ITC28The 28th International Toki Conference on Plasma and Fusion ResearchCeratopia Toki, Toki-city, Gifu, Japan, November 5-8, 2019

Book of Abstracts

Printed on October 23, 2019

Nov. 5, 2019 (Tue) Nov. 6, 2019 (Wed) Nov. 7, 2019 (Thu) Nov. 8, 2019 (Fri)

Opening

(Group Photo)

PL1: K. Kurihara

I1-1: G. McKee

I1-2: D. Kato

16:10

18:20

18:50

Magnetically-Confined PlasmasInertially-Confined PlasmasFusion Engineering and Reactor DesignBasic Plasma Research and Plasma ApplicationFundamental theory and simulation techniqueSpace and Astrophysical PlasmasApplied Superconductivity and Cryogenic Systems

20:30

ITC28 Program

I1-4: K. Ibano

O1-2: T. Bando

O1-3: M.S. Islam

O1-4: H. Yamaguchi

17:00Break

17:20

17:40

18:00

9:30

10:20

11:10

11:40

12:10

12:40

13:40

15:40

16:00

9:30

12:30

I2-1: N. Yanagi

I2-3: M. Wolf

I2-4: M. Tomita I4-4: S. Saito

I4-2: S. Lazerson

10:20

PL2: W. Stautner

I2-2: N. Amemiya

Lunch

I3-2: H. Utoh

O3-1: T. Kitasaka

I3-3: A. Houben

I3-1: S.N. Jiang

I3-4: G.N. Luo

11:40

12:0012:10

10:20

11:10

PL4: M. OsakabePL3: R.J. Kurtz

9:30

11:10

11:40

I4-1: N. Ezumi

I4-3: M. Tanaka

15:30

14:50

16:00

12:40

O4-3: M. van Berkel

O4-4: H. Wang

Closing

14:10

14:30

15:10

I1-3: Y. Mori

Excursion

16:30

15:40

13:40

O1-1: K. Itoh

Poster 1

18:30

13:00

16:20

16:50

17:30

Lunch

Poster 2

O2-1: W. Chen

O2-2: M. Bakr

O2-4: T. Murase18:30

O2-3: A.V.B Catapang

I2-6: T. Nozaki

17:10

I2-7: A. Shimizu18:00

Banquet

I2-5: T. Watanabe

10:40

11:30

12:00

Break

9:30

10:20Break

10:50Break

10:50Break

10:50Break

12:30

Lunch

13:40

Category

Break

15:50

O4-1: H. Hachikubo

I4-5: H. Yamaguchi

O4-2: N. Tsujii

List of Exhibitors (in alphabetical order)

Fujikura Ltd. https://www.fujikura.com/

Furukawa Electric Co., Ltd. https://www.furukawa.co.jp/

Hitachi, Ltd. http://www.hitachi.com/

Mitsubishi Electric Corporation, Chubu branchhttps://www.mitsubishielectric.com/

Sumitomo Electric Industries, Ltd. https://global-sei.com/

SuperOx Japan LLC http://www.superox.jp/

Toshiba Energy Systems & Solutions Corporation https://www.toshiba-energy.com/

Day1 - November 5, 2019 (Tue)

09:30-10:20 Opening

10:20-10:40 Break

10:40-11:30 PL1: Kenichi KURIHARAThe Current Project Status of the ITER and the Nearly-Assembled JT-60SA

11:30-12:00 I1-01: George Raymond MCKEEImpact of Fuel Ion Isotope Mass on Confinement, Turbulence and the L-H Transition in DIII-D

12:00-12:30 I1-02: Daiji KATOVisible M1 Line Emission of Highly Charged Tungsten Ions in LHD Core Plasmas

12:30-13:40 Lunch

13:40-15:40 Poster 1

15:40-16:00 O1-01: Kimitaka ITOHMulti-ferroic turbulence in non-equilibrium plasmas

16:00-16:30 I1-03: Yoshitaka MORIActivities of Inertial Fusion Energy program at GPI Hamamatsu

16:30-17:00 I1-04: Kenzo IBANOVapor Shielding Simulation by Weighted PIC Code for Erosion Estimation during Transient Loads

17:00-17:20 Break

17:20-17:40 O1-02: Takahiro BANDOExperimental observations of n = 1 helical cores accompanied with saturated m/n = 2/1 tearing modeshaving low mode frequencies in JT-60U

17:40-18:00 O1-03: Md Shahinul ISLAMStudy of the Transient Behavior of Detached Plasma during Xe Gas Injection into the D-module of GAMMA 10/PDX

18:00-18:20 O1-04: Hiroyuki YAMAGUCHIStudy of magnetic configuration based on helical coil system and attempt for numerical optimization

