Confinement and Stability in the Large Helical Device
OV1-4 ConfinementConfinement andand MHDMHD StabilityStability inin thethe LargeLarge HelicalHelical DeviceDevice
Osamu Motojima LHD Experimental Group National Institute for Fusion Science
IAEA2004, Vilamoura, Portugal, Nov. 1, 2004 1 Content of my paper • 1. Change of NIFS Organization • 2. MHD study • 2.1. High beta experiment • 2.2. Healing of magnetic island • 3. Extended operation regime in LHD • 3.1 Edge control by Local Island Divertor • 3.2 Control of radial electric field by shift of magnetic axis • 3.3 Density limit and radiation collapse • 3.4 Achievement of high ion temperature with NBI • 3.5 Long pulse operation with ECH • 4 Transport study • 4.1 Properties of particle and heat transport • 4.2 Electron transport • 5 Summary
2 Big Challenge! National Institute for Fusion Science (NIFS) joined a new organization in April, 2004 “National Institutes of Natural Sciences (NINS)”
NIFS will be incorporated into a new academic agency, as an inter- university research institute for universitiesNational all Astronomical over Japan Observatory of Japan Together with Okazaki National Research Institutes (Institute for Molecular Science, National Institute for Basic Biology Okazaki National Research Institute and National Institute for Physiological Sciences) • Institute for Molecular Science and National Astronomical Observatory of Japan National Institute • National Institute for Basic Biology for Fusion Science • National Institute for Physiological Sciences Fusion, Astronomy, Materials, Biology, etc.
• Increase the activities of mutual collaboration and exchange program with domestic and international universities and institutions to develop individual research areas and new science categories • Provide adequate research opportunities for scientists • Work with Universities to educate graduate students Work with Universities to educate graduate students 3 Objectives of NIFS/LHD Project National Institute for Fusion Science • Established in May, 1989 An inter-university National Institute to promote scientific research of fusion plasmas and their application Report of National Council for Science and Technology in 1984 • NIFS promotes experimental and theoretical research into fusion and plasma physics using the world’s largest superconducting helical experiment, the Large Helical Device (LHD) and by means of theoretical and simulation studies Based on the Heliotron, an original Japanese concept • Increased effort in fusion technology
4 Ground Design of Japanese Fusion Research Now, role of NIFS is clear . Keep a close relation with universities . Increase collaborationStratifiedas a Structurecenter of excellenceof Research of fusion research towards Fusion Reactor . Increase educational function in cooperation with the Graduate University Development for Advanced Studies Fusion Reactor
ITER
IFMIF Tokamak LHD Laser (Development) (Science)
Reactor engineering Fusion science Plasma Science Scientific Research Basis Science
Working Group on Fusion Research Future Direction of National Fusion Research Special Committee on Basic Issues Subdivision on Science Council for Science and Technology
5 External diameter 13.5 m Present View! Plasma major radius 3.9 m Plasma minor radius 0.6 m Large Helical Device Plasma volume 30 m3 (LHD) Magnetic field 3 T Total weight 1,500 t
ECR Plasma vacuum vessel 84 – 168 GHz
World largest superconducting coil system Magnetic energy 1 GJ Cryogenic mass (-269 degree C) 850 t Local Island Tolerance < 2mm Divertor (LID)
ICRF 25-100 MHz NBI NBI
6 MissionMission ofof LHDLHD Physics Issues Contributing to Fusion Research τ (a) To realize high n ET plasmas and to study transport physics applicable to fusion plasmas, (b) To demonstrate high β stable plasmas (<β> ≥ 5 %) and to study related physics, (c) To develop physics and technology for long pulse or steady state operation and control using divertor, (d) To study energetic particle behaviors to simulate a particles in fusion plasmas, (e) To increase the physics understanding of toroidal plasmas by an approach which is complementary to tokamaks
