that coils of the size of ITER and with the ITER
superconductor technology can be used. The paper will report on the main conclusions from theassessment of the interdependencies of the different physics and technology requirements and givean initial comparison of the power plant models.

276

Plenary / Invited / Oral
Sessions(Thursday 1st December)PL-6
History and prospects of applied superconductivity technology for fusion magnets
M. Noe, P. Komarek
Karlsruhe Institute of Technology, Institute for technical Physics, D-76344 EggensteinLeopodshafen, Germany
One of the most fascinating and challenging large scale applications of superconductivity aremagnets for fusion reactors. Up to now, many tokamak and a few stellarator facilities using
superconducting magnets have been built or are under construction with the most prominent being
the ITER project currently realized in Cadarache, France. The magnetic confinement in fusion
devices places heavy requirements on the different magnets. In general, currents of more than 10
kA have to be handled in magnetic fields of more than 10 T resulting together with the large
dimensions in huge forces and high energies. As a consequence, the investment insuperconducting magnets and cryogenics takes a significant fraction of the overall cost for a fusion
device.
This contribution summarizes the history, the state-of-the-art and the future developments of thesuperconductivity technology for fusion magnets. Major highlights of the significant progress inthis field are shown and the magnet technology for large scale fusion magnets e.g. Tore Supra,
LCT, TFMC, ITER, LHD and W7-X is presented. As far as possible, material aspects, conductor
technology as well as magnet design and manufacturing issues are covered.
Finally, the prospects for high temperature superconducting materials in fusion magnets are
summarized because these materials have a great potential to decrease magnet cost, increase
reliability and efficiency of fusion power plants and simplify the construction of a fusion power
plant.

279

I-14
What is the nature of the critical current density in superconducting materials
under strain in high magnetic fields?
J. S. Higgins, P. Sunwong, Y. Tsui, M. J. Raine, and D. P. Hampshire

Centre for Materials Physics, Superconductivity Group, Physics Department,
Durham University, South Road, Durham, DH1 3LE, UK
Understanding the magnetic field (B), temperature (T), strain (ε) and angular dependence (θ) of
the critical current density (Jc) in practical superconducting wires/tapes is critical for optimizing
magnet design using superconducting technology. We review critical current measurements
reported on both LTS and HTS superconductors. For the LTS materials, we present the evidence
for the well-established scaling law for Jc, that can be best understood in terms of flux pinning.
For HTS materials, we present recent measurements that make it possible to derive a scaling
relation for Jc(B, T, ε, θ) and discuss whether in contrast to LTS materials, wave-like properties
provide the best description of the mechanism that limits Jc .

280

O-17
Neutron Diffraction Study of Internal Strain in Nb3Sn Cable-In-Conduit Conductors
T. Hemmi, Y. Nunoya, S. Harjo1), H. Kajitani, N. Koizumi, H. Nakajima, K. Aizawa1),
S. Machiya2), K. Osamura3)

Fusion Research and Development Directorate, Japan Atomic Energy Agency, 801-1 Mukoyama,Naka-shi, Ibaraki, 311-0193 Japan
1)J-PARC center, Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai-mura, Naka-gun,Ibaraki, 319-1195 Japan
2) Daido University, 10-3 Takiharu-cho, Minami-ku, Nagoya-shi, Aichi 457-8530 Japan
3) Research Institute for Applied Sciences, 49 Tanaka Ohi-cho, Sakyo-ku, Kyoto, 606-8202 Japan
Internal strain in Cable-In-Conduit Conductors (CICC) is caused by differences in the coefficientsof thermal expansion between Nb3Sn strands and the stainless steel jacket over a temperature
range of 5 - 923 K. In addition, transverse electromagnetic loading of 800 kN is generated by acurrent of 68 kA and a magnetic field of 11.8 T in the case of ITER TF coils. The superconducting
performances of Nb3Sn strands change significantly, depending on the presence of strain in
Nb3Sn. The presence of internal strain in Nb3Sn cables is important to evaluate thesuperconducting performance. However, the strain of strands in the conductor has not beenmeasured so far because of the cabling configuration and their location in a jacket. Internal strain
can be determined by neutron diffraction measurement using Takumi of J-PARC[1]. Neutron
diffraction measurement becomes a strong tool for evaluating directly the internal strain of Nb3Sn
in CICC. The neutron diffraction measurement will be applied to evaluate the large bending due tothe buckling of Nb3Sn strands in CICC samples, especially the slippage of the superconducting
cable from low field zone to high field zone along the short CICC sample. In this presentation, test
results of high field zone in a cut CICC sample using neutron diffraction procedure and the future
plan of the neutron diffraction study will be presented and discussed.
References[1]T. Hemmi et al., IEEE Trans. Appl. Supercond., 21-3(2011) 2028
The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