The Current Project Status of the ITER and the

Nearly-Assembled JT-60SA

Kenichi KURIHARA

1. National Institutes for Quantum and Radiological Science and Technology

The international collaboration project to build the fusion experimental reactorITER has been aggressively progressing in construction of the buildings at theITER site as well as in manufacturing of the components such as large supercon-ducting (SC) coils, vacuum vessel (VV) sectors, etc., through in-kind contribu-tions by 7 parties. As of September, 2019, the achievement of the project towardthe first plasma in 2025 has come to be approximately 65 percents. Japan is incharge of procurement of the major state-of-the-art components as an associatehost for the ITER project. For the major tokamak components, in particular,the first SC toroidal field coil (TFC) is expected to be completed in this comingDecember, and will be transferred from Japan to the ITER site as a Japanesein-kind procurement. The first and second TFCs will be sub-assembled with thefirst 40-deg VV sectors at the tokamak assembly building. The current status ofother major components procured by Japan is as follows: (a) The neutral beaminjector test facility is also being promoted in Italy. The 1-MV DC power supplycontributed by Japan is now tested with a 1.2-MV high voltage. (b) Gyrotronsfor radio-frequency heating are also steadily tested. Factory acceptance tests for 2out of 8 have been completed. (c) In order to determine the detailed design of theblanket remote handling system, discussion between Japanese domestic agencyand ITER organization is intensively carried out. (d) The tungsten divertor,the detritiation system of the tokamak complex, and the diagnostics system arebeing developed toward finalizing their manufacturing designs and interface con-ditions. For the satellite tokamak JT-60SA project under the Japan-EU BroaderApproach activities, the assembly work has been progressing on schedule. Themajor objectives of JT-60SA are (a) to support ITER-related plasma studies inadvance for optimization of its operation scenarios, and (b) to study a high pres-sure plasma for a compact DEMO. As of September, 2019, the cryostat vesselbody has almost covered the tokamak core, and the assembly will be accom-plished in March, 2020. Other components necessary for the first plasma havebeen installed and tested in parallel to the tokamak assembly: The power sup-plies for the SC toroidal and equilibrium field coils had been newly developedand manufactured for JT-60SA, together with the quench protection circuits. Inorder to suppress instabilities in a high pressure plasma, in-vessel coils will beinstalled along the wall surface of the vacuum vessel. The progressing projectstatus of the ITER and the nearly-assembled JT-60SA will be presented in detailwith the updated pictures and movies.

PL1

Impact of Fuel Ion Isotope Mass on Confinement,

Turbulence and the L-H Transition in DIII-D

George Raymond MCKEE1, Zheng YAN1,

Punit GOHIL2, Terry L. RHODES3, Lothar SCHMITZ3

1. University of Wisconsin-Madison, Madison, Wisconsin, USA2. General Atomics, San Diego, California, USA3. University of California-Los Angeles, Los Angeles, California, USA

The dependence of turbulence characteristics, transport scaling and the L-H tran-sition power threshold on the fuel ion mass has been investigated in hydrogen(A=1) and deuterium (A=2) L-mode plasmas on DIII-D. Normalized energyconfinement time (B ∗ τE) is approximately two times higher in deuterium (D)plasmas compared to similar hydrogen (H) plasmas, while the L-H transitionpower threshold decreases in Deuterium in a density-dependent manner. Dimen-sionless parameters other than ion mass (A), including ρ∗, q95, Te/Ti, β, ν

∗, andMach number, were maintained nearly fixed during the transport comparison.The normalized turbulence amplitude (n/n) is approximately twice as large inH as in D, which may partially explain the increased transport and reduced en-ergy confinement time. Radial correlation lengths of low-wavenumber densityturbulence in hydrogen are similar to or slightly larger than correlation lengthsin the deuterium plasmas and generally scale with the ion gyroradius, which wasmaintained nearly fixed in this dimensionless scan.

PLH is higher in Hydrogen than in Deuterium by nearly a factor of two nearthe density minimum in PLH , but the difference decreases and converges at higherdensity, and diverges further at lower densities. Measurements of long wavelengthdensity fluctuation characteristics in the edge across the L-H transition demon-strate the existence of single or double modes of turbulence, which are stronglycorrelated with the L to H-mode transition power threshold (PLH) and may ex-plain the isotopic and density dependence of PLH , and how the PLH difference isreduced at higher density. The increased edge fluctuations, increased flow shear,and the dual-band nature of edge turbulence correlating with lower PLH in Deu-terium suggests a complex behavior that can inform a more complete model ofthe L-H transition threshold.

Predicting the energy confinement and the L-H power threshold in D-T burn-ing plasmas requires an understanding and accurate prediction of the large andbeneficial isotope scaling of turbulence behavior, transport and PLH .

*Supported by US DOE under DE-FG02-08ER54999, DE-FC02-04ER54698,DE-FG02-08ER54984.

I1-01

Visible M1 Line Emission of Highly Charged

Tungsten Ions in LHD Core Plasmas

Daiji KATO1,2, Shota ERA2, Hiroyuki A. SAKAUE1,

Izumi MURAKAMI1,3, Motoshi GOTO1,3,Tetsutaro OISHI1,3, Shigeru MORITA1,3, Keisuke FUJII4,

Nobuyuki NAKAMURA5

1. National Institute for Fusion Science2. Kyushu University3. SOKENDAI4. Kyoto University5. The University of Electro-Communications

Forbidden transitions in ground term fine-structures of highly charged heavy ionsfall into the visible range. Such the transition well known is the magnetic-dipole(M1) transition of Fe13+ in the solar corona that emits a green line at 530.3 nmobserved in eclipse. Doschek and Feldman [1] discussed possibilities of observingsuch forbidden lines in tokamak plasmas and pointed out potential usefulness forplasma diagnostics. Visible lines are particularly useful for fusion plasmas be-cause neutron radiation damage on detectors can be avoided by usage of opticalfibers. Since tungsten is selected as divertor materials for the next fusion de-vice, ITER (international thermonuclear experimental reactor), visible M1 linesof highly charged tungsten ions will be useful for ITER, and future DEMO. Thevisible M1 line emission of highly charged tungsten ions in high-temperaturemagnetically confined plasmas was observed for the first time by tungsten pelletinjection experiments at Large Helical Device (LHD) [2]. We will present obser-vation of near ultra-violet and visible M1 lines of highly charged tungsten ionswith 4f open-shell at LHD which are identified by comparing with wavelengthsmeasured by the compact electron beam ion trap (CoBIT) [3] as well as withatomic structure calculations. We will also present results of its applications totungsten measurements at LHD [4,5].