7 Steady progress of plasma parameters
Electron temp. NBI 10 12 Ion temp. ICH 800 ECH 5 10 700 Pulse length (s) length Pulse Heating 600
Power(MW) 0 8 500 6 400 1.0 4 300 Pulse length 200 2 Temperature (keV) 0.5 1st(1998) 5th(2001) 100 2nd(1998) 6th(2002) '98 '99 '00 '01 '02 '03 0 0 Stored Energy (MJ) Energy Stored 3rd (1999) 7th(2003) 1234567 4th(2000) 0 Experimental campaign 0 10000 20000 30000 40000 50000 Shot Number
Stored energy has reached 1.3MJ comparable to big tokamaks. Electron temperature 10 keV Ion temperature 10 keV Beta 4.1 % Density 2.4x1020 m-3 Pulse length 756 s 8 Beta value has reached 4.1%
LHD has been pushing beta 9 γ . κε Aspect-ratio ( =n/m ac/R= ) Control in LHD
• Aspect ratio of plasma is optimized by controlling the central position of HC current
- Aspect-ratio of 5.8 ~ 8.3 is available in the Rax = 3.6 m P oloidal Coils configuration - High aspect-ratio configuration has high rotational transform which restrains the plasma shift
HC-O HC-MH- M HC-I
Helical Coil Plasma Large Aspect Ratio 10 Scenario to achieve high β plasma in LHD • γ optimization to minimize Shafranov shift is a key
β(%) RT_NBI Transport 0.4 improvement γ =1.254 MHD limit 0.3 γ =1.22 R 0.2 T_NBI
0.1 transport limit shafranov_shift [m] Small Shafranov 0.0 (γ 01234 shift =1.22)
11 High beta discharge at B=0.45T
Beta value has reached 4.1%
. The core fluctuation (ex.n/m=1/2) disappears because of spontaneous generation of magnetic well
. Even edge fluctuation (n/m=1/1) is mitigated because of flattening of pressure gradient
Further progress expected
12 K.Y.Watanabe EX/3-3 Study on MHD stability limit of high beta plasma Role and Function of Boundary
currentless m/n=1/1 unstable ρ=0.9 (ι~1) Achieved beta is close γ/ω =10-2 12 dβ /dρ A kin to the m/n=1/1 ideal 0.5x10-2 mode limit
8 Achieved beta Mercier significantly exceeds the Mercier limit 4
γ=1.22 0 γ . κε 01234 ( =n/m ac/R= ) T (T ): Thomson <β > (%) e i dia ne:FIR . β values achieved significantly exceeds the Mercier limit and increases up to m/n=1/1 ideal MHD limit
kinetic beta gradients at ρ=0.9 (ι/2π = 1) in <β>-dβ/dρ diagram. 13 K.Y.Watanabe EX/3-3 Confinement property of high β regime 2.0 (a) 1.5
1.0 iss95-Diamag
H 0.5 low β high β 0.0 01234 <β > (%) dia
. H-factor with respect to the ISS95 scaling decreases as the β is increased . This decrease is due to the possible reduction of ISS95 scaling at higher collisionality . Degradation of energy confinement due to high β is not observed
The improvement factor of effective energy confinement as a function of K.Y.Watanabe14 EX/3-3 (a) beta value for the plasma with R =3.6m, B=0.45-1.75T and γ = 1.22 and ax B.J.Peterson EX/6-2 Healing of magnetic island . Magnetic island appearing after the pellet injection is healed by itself as the collisionality decreases . Positive shear(ι/2π) plays an important role on MHD property of LHD
2 (a) Shot 38937 (b) Island Width (β=0.77% plasma) 1.35 healing of 0.2 1.5 magnetic Vacuum Width
1.25 island W(m) (keV) 1 e 0.1 T 1.15 0.5 1.05ms (after pellet) 0 0 00.10.2 4 4.2 4.4 4.6 W (m) R (m) ex high ν * Low ν * b b . Magnetic island existing in the Magnetic island Magnetic island vacuum field is healed, when the size is growing is healed of magnetic island is small (w<0.3a)
(a) Time evolution of the Te profile with the n=1 external field. 15 (b) Normalized coil current vs. island width (w) in vacuum (open circles) and in plasma (closed circles). Y.Nagayama EX/P5-14 Extended operation regime in LHD •• Edge control by Local island divertor (LID) •• Control of radial electric field by shift of magnetic axis •• Density limit and radiation collapse •• Achievement of high ion temperature with NBI •• Long pulse operation with ECH
Transport study •• Properties of particle and heat transport •• Electron transport
16 Energy Confinement and Targets of LHD