281

O-18
Internal strain states of high strength Nb3Sn wires and cables
S. Awaji1), K. Takahashi1), H. Oguro1), K. Watanabe1), S. Harjo2), T. Ito2), H. Arima2), K. Aizawa2)
1) Institute for Materials Research, Tohoku University, Sendai 980-8577, Japan2) J-PARC Center, High Energy Accelerator Research Organization, Tokai-mura 319-1195,Japan
For high field superconducting magnet applications, it is necessary to develop the high strength
superconducting wires and cables. Rutherford and cable-in-conduit conductors are typicalconductors of Nb3Sn. However, internal strain states of Nb3Sn phase in the strands and cables arecomplicated but very important. We evaluated the internal strain of the high strength CuNb/Nb3Sn
strands and twisted cables with stainless steel under tensile stress by neutron diffraction. We foundthat 3-dimensional strain states of Nb3Sn in the strands under tensile stress are determined by theelastic-plastic deformation of the constituents of the composite wires, by the detailed analysis of
the each strain of the composed materials of the strand. Therefore, the mechanical deformation
history such as pre-bending and pre-loading changes the internal strain state, which is closely
related to the superconducting properties. In addition, the strain states of the cables, which consistof many strands, are more complicated. We also evaluated internal strain properties of the twisted
cable under the tensile stress which consists of 3 CuNb/Nb3Sn and 4 stainless steel strands. The
reduction of the internal strain of the Nb3Sn under tensile stress due to the reinforcement could beseen directly in the twisted cable. This suggests that a reinforcement effect by the stainless steel iseffective for the twisted cable.

282

I-15
Particle Simulations on GPU supercomputing systems
V. K. Decyk and T. V. Singh
University of California, Los Angeles
Graphical processing units (GPUs) have become a strong candidate for high performancesupercomputing this decade. GPUs have evolved from specialized devices for graphics to general
purpose computing. One attractive feature of GPUs is that the cost of floating point operations perwatt is substantially lower than with traditional processors. Although GPUs appear to be an exotic
and specialized architecture, they are constructed from familiar parts: a hierarchical collection of
SIMD (vector) processors. As a result, one can develop an abstract view of the hardware andprogramming languages that support this view (such as OpenCL and CUDA). This abstract
machine not only represents GPUs, but also other devices such as Intel multi-core processors. As aresult, there is hope in developing portable, scientific applications that will support emerging
supercomputing architectures.
At UCLA, we have recently installed a 96 node cluster with 288 NVIDIA GPUs called Dawson2.
The system ranked 148 in the most recent compilation of the top500 computers in the world, andhas 129,024 GPU processing cores and 1152 cpu cores. In preparation for this machine, wedeveloped algorithms for Particle-in-Cell (PIC) codes which run on a single GPU[1]. Thisalgorithm reorders particles every time step and makes use of a sorting cell that contains a small
number of grids. The algorithm has a small number of parameters that allow one to adapt the code
to different hardware architectures. We obtained speedups of around 50-60 compared to traditional
CPUs. We are now designing algorithms that will allow multiple GPUs to work together using a
hybrid MPI-CUDA domain decomposition. Our design goal is to create a PIC code withinterchangeable parts that could be adapted to emerging exascale supercomputers.
References[1] V. K. Decyk and T. V. Singh, “Adaptable Particle-in-Cell algorithms for graphicalprocessing units,” Computer Phys. Comm. 182, 641 (2011).
*Work supported by USDOE and UCLA IDRE.