[1] G.A. Doschek and U. Feldman, J. Appl. Phys.47 (1976) 3083.[2] D. Kato et al., Phys. Scr. T156 (2013) 014081.[3] N. Nakamura, H. Kikuchi, H. A. Sakaue et al., Rev. Sci. Instrum. 79,

063104 (2008).[4] D. Kato et al., 26th IAEA Fusion Energy Conference (17-22 oct., 2016,

Kyoto, Japan), EX/P8-14.[5] K. Fujii, D. Kato et al., J. Phys. B: At. Mol. Opt. Phys. 50, 055004

(2017).

I1-02

Multi-ferroic turbulence in non-equilibrium

plasmas

Kimitaka ITOH1,2,3, Yusuke Kosuga2,4,Tatsuya Kobayashi3, Sanae-I. Itoh2,4

1. Institute of Science and Technology Research, Chubu University2. Center for Plasma Turbulence, Kyushu University3. National Institute for Fusion Science, National Institutes of Natural Sci-

ences4. Research Institute for Applied Mechanics, Kyushu University

Plasmas in laboratories and in nature are often associated with multiple originsthat deviate the plasma far from thermodynamical equilibrium, e.g., gradients ofscalar fields and vector fields, etc. Under such circumstances, turbulence-inducedfluxes (the particle and heat fluxes, momentum fluxes, vorticity fluxes, those ofvector potential, etc.) can mutual interference so as to generate new types of‘ turbulent-structure’. Such states of turbulence is called multiferroic turbulence[1]. (In the concept of‘ interference’, such phenomenon that the relaxation of ve-locity gradient induces an up-hill particle flux, and/or the double helix flow withopposite chirality is generated, etc.) In addition to various (spatial) symmetry-breaking (including that of turbulence [2], the time-reversal symmetry is alsobroken due to sources. We here extend the concept of multiferroic turbulence toinclude the effects of sources, and discuss such an extension will give us possi-bility to solve long-standing mysteries (e.g., effect of wall material on improvedconfinement, etc.) Importance of the study of higher order correlation functionsis also addressed [3]. This presentation is a temporary review of work, which hasbeen developed and performed by the support of JSPS Grant-in-Aid of ScientificResearch (A) 15H02155.

[1] S.-I. Itoh et al., Japan Physical Society Autumn Meeting (Kanasai Univ.2015 Sep.)‘ Cross-ferroic turbulence ’, 17aCN-10.

[2] K. Itoh, et al.: Symmetry-Breaking of Turbulence Structure and PositionIdentification in Toroidal Plasmas, Plasma and Fusion Research - Review Articles,13 (2018) 1102113

[3] Sanae-I. ITOH, et al.: Plasma and Fusion Research: Review Articles Vol-ume 12, 1101003 (2017)

O1-01

Activities of Inertial Fusion Energy program at

GPI Hamamatsu

Y. Mori, R. Hanayama, S. Okihara, K. Ishii,Y. Kitagawa, A. Sunahara, Y. Sentoku, E. Miura,

A. Iwamoto, H. Sakagami, T. Johzaki

1. The Graduated School for the Creation of New Photonics Industries2. Purdue University CMUXE3. ILE, Osaka University4. National Institute of Advanced Industrial Science and Technology5. National Institute for Fusion Science6. Hiroshima University

The Inertial Fusion Energy (IFE) reactor relies on pulsive operation that creatingan instantaneous high-energy density state in a reactor chamber to explode fusionreaction energy and repeating this explosion. Diode-Pumped-Solid-State Laser(DPSSL) is a leading candidate for IFE driver. The Graduated School for theCreation of New Photonics Industries (GPI) have conducted IFE program [1]beyond 10 years with collaboration Hamamatsu Photonics K. K., and ToyotaMotor Corporation by utilizing 10 Joule class DPSSL KURE-1 [2] or DPSSL-assist laser HAMA [3] those are installed in Hamamatsu Photonics industrialdevelopment laboratory in Hamamatsu. The original mission of our program wasresearch and development of IFE key technologies scalable toward future IFEplant such as pellet injection systems [4, 5]. Our activities also have contributedverifications of physics related to fast ignition scheme [6-9]. In the program, wepropose a mini-reactor CANDY [10] that driven by a kJ-class repetitive laserdriver based on DPSSL for an engineering feasibility study of the power plant inthe counter beam fast ignition scheme fusion. Toward the CANDY, feasibilitystudies on pellet injection systems both beads pellet and spherical shell pelletare on-going [11, 12]. In this talk, we present progresses of IFE program in GPI,Hamamatsu.

[1] Y. Kitagawa et al., Plasma Fusion Res. 8, 3404047 (2011). [2] T. Sekineet al., Opt. Express 21, 8393 (2013). [3] Y. Mori et al., Nucl. Fusion 49, 075006(2013). [4] O. Komeda et al., Sci. Reports 3, 2561 (2013). [5] Y. Nishimura et al.,J. Plasma Fusion Res. 91, 544 (2015). [6] Y. Kitagawa et al., Phys. Rev. Lett.108, 155001 (2012). [7] Y. Mori et al., Phys. Rev. Lett. 117, 055001 (2016).[8] Y. Mori et al., Nucl. Fusion 57, 116031 (2017). [9] Y. Kitagawa et al., Phys.Rev. Lett. 114, 195002 (2015). [10] Y. Kitagawa et al., Plasma Fusion Res. 8,3404047 (2013). [11] Y. Mori et al., Fusion Sci. and Technol. 75, 36 (2019). [12]Y. Mori et al., Nucl. Fusion 59, 096022 (2019).