τ A factor of 1.5 improvement of the energy confinement time E over the ISS95 Comparable to ITER ELMy H-mode International Stellarator Scaling 95 (ISS95) τISS 95 = 0.26B0.80 P −0.59n 0.51R0.65a 2.21q −0.40 E e 2/3 102 Ignition Commercial Reactor ∝ τ ρ−0.71β−0.16ν −0.04 B * * JT-60U ITER Q=1 101 JET W7-AS JT-60U LHD 2nd target ATF DIII-D 100 JET 0 CHS LHD 1st target 10 H-E keV s) LHD W7-A -3 10-1 m LHD R3.75 ToreSupra 20 LHD R3.6 -2 -1 10 (10
10 E τ Tokamaks i0 (s) -3 T 10 0 exp Fusion triple product product triple Fusion n E -2 Preceding TRIAM-1M τ 10 10-4 Helical Dev. min. hour day month Tokamaks 10-5 ELMy H-mode 10-1 100 101 102 103 104 105 106 107 10-3 Operated Duration (s) 10-3 10-2 10-1 100 τ ISS95 (s) 17 E H.Yamada EX/1-5 4 LID configuration 0.08 (a) Create sharp 3 LID LID edge (large Te config. gradient) 2 Confinement (s) (keV) exp e improvement E T τ 1 Helical divertor config. 0 0 3.8 4.0 4.2 4.4 4.6 00.08τ ISS95 (s) 20 R (m) 6 E MI (c) without MI (b) with MI 5 Produce edge 10 X-point Prevent (limiter config.) Positive E (LID config.) 4 r radiation with MI 0 collapse 3 (LID config.) (kV/m) r
Prad(MW) 2 E -10 by impurity without MI 1 shielding effect -20 (limiter config.) 0 3.8 3.9 4.0 4.1 4.2 0.0 0.5 1.0 1.5 2.0 R(m) Time(s) . Basic function of LID demonstrated 1) confinement improvement 2) prevent radiation collapse 18 A.Komori EX/10-4, K.Ida EX/P4-6 Study on high ion temperature • Central ion temperature increases up to 10keV with strong impurity puff
12 12 (a) (b) 10 Te (ECE) 10 T i R=3.822m 8 8 (ArXVII) (keV) 6 e
(keV) 6 i T
4 & T i 4
2 T 2 0 02468101214 0 19 -3 01234 Pabs/ne (MW/10 m ) Time (sec)
at present: 3 tangential beam lines with 180keV/14MW H: electron heating Ar, Ne: ion heating next year: one perpendicular beam line added with 40kev/3MW more efficient central ion heating
(a) Ion temperature as a function of the direct ion heating power normalized by the ion density in the plasma with Ar- and Ne-puff and 19 (b) time evolution of electron and ion temperature in a low-density high-Z plasma. Y.Takeiri EX/P4-11 Particle transport in L-mode plasmas Density profile is flat in the core region in LHD Transient transport analysis with gas puff modulation is required to derive diffusion coefficient in the core Peaked ne Hollow ne 1 8 (a) (b) 4 V D core
/sec) edge Outward 2 0 0.1 V(m/sec) D(m -4 D core V -8 edge Inward 0.5 1 110 T (keV) -dT /dr (keV/m) low P e high P e nbi nbi low P high P 1.7+-0.9 nbi nbi Dcore ~ Te 1.1+-0.14 Dedge ~ Te . Consistent with Gyro Bohm . Consistent with density profile measured in the steady state . 1.5 Te dependence (Te ) 20 (a) Electron temperature dependence of diffusion coefficient (b) convective velocity as a function of the temperature gradient K.Tanaka EX/P6-28 Heat transport in ITB plasma χ Steady-state transport analysis Power balance e χ χ Transient transport analysis Cold pulse propagation e and d e/dT
15 (b) /s) (a) Power 2 LHD LHD balance 0 (m e 10 ITB χ χ e α ) β 5 -5 = (1+
cp 00 0.2 0.4 0.6 0.8 1 χ 0 0.2 0.4 0.6 0.8 1 ρ ρ Temperature dependence . α χ χ Significant reduction of thermal = (Te/ e)(d e/dTe) diffusivity inside the ITB is observed Inside ITB : α < 0 negative T dependence both in steady-state and transient e Outside ITB : α > 0 positive T dependence transport analysis e (a) The radial profiles of the electron heat diffusivity χ α (b) Te dependence factor of e, , estimated by cold pulse propagation. The heat diffusivity estimated by power balance is also plotted 21 S.Inagaki EX/P2-12, T.Shimozuma EX/P3-12 Obtained Physics and Achieved Parameters of LHD Experiments LHD 1. Plasma performance was improved remarkably through 7 experimental campaigns in these 6 years 2. Quality and amount of database increased remarkably for MHD and transport study
3. With NBI of 12MW, Te=4.5keV and Ti=10.1keV were × 18 -3 obtained at
Joint Report of EU/JA Expert Group Meeting 18th / 19th April 2004, Culham on A Broader Approach to Fusion Power ITER/DEMO oriented
Strong accompanying physics programmes are needed in the parties during ITER construction and operation. Their functions should include, in particular, to directly support ITER and to complement ITER outputs in the preparation of DEMO.