283

I-16
Kinetic Simulations of Neoclassical and Anomalous Transport Processes in Helical Systems
H. Sugama1,2), T.-H. Watanabe1,2), M. Nunami1), S. Satake1)
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan2) The Graduate University for Advanced Studies, Toki, Gifu 509-5292, Japan
Quantitative predictions of transport fluxes of particles, momentum, and heat in magneticallyconfined plasmas are a critical issue for the design of fusion reactors. In this paper, recent resultsfrom kinetic simulations of neoclassical and anomalous transport processes in helical systems suchas the Large Helical Device (LHD) are reported. In contrast to axisymmetric systems such as
tokamaks, the radial electric field in helical systems can be determined by the condition of
ambipolar neoclassical radial particle fluxes. Using the delta f particle simulation code, FORTEC3D [1], the radial profiles of the neoclassical transport fluxes and the radial electric field in the
LHD are calculated and compared with experimental results. It is observed that the radial electricfield profile is well predicted by the ambipolar neoclassical particle flux condition even though theanomalous transport is dominant over the neoclassical one in LHD plasmas. Using the gyrokinetic
Vlasov simulation code, GKV-X [2], simulations of the ion temperature gradient (ITG) turbulence
are done including precise equilibrium conditions corresponding to the LHD experiments.
Comparisons between the profiles of the turbulent ion heat flux obtained from the GKV-X
simulations and from the high ion temperature LHD experiment show a good agreement. Besides,
the simulations show that generation of the zonal flows becomes more evident when the rotationaltransform increases or the inward shifted plasma configuration is realized. It is theoretically
predicted that, in helical systems, the macroscopic or background radial electric field can enhance
generation of zonal flows leading to the further reduction of turbulent transport and the favorable
isotope effect of the ion mass [3]. These effects of the background electric field are confirmed by
the poloidally global simulations of the ITG turbulence using the extended gyrokinetic Vlasov
simulation code [4].
References[1] S. Satake, R. Kanno, and H. Sugama, Plasma Fusion Res. 3, S1062 (2008).
[2] M. Nunami, et al., Plasma Fusion Res. 6, 1403001 (2011).
[3] H. Sugama and T.-H. Watanabe, Phys. Plasmas 16, 056101 (2009).
[4] T.-H. Watanabe, H. Sugama, and M. Nunami, 23rd IAEA FEC (2010), THC/6-1.

284

I-17
High Performance Integrated Simulation of Fusion Plasmas
C.S. Chang, S. Ku, and the CPES Team1)
Princeton Plasma Physics Laboratory, Princeton, NJ, USA
1)SciDAC Proto-FSP Center for Plasma Edge Simulation, USA
A high fidelity integrated simulation requires the component codes to be as first-principles as
possible. The most fundamental first-principles component codes available today are full-function
gyro-kinetic codes for microturbulence, neoclassical, RF-plasma interaction, neutral beam
particles, and neutral transport physics; and MHD/fluid codes for fast and large scale instabilities.
Such a simulation must be performed on leadership-class high performance computers (HPCs).
Due to the large scale data I/Os from large number of processors, the elevated fault tolerance
problem, and the heterogeneous computing methods on different grids, compilers, and numericallibraries between codes, there are numerous issues we need to understand and resolve in order toperform integrated simulations on leadership-class HPCs. To solve these problems, we havedeveloped the EFFIS (End-to-end Framework for Fusion Integrated Simulation) in service
oriented architecture. EFFIS utilizes a new adaptive I/O method, ADIOS, for simultaneous in-situ
coupling through memory-to-memory and files. DataSpace method will allow optimal in-HPC
coupling and in-situ data-compression and analysis. For the internal coupling workflow, an
extended ADIOS technology with DataSpace is used. For the external coupling workflow,
together with the data storage, the provenance information, the external analysis and thevisualization, the old Kepler technology is utilized, in cooperation with a fast data movement
technology. Web-based eSiMon dashboard is used for remote in-situ data analysis andvisualization on personal computers. New integrated scientific discoveries which were achieved
using the EFFIS technology on HPCs will be presented, which include the global core-edge full-f
gyrokinetic simulation, kinetic RMP penetration, and pedestal-buildup and ELM cycle. Utilization
the 10-100 peta flop HPCs, and beyond, will also be discussed.
Work supported by US Department of Energy, Office of Fusion Energy Science and Office of Advanced
Scientific Computing Research.