I1-03

Vapor Shielding Simulation by Weighted PIC

Code for Erosion Estimation during Transient

Loads

Kenzo IBANO1, Yoshio UEDA1, Tomonori TAKIZUKA1

1. Graduate School of Engineering, Osaka University

A large concern for ITER and future fusion devices is erosion of wall materialsduring transient events such as ELM and disruptions. Melting, vaporization andablation of wall materials will be caused by the extensive heat loads. On theother hand, an inherent protection mechanism for the wall erosion is known andcalled as vapor shielding. Vapor emitted from the surface interacts with theplasma incoming to the wall. This vapor-plasma interaction can dissipate theincoming plasma energy. We have developed a particle-in-cell (PIC) simulationcode, called PIXY [1, 2], and applied to the vapor shielding at fusion devices.One of challenges in the vapor shielding simulation by a PIC method is to treatsurface ejected particles. The amount of ejected flux spreads over a very widerange as a function of the surface temperature. In order to simulate this fluxwithin a calculatable numbers of numerical super-particles, a weighted PIC modelis adopted in PIXY. In the weighted PIC model, each super-particle has differentweight (number of actual particles in a super-particle). Then, the smaller/highervapor fluxes can be treated by super-particles with smaller/larger weights. Usingthe PIXY, we have been investigating the wall erosions under various transientevents. In this paper, we summarize our findings on vapor shielding efficiencies fordifferent wall materials, incoming plasma energy, pulse width, and pulse shape.Based on these findings, predictions of wall lifetime in the future fusion deviceswill be shown and discussed.

[1] K. Ibano, S. Togo, T.L. Lang, Y. Ogawa, H.T. Lee, Y. Ueda, and T.Takizuka, Contrib. Plasma Phys. 56 (2016) 705

[2] K. Ibano Y. Kikuchi, A. Tanaka, S. Togo, Y. Ueda, and T. Takizuka, Nucl.Fusion, 59 (2019) 076001

I1-04

Experimental observations of n = 1 helical cores

accompanied with saturated m/n = 2/1 tearing

modes having low mode frequencies in JT-60U

Takahiro BANDO1, Go MATSUNAGA1,Manabu TAKECHI1, Akihiko ISAYAMA1,

Naoyuki OYAMA1, Sizuo INOUE1, Maiko YOSHIDA1,Takuma WAKATSUKI1

1. National Institutes for Quantum and Radiological Science and Technology,Naka, Ibaraki 311-0193, Japan

Helical cores (HCs) in plasmas are one of the important phenomena becausethe helical structure affects impurity transport and induces loss of confined ener-getic particles or loss of toroidal momentum. Substantial studies on experimentalobservation of HCs were reported, because HCs would be excited in hybrid sce-narios in ITER or DEMO. In addition to the effect on transport by HCs, recentnumerical studies suggested that MagnetoHydroDynamics (MHD) dynamo ac-companied with HCs redistributes the current profile in the core and realizesthe experimentally observed sawtooth-free plasmas of hybrid scenarios in DIII-D.Though substantial experimental observations were reported, the experimentalobservations focusing on the detailed relationship between the mode structuresof the HC and other MHD modes have not been reported well.

In this paper, we report experimental finding of n = 1 HCs accompanied withsaturated m/n = 2/1 Tearing Modes (TMs) having low mode frequencies in JT-60U. The TMs with HCs are observed after an increase of the mode amplitudeand a decrease of the mode frequency of m/n = 2/1 precursors having the tearingparity. The decreased mode frequency is lower than 20 Hz typically. With variousdiagnostics, the coupling of n = 1 helical cores and m/n = 2/1 TMs have beenclearly observed. Because the coherent oscillations in the ion temperature areobserved in the core region and in the edge region, the flux surfaces including them/n = 2/1 magnetic island appear to have m = 1 helical deformation. It hasalso been suggested that the m/n = 2/1 TM and the HC rotate in the electrondiamagnetic direction keeping fm/n=1/1(HC) = 2fm/n=2/1(TM) in several plasmas.Here, fm/n=1/1(HC) is the mode frequencies of HCs and fm/n=2/1(TM) is the modefrequency of TMs. In addition, the core seems to be shifted to the high field sidewhen the O-points of the m/n = 2/1 magnetic island line up in the midplane,which is confirmed by the reconstruction of MHD equilibria with Motional StarkEffect measurement and MEUDAS code. Our observations on m/n = 2/1 TMshaving HCs would contribute to the understanding of the excitation mechanismof HCs in tokamak plasmas.

O1-02

Study of the Transient Behavior of Detached

Plasma during Xe Gas Injection into the

D-module of GAMMA 10/PDX

M.S. Islam1, Y. Nakashima2, T. Iijima2, T. Hara2, K. Nojiri2,A. Terakado2, N. Ezumi2, M. Yoshikawa2, T. Kariya2, R. Minami2,K. Hoshino3, A. Hatayama3, H. Hasegawa1, S. Ishiguro1, M. Sakamoto2

1. National Institute for Fusion Science, Toki, Gifu, Japan2. Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki, Japan3. Faculty of Science and Technology, Keio University, Hiyoshi, Japan

A divertor simulation experimental module (D-module) has been installed at thewest end cell of the device to explore the physics of plasma detachment and energyloss processes [1-2]. A V-shaped target has been installed in the D-module anda set of Langmuir probes and calorimeters are installed on the target plate tomeasure the plasma parameters [1-2]. Plasma detachment state has been shownby injecting impurity gases into the D-module [1].

Transition phenomenon of plasma detachment during additional heating ef-fect (ECH) has been recently studied by applying a short pulse of ECH at theeast plug-cell [2]. During transient heating condition, however, the distributionof plasma parameters on the target plate has not been investigated precisely inthe previous study. In this paper, the detailed plasma parameters on the targetplate are studied. The distribution of the heat flux (PH) on the target plateduring transient heating pulse is observed to boost up for additional heating casein comparison to without heating pulse. Xe injection significantly reduces thePH on the target plate for only ICRF heated plasma, however, the PH reducesslightly for additional heating case. In particular, the increase ratio of the PH

profile on the target plate with ECH to without ECH increases with the increas-ing Xe plenum pressure. Time behavior of electron density (ne) shows that thene increases remarkably during ECH application timing. Furthermore, the ne

increases according to the increment of Xe plenum pressure. For without Xe in-jection, the electron temperature (Te) increases when ECH is applied. However,the Te did not increase during ECH application time when Xe is injected. Spec-troscopic data show that the emission intensity of Xe ion (XeII) is found to bevery small before and after the ECH injection time. However, it is also observedthat the emission significantly increases during the ECH injection. These out-comes represent that Xe is not a suitable radiator gas for sustaining the detachedplasma during transient heating pulse case.

[1] Y. Nakashima et al., Nucl. Fusion. 57, 116033 (2017).[2] Y. Nakashima et al., Nuclear Mater. Energy 18, 216 (2019).

O1-03

Study of magnetic configuration based on helical

coil system and attempt for numerical

optimization

Hiroyuki YAMAGUCHI1

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

The present-day advanced stellarators [1-4] are based on modular coil system togenerate magnetic configuration designed to have improved stability and confine-ment properties. For the modular coils systems, however, method to producedivertor configuration that is robust in the presence of plasma pressure has notbeen established. Modular coil ripples that are similar to those of toroidal fieldcoil ripples in tokamak may lead ripple transport and restrict the level of fast ionconfinement below that of tokamak. On the other hand, stellarators/heliotronswith continuous helical coil, such as Large Helical Device [5], can operate atsteady state with robust divertor legs. In addition, if toroidal field coils arenot used, there would not be ripple magnetic fields that exist in tokamaks andmodular-coil-based stellarators.

In this study, we construct advanced magnetic configurations based on continuous-helical coil system, greatly extending freedom in shape and arrangement of coilsfrom conventional ones. As a result, we have found magnetic configurations withhigh helical symmetry and quasi-omnigeneity with continuous helical coils with-out support of toroidal field coils. We also develop the first numerical optimizerfor magnetic confinement with helical coil system, OPTHECS (OPTimizer forHElical Coil System), that can optimize coil system aiming at target values ofphysics variables such as symmetry of field strength and neoclassical transport,and engineering ones such as plasma-coil distance. We will present the foundconfigurations and demonstrate the optimization using OPTHECS.

[1] C. D. Beidler et al., Fusion Technol. 17, 148 (1990)[2] D. Spong et al., Nucl. Fusion 41, 711 (2001)[3] J. Talmadge et al., Phis. Plasmas 8, 5165 (2001)[4] G. Neilson et al., J. Plasma Fusion Res. 78, 214 (2002)[5] A. Iiyoshi, M. Fujiwara, O. Motojima, N. Ohyabu, and K. Yamazaki, FusionTechnol. 17, 169 1990

O1-04

P1-01 Tetsuo OZAKIMeasurements of D/H Ratio Using Compact Neutral Particle Analyzer in LHD Deuterium Experiments

P1-02 Jing WUComplex Visible Spectrum Simulation and Experimental Study of Tokamak Plasma

P1-03 Xuan Nhat BUIEVALUATION OF DIVERTOR HEAT FLUX USING COMBINED PROBES ARRAY

P1-04 Kunihiro OGAWA1 MeV triton orbit analysis in EAST

P1-05 Hideaki MATSUURANumerical Study on Large-Angle-Scattering Effects Observed in the LHD Deuterium Plasma Using the Boltzmann-Fokker-Planck Model

P1-06 Withdrawn

P1-07 Liang LIUMeasurement of absolute carbon and helium impurity density profiles based on CXRS diagnostic on HL-2A tokmak

P1-08 Chunhua LIUThe stray laser light simulation and optimized design of beam dump for Thomson scattering system in HL-2Mtokamak

P1-09 Maoyuan LUOMeasurement of Electron Temperature Profile and Fluctuation with ECE Radiometer System in Heliotron J

P1-10 Toshiki KINOSHITAMacroscopic and microscopic instabilities in detached plasma of LHD

P1-12 Dechuan QIUDESIGN OF COMPACT MULTI-PATH THOMSON SCATTERING DIAGNOSTIC WITH SIGNAL DELAY SYSTEMIN HELIOTRON J

P1-13 Yoshio NAGAYAMAMeasurement of Electron Density Fluctuations by Using the O-mode Microwave Imaging Reflectometry (O-MIR) inTST-2 Spherical Tokamak

P1-14 Masaharu FUKUYAMAEvaluation of Quasi-Optical Mirror in Millimeter- and Submillimeter-Wave Range Using Three-DimensionalPrinter Technique

P1-15 Osamu WATANABEEvaluation of X-ray penetration pass through shield gap of a hard X-ray measurement system

P1-16 Naoto IMAGAWAParticle Transport Analysis by Pellet Injection for Characterization of Isotope Effect in LHD

P1-17 Shu ITOThe response of RMP on interchange instability in LHD

P1-18 Peerapat BOONYARITTIPONGProton Beam Production for Divertor Plasma Simulation Experiments in DT-ALPHA Device

P1-19 Kenji SAITOA scheme for ICRF heating of high-density core plasma in LHD

P1-20 Mitsutaka ISOBEMEASUREMENT OF NEUTRON YIELD FOR STRICT OBSERVANCE OF ITS BUDGET IN LHD

P1-21 Ryohei IKEDAEquilibrium analysis of tokamak plasma including the eddy current effects in TOKASTAR-2

P1-22 Dogyun HWANGBOInspection of Arc Trails Formed in Stellarator/Heliotron Devices W7-X and LHD

P1-23 Daichi KOBAYASHIEvaluation of Translation Velocity Control by Auxiliary Coils for Collisional Merging Formation of FRCs by 2-DResistive MHD Simulation

Poster 1 [November 5 (Tuesday) 13:40 - 15:40]

P1-24 Hisamichi FUNABAImpurity Transport of Externally Injected Metal impurities in the LHD Plasma with Impurity Hole

P1-25 Hiroshi OKAWAApplication to Plasma Immersion Ion Implantation by Inertial Electrostatic Confinement (IEC ) Device

P1-26 Akifumi IWAMOTOVoid-free Fuel Solidification in a Foam Shell FIERX Target

P1-27 Hirotaka CHIKARAISHIRenewal of large current dc power system for fusion plasma experiment facility LHD

P1-28 Mamoru SHOJIRadiation Resistant Camera System For Monitoring Deuterium Plasma Discharge In The Large Helical Device

P1-29 Osamu MITARAIComparative Studies on the Control Algorithm for the High-density Ignition Regime in FFHR-d1

P1-30 Kento MIYAMAEStudy on System Dynamics of Fuel and Burning in DD Start-up Scenario of a Fusion Reactor

P1-31 Kaoru OHYASimulation of Deuterium Retention in Tungsten by Repetitive Irradiation with High-Flux Deuterium Ions

P1-32 Yifan ZHANGDevelopment of a bipolar power supply for Ohmic heating on QUEST spherical tokamak

P1-33 Kosuke ASAITemperature dependence on co-deposition layer formed by sputtered tungsten and helium plasma

P1-34 Sho NAKAGAWAENGINEERING ANALYSIS FOR VACUUM VESSEL OF CFQS QUASI-AXISYMMETRIC STELLARATOR

P1-35 Ichihiro YAMADAInfluence of neutron irradiation on LHD Thomson scattering system

P1-36 Shunya NAKASONEThe investigation of pretreatment methods for liquid scintillation measurement of environmental water-samplesusing ion exchange resin

P1-37 Nopporn POOLYARATDESIGN OF POWER SUPPLY SYSTEM FOR MAGNET SYSTEMS OF THAILAND TOKAMAK -1

P1-38 Jiraporn PROMPINGNumerical Study of Supersonic Molecular Beam Injection System in Future Thailand Tokamak

P1-39 Kwangho JANGDesign of VFT and Multipactor Test Chamber for High Power Helicon Current Drive System in KSTAR

P1-40 Myoung-Jae LEEPropagation of lower-hybrid surface waves in semibounded warm plasmas

P1-41 Gediminas GAIGALASTheoretical investigation of E1 transitions for Pr II - Gd II ions

P1-42 Yoshihide SHIBATAStudy of the plasma response on external RMP field in a small tokamak device HYBTOK-II

P1-43 Kazuya HASHIGUCHIExperimental studies on ion-ion separation and their collection for direct energy conversion

P1-44 Tetsuya AKITSUA Study on the Detection of Plasma Generated Radicals in the Degradation of Dibromophenols in Water Solution

P1-45 Hayato TSUCHIYAStudy of reconstruction of microwave reflectometry image by machine learning

P1-46 Tomoko KAWATEDetection of Non-thermal Electrons in LHD Plasmas via Fe-line Spectroscopy

P1-47 Yoshiki MARUKIHigh density plasma generation in a thruster using a helical antenna with variable axial length

P1-48 Atsuya KURISUNOImprovement of electron collection using a magnetic field with low mirror ratio in a secondary electron directenergy converter simulator

P1-49 Hiromune YAMAZUIPlasma generation study on dependence of magnetic fields ratio between upstream and downstream for a singlehelical antenna thruster

P1-50 Quan SHIThe Formation of Cone Structure On Silicon Irradiated with Low Energy Helium Plasma: Flux and SampleTemperature Dependences

P1-51 Hiromasa TAKENOSimulation experiment of thermal load reduction by direct energy conversion using an ion beam injector

P1-52 Daichi OGATAMixer Comb-frequency Generator for Differential Phase Reflectometry

P1-53 Shota ERANEW IDENTIFICATION OF UV-VISIBLE EMISSION LINES FROM HIGHLY CHARGED TUNGSTEN IONS INCOBIT

P1-54 Hiroki TAKANOTwo-point measurements along the magnetic field in detached linear plasmas with laser Thomson scatteringmethod

P1-55 Koji ASAOKAMeasurement of Nitrogen Atom Density Generated by Spiral Shape Recombining Plasma

P1-56 Ryosuke OCHIAIDevelopment and evaluation of ion energy analyzer for energetic ion measurement in a linear plasma deviceNUMBER

P1-57 Shogo HATTORIPlasma potential measurements in detached plasmas by using electrostatic probes

P1-58 Hideaki MIURALarge eddy simulations of magnetized plasmas described by extended MHD model

P1-59 Ryutaro KANNOGlobal effect on collisional transport of tungsten impurity

P1-60 Hideo SUGAMAImproved Linearized Model Collision Operator for Kinetic Plasma Simulation

P1-61 Shinichiro TODAStudy on Effects of Zonal Flows and Trapped Electrons on Turbulent Transport by Reduced Models for HelicalPlasmas

P1-62 Kota YANAGIHARAQuasioptical modeling of the electron cyclotron resonance heating

P1-63 Shabbir Ahmad KHANKinetic full wave analysis of electron cyclotron waves in a tokamak plasma using integral operator method

P1-64 Isaya SAEKISimulation study of detached helium plasma in NAGDIS-II by using the fluid code LINDA

P1-65 Jie HUANG3-D equilibrium reconstruction under resonant magnetic perturbations on EAST

P1-66 Takazumi YAMAGUCHIEquivalent-Circuit Model for Superconducting Linear Acceleration System: Improvement of AccelerationPerformance

P1-67 Ayumu SAITOHHybrid Method Incorporated with Meshless approach for Electromagnetic Wave Simulation

P1-68 Haolun LIMolecular Dynamics Simulation on Structural Change of Tritium-substituted Polyethylene Using Reactive ForceField

P1-69 Shinsuke SATAKEOptimization study of heliotron devices

P1-70 Ryoya FUNABASHIHigh-speed analysis of heating and current drive with neutral beam injection in tokamak plasma

P1-71 Katsuji ICHIGUCHINumerical Study of Net Toroidal Current Effects on MHD Stability of LHD Plasmas

P1-72 Daisuke AOKINeutral effects on the structure of minimum enstrophy flows

P1-73 Hiroki HASEGAWAIsotope Effect on Plasma Coherent Structure Propagation

P1-74 Shuhei TOMIMATSUDynamics of resistive ballooning instability in PLATO tokamak plasma

P1-75 Gakushi KAWAMURAReconstruction of Radiation Profile of Neon-Seeded LHD Plasma from Bolometer Measurements with the Aid ofEMC3-EIRENE

P1-76 Rudolf TRETLERMHD Simulation of Plasma Sheet Thinning Due to Loss of Near-Earth Magnetotail Plasma

P1-77 Mieko TOIDASimulation study of trapped electron effects on positron acceleration by a shock wave in an electron-ion-positronplasma

P1-78 Arata NISHIMURANeutron Irradiation Effect on Critical Current and Critical Magnetic Field of Nb3Sn wire

P1-79 Hiroaki OHTANIOptimization of electromagnetic particle simulation code PASMO

P1-80 Yasuhiro SUZUKIImpacts of RMP fields on ELMs in a JT-60SA plasma

P1-81 Ritoku HORIUCHIEnergy Partition and Temperature Anisotropy in Merging Processes of two Spherical-Tokamak-type Plasmoids

Measurements of D/H Ratio Using Compact

Neutral Particle Analyzer in LHD Deuterium

Experiments

Tetsuo OZAKI1, Shuji KAMIO1, Kenji SAITO1,LHD Experimental Group1

1. National Institute for Fusion Science

The Compact Neutral Particle Analyzer (CNPA) [1] is one of the few devices,which can directly measure radial energetic particle spectra by combining witha pellet charge exchange measurement. Higher plasma performance can be ex-pected in LHD deuterium experiment. The monitoring of D/H ratio is veryimportant because the hydrogen is not replaced immediately by the deuteriumin LHD. A spectroscopic measurement is often used as the D/H monitor. How-ever, this is an information especially at the lower energy region near the plasmaedge. We are also interested in D/H ratio at higher energy region because theenergetic-ion-driven resistive interchange mode is relevant to the energetic par-ticle species. The particle energy distribution can be obtained by a magneticfield created by a permanent magnet in the CNPA. The mass separation can beperformed by an electric field created by condenser plates. Since it is designedto measure especially the hydrogen energy spectra, it is necessary to change theplate voltage in order to measure the deuterium. At a certain magnetic field,the deuterium energy, which has the same orbit as the hydrogen orbit in theequatorial plane, is half of the hydrogen energy. The deflection voltage for thedeuterium detection is half of the voltage in the hydrogen detection in order tointroduce the deuterium beam to the detector. At this time, the hydrogen beamreaches between the detector and the mid-plane. This means the mass rejectionfactor may be reduced because deuterium/hydrogen beam diameters are consid-erably large. The deuterium and the hydrogen have been measured over severalthousand shots by changing the plate voltage shot by shot. D/H ratios couldbe obtained by comparing similar two discharges. Figure 1 shows the D/(D+H)history over 3400 shots obtained from the energy integrated total flux. The ra-tios of D and H obtained by a visible spectroscopy are also plotted in Fig. 1. Inthis estimation, low energy components are dominant because lower componentis large in the CNPA. In the spectroscopic measurement, the contribution of lowenergy components in the periphery is also dominant. Therefore, both resultsare agreed in a wide range. About 300 discharges are necessary to replace thedeuterium gas to the hydrogen gas completely due to the large gas absorptionin the wall. We can conclude that the deuterium measurement is possible byadjusting the plate voltage. The tendency of the ratio in high energy components

P1-01

will be also described. [1] T. Ozaki et al., Rev. Sci. Instrum. 79 (2008), 10E518.

P1-01

Complex Visible Spectrum Simulation and

Experimental Study of Tokamak Plasma

Jing WU1, Lieming YAO1, Peng CHEN1, Hao WEN1,

Hangyu ZHOU2

1. University of Electronic Science and Technology of China2. Southwestern Institute of Physics

This research consists of two parts: the first section was the simulation part.In the simulation part, we created a main visible spectrum diagnosis of Toka-mak, which consisted of charge exchange complex spectrum(CXRS), neutralbeam emission spectrum(BES), Stark effect spectrum(MSE), fast ion Dα spec-trum(FIDA). The tokamak simulation programs were based on physical problemssuch as neutral beam and plasma interaction under different conditions, providingtheoretical data reference for diagnostic system design. The second part was theexperimental real-time data fitting procession, which aims to find the tokamakplasma visible spectrum diagnostic experimental results, to obtain the relevantphysical parameters, such as ion temperature Ti, ion density Ni, safe factor q,velocity of toroidal rotation Vt, velocity of polar rotation Vp .The study modelsdeal with the complex spectrum, establishes different elements, such as carbon,helium, hydrogen and other spectral models; establishes different materials ofwalls (inner wall and outer wall), such as carbon, tungsten and other reflectionmodels, to debug the inversion program. Then we established an electric fieldclose to the actual situation, magnetic field and other environments, to obtaina model of spectral Stark splitting and Zeeman splitting. We also establishedaccurate spectral inversion procedures for hydrogen/ deuterium edge fuel recy-cling and ash removal. The final result is compared with the actual experimentalresults to meet the accuracy and real-time requirements of the final measurementresults. [1] D .L. Yu, Y. L. Wei et al.,“ The motional Stark effect polarimeterin the HL-2A tokamak”, Rev. Sci. Instrum, vol. 8,. no.5, pp. 053508(1-4),May. 2014. [2] F. M. Levinton, Fonck. R. J et al.,“Magnetic field pitch anglemeasurement in the PBX-M Tokamak using the motional Stark effect”, Phys.Rev. Lett, vol. 63, no. 6, pp. 2060-2063, Feb. 1989. [3] D. Wroblewski. andL. L. Lao,“ Polarimetry of motional Stark effect and determination of currentprofiles in DIII-D”, Rev. Sci. Instrum, vol. 63, no. 10, pp. 5140-5147, Mar.1992. [4] B. C. Stratton.,“ Instrumentation for the joint European torus mon-tional Stark effect diagnostic”, Rev. Sci. Instrum, vol. 70, no. 1, pp 898-901,Jan. 1990. [5] B.Rice. W et al.,“Motional Stark effect upgrades on DIII-D”,Rev. Sci. Instrum, vol. 66, no. 1, pp. 373-375, Jan. 1995. [6] R. J. E. Jasperset al.,“ Validation of the ITER CXRS design by test on TEXTOR”, Rev. Sci.Instrum, vol. 79, no. 10, pp. F526(1-4), Oct. 2008. [7] N. A. Pablant et al.,

P1-02

“Measurement of the magnetic field on DIII-D using Intensity and spacing of themotional Stark multiplet”, Rev. Sci. Instrum, vol. 79, no. 10, pp. F517(1-4),Oct. 2008.

P1-02

EVALUATION OF DIVERTOR HEAT FLUX

USING COMBINED PROBES ARRAY

S. Bui1, H. Muraoka1, Y. Yamamoto1, H. Matsuura1,R. Matoike2, S. Ohshima 2, T. Mizuuchi2

1. Osaka Prefecture University2. IAE, Kyoto University

In a fusion reactor, one of the most important components is the divertor andits purpose is to remove impurities from the core plasma. Because the divertorreceives a very large heat flux as well as various surface damages, divertor heatflux needs to be estimated and diminished for safe operation. To do it, precisebalance model must be established with suitable boundary conditions.

Special calorimeters which include two or more thermocouples were installedto the armer tile of the heating beam in Compact Helical System (CHS) [1]and Large Helical Device (LHD) [2]. Temperature gradient in the target canbe measured directly and it is easy to estimate the heat flux evolution with thegradient method. In GAMMA 10/PDX experiment [3], gradient method has beenused for estimating ICRF plasma of 400ms. However short pulses (10ms) heatflux with additional ECH cannot be reproduced well due to the signal ’s delayof approximately 100ms.

In terms of divertor heat flux studying, thin film type calorimeter can be apromising and useful device due to its fast response to temperature data. InHeliotron J experiment, divertor probes array is upgraded, it has thin film typecalorimeters in order to improve the signal responses. From our calculation, heatpulses can be reproduced in a wider range of durations compared to that ofGAMMA 10/PDX experiment [3].

This work is partially performed with the support and under the auspices ofthe NIFS Collaborative Research Program. (NIFS18KLPR044/ NIFS18KUHL084/NIFS18KUGM134).

REFERENCES: [1]M. Osakabe et al,“Development of fast response calorime-ter for neutral beam shine-through measurement on CHS”, Rev. Sci. Instrum.72, 586 (2001).

[2]H. Matsuura et al,“ The calorimetric estimate of shine-through power ofthe neutral beam system for plasma