The main functions in support to DEMO will be to explore operational regimes and issues complementary to those being addressed in ITER. In particular these will include:
23 Contributor O.Motojima 1), K.Ida 1), K.Y.Watanabe 1), Y.Nagayama 1), A.Komori 1), T.Morisaki 1), B.J.Peterson 1), Y.Takeiri 1), K.Ohkubo 1), K.Tanaka 1), T.Shimozuma 1), S.Inagaki 1), T.Kobuchi 1), S.Sakakibara 1), J.Miyazawa 1), N.Ohyabu 1), K.Narihara 1), K.Nishimura 1), M.Yoshinuma 1), S.Morita 1), T.Akiyama 1), N.Ashikawa 1), C.D.Beidler 2), M.Emoto 1), T.Fujita 3), T.Fukuda 4), H.Funaba 1), P.Goncharov 5), M.Goto 1), H.Idei 6), T.Ido 1), K.Ikeda 1), A.Isayama 3), M.Isobe 1), H.Igami 1), K.Itoh 1), O.Kaneko 1), K.Kawahata 1), H.Kawazome 7), S.Kubo 1), R.Kumazawa 1), S.Masuzaki 1), K.Matsuoka 1), T.Minami 1), S.Murakami 8), S.Muto 1), T.Mutoh 1), Y.Nakamura 1), H.Nakanishi 1), Y.Narushima 1), M.Nishiura 1), A.Nishizawa 1), N.Noda 1), T.Notake 9), H.Nozato 10), S.Ohdachi 1), Y.Oka 1), S.Okajima 11), M.Osakabe 1), T.Ozaki 1), A.Sagara 1), T.Saida 12), K.Saito 1), R.Sakamoto 1), Y.Sakamoto 3), M.Sasao 12), K.Sato 1), M.Sato 1), T.Seki 1), M.Shoji 1), S.Sudo 1), N.Takeuchi 9), N.Tamura 1), K.Toi 1), T.Tokuzawa 1), Y.Torii 9), K.Tsumori 1), T.Uda 1), A.Wakasa 13), T.Watari 1),H.Yamada 1),I.Yamada 1), S.Yamamoto 9), T.Yamamoto 9), K.Yamazaki 1), M.Yokoyama 1), Y.Yoshimura 1)
1) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, 509-5292, Japan 2) Max-Planck Institut fuer Plasmaphysik, Greifswald D-17491, Germany 3) Japan Atomic Energy Research Institute, Naka, 311-0193, Japan 4) Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871, 5) Department of Fusion Science, School of Mathematical and Physical Science, Graduate University for Advanced Studies, Hayama, 240-0193, Japan 6) Research Institute for Applied Mechanics, Kyushu University, Kasuga, 816-8580, Japan 7) Graduate School of Energy Science, Kyoto University, Uji 611-0011, Japan 8) Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan 9) Department of Energy Engineering and Science, Nagoya University, 464-8603, Japan 10) Graduate School of Frontier Sciences, The University of Tokyo 113-0033, Japan 11) Chubu University, Kasugai, Aichi, 487-8501, Japan 12) Graduate School of Engineering, Tohoku University, Sendai, 980-8579, Japan 13) Graduate School of Engineering, Hokkaido University, Sapporo 060-8628, Japan e-mail contact of main author: [email protected] 24