285

O-19
Theory of driven magnetic islands
F.L. Waelbroeck1), A. Ishizawa2), R.E. Waltz3), A.Y. Aydemir1), R. Fitzpatrick1)
1) Inst. Fusion Studies, Univ. Texas at Austin, TX 78712-0264, USA
2) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan3) General Atomics, P.O. Box 85608, San Diego, CA 92186-5608, USA
Electric currents flowing in coils located outside magnetically confined plasma give rise tomagnetic islands. Such externally driven islands are of growing importance as a means to control
the edge conditions in stellarators and tokamaks alike. They are experimentally observed to affectthe profiles of density, temperature, and electric field, as well as to modify the properties of the
turbulence in their vicinity. In many respects, these islands are analogous to airfoils in windtunnels. In particular, two parameters play a central role in the description of their interaction withthe plasma: the local drive, and the electromagnetic force acting on the plasma. These parameters
are loosely analogous to the lift and the drag acting on an airfoil. The central problem for thetheory of driven islands has been to determine the dependence of these two parameters on
conditions such as the strength of the external drive and the scale-lengths of the equilibrium fields
such as density, pressure, and plasma rotation velocity.
Several regimes can be distinguished according to the size of the island and the collision
frequency. The relevant dimensionless measure of the collision frequency is the ratio of the skindepth at the diamagnetic frequency and the ion Larmor radius. In large machines with a hot edge,
this parameter is small and the effects of diamagnetic drifts are important. The island is thenanalogous to a sailboat, with the electrons playing the role of the wind and the ions that of thewater. Models for the plasma response that assume quiescent conditions find that in all regimes,
the electrons inside the separatrix of the island are trapped or frozen into the island. This
circumstance, together with the continuity of the ion velocity, determines the electric field inside
the island in terms of the other plasma parameters. The density and temperature for electrons andions, by contrast, are determined by the competition between transport rates parallel to and acrossthe magnetic field. The presence of turbulence significantly alters the picture, giving rise to island
amplification by correlated eddies and a turbulent drag. Gyrokinetic simulations for large islands
show that the drag force reverses when the electric field is such that the ions outside the separatrix
are almost at rest with respect to the island, in contrast with small islands where the reversal
occurs when the outside electrons are at rest. Forcing the plasma to flow past the island gives rise
to significant steepening of the density and temperature profiles, an effect that is only observed formuch smaller islands in quiescent plasma. Lastly, sufficiently strong flows are found to unlock amagnetic island from the coils that generated it. The island will then heal as it rotates, leaving in
its place a much smaller, “suppressed” island.

286

I-18
Cross-cutting Ideas for a Fusion DEMO Plant with Current and Generation IV Nuclear
Power Plants
Hyuck Jong Kim, Myeun Kwon
National Fusion Research Institute, 113 Gwahangno, Yuseong-gu, Daejeon 305-806, Korea
Ever since the beginning of the ancient civilization, human beings have obtained energy fromnatural resources to enhance their welfare and sustain economic growth. To successfully realize
fusion energy that is to be the first energy not from natural resources but from technology, there
still are technology barriers to break through and the risks associated with investing to thetechnologies that have not been fully developed. The easiest way of mitigating the risks is to be
postponement of developing fusion energy until the technologies will be matured. However, as the
economic benefits that an early mover of fusion energy is to take are so huge [1], it is worthwhile
taking due risks associated with developing it [2].
A practical way of mitigating the risks is to be cross-cutting of fusion DEMO development making
use of the commonalities between a fusion DEMO systems and existing systems [3]. Among theexisting systems, current and generation IV nuclear power plants (NPPs) could have many areas ofcommonalities that could be utilized for the of development of fusion DEMO systems. The crosscutting ideas for the development of fusion DEMO systems, based on these commonalities, may
include the following areas:- Regulatory requirements and licensing processes;- Codes and standards;- Safety analysis methods;- Design methods;- Computational codes and Thermo-hydraulic test facilities;- Material development and irradiation test facilities;- Radioactive waste management and environmental protection;- Program lifecycle management skills.
References[1] D.J. WARD, I. COOK, Y. LECHON, R. SAEZ, The economic viability of fusion power,
Fusion Engineering and Design,75-79, 1221 (2005)[2] HJ. Kim, HC. Kim, Chul-Sik Lee, M. Kwon, and GS. Lee, “The strategic plans for thefusion DEMO program of Korea”, Proceedings of TOFE 19, Transactions of Fusion Scienceand Technology, vol. 60-2, pp 443-440 (2011)

287

I-19
Research platforms of laser and ion beam induced ablation plasmas
as interdisciplinary reactor studies
Kazuo A Tanaka, Masayoshi Nagata1), and Hitoki Yoneda2)
Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871 Japan
1)Graduate School of Engineering, University of Hyogo, Hyogo 671-2280 Japan
2)The University of Electro-Communications, Tokyo 182-8585, JapanThe extreme states of density ranging from vacuum to solid density, temperature eV or above are
created in both magnetic and laser fusion reactors due to disruption, ELM, explosions of fusion
reactions etc. These states can be studied using laser and ion beam induced ablation schemes as
experimental platforms. Considering this laser irradiation as a heat pulse source, the ablated
plasma plumes can be studied in various way.
Our recent study focuses on plasma shielding for material under extreme conditions. In
orthogonally oriented two targets are placed in a vacuum chamber where two laser pulses canirradiate. The pulse laser creates two plasma plumes crossing each other at the densities 1011 to
1013/c.c. at temperature at around eV [1, 2]. In case of Carbon targets are used, the crossing points
show stagnation up to microsecond, considerably longer than the laser pulse width 6 nsec. Further
processes include the molecular formation leading to the formation of Carbon Fullerene andCarbon nano tubes. These collision processes is observed to absorb the kinetic energy of incoming
plasma plumes up to 50 % at the laser energy density 10 J/cm2. This may be considered as aenergy absorption mechanism of incoming plasma flux when another plasma plume is prepared inadvance in front of the material to be protected.
Magnetized Coaxial Plasma Gun has been upgraded at the University of Hyogo. The plan is toreach 2MJ/m2 ion beam irradiation energy density that is expected in the Type-1 ELM at ITER.
The upgrade included two categories: (1) Increase of capacitance from 1mF to 2.9 mF and (2)Installation of tapered drift tube to increase the ion beam flux. The test operation has been
conducted to confirm the designed output energy density.
References[1] Y. HIROOKA et al., J. Phys. Conf. Ser. 244, 032033(2010); Y HIROOKA et al., Fus. Sci.
Tech., 60, 804(2011).
[2] KA Tanaka et al., Fus. Sci. Tech., 60,329 (2011).
Under the LHD Plan Collaboration Research Program: NIFS10KOAF001.

288

I-20
Particle and heat control for steady state burning plasma in helical reactor
S. Masuzaki, R. Sakamoto, M. Tokitani, H. Tamura, I. Murakami, D. Kato, H. Sakaue, T. Goto,
J. Miyazawa, T. Tanaka, M. Kobayashi, M. Tanaka, M. Shoji, T. Watanabe, T. Morisaki,
T. Muroga, A. Sagara
National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Particle and heat control is essential for steady state burning plasma. The control scenario in
helical reactor has been studied on the basis of knowledge obtained by the Large Helical Device(LHD) experiment. In this presentation, the scenario in the helical reactor, FFHR-d1, which isunder conceptual design, will be shown.
Fueling has to be well done, and helium ash and unreacted deuterium and tritium have to bepumped to sustain steady state burning plasma. In FFHR-d1, fuelling will be conducted mainlywith ice-pellet injection. The fuelling efficiency is estimated to be 0.18%, and necessary injection
amount of DT pair is 6x1023 DT/s for 3 GW fusion-power operation in which about 1.1x1021
reactions occur during 1s. The helical divertor which is intrinsically equipped in heliotron-typemagnetic configuration will be utilized for the compressing and pumping of residual fueled
particles and helium ash. The required pumping flux will be the order of 1000 Pa·m3/s.
Heat load to plasma facing components has to be less than that which causes serious damages of
the components especially divertor plates. In this presentation, the heat control scenario will be
discussed mainly from the point of view of plasma operation. One of the characteristics of thehelical divertor is non-uniform profiles of the particle and the heat flux on the divertor both in thetoroidal and poloidal directions. Studies of the particle and the heat loads on the helical divertor
have been conducted in LHD, and the understandings obtained from them can be applied to theestimation of the divertor heat load in FFHR-d1. To reduce the divertor heat load, impurity gaspuffing is planned. It can enhance radiation in the scrape-off layer, and reduce plasma temperature.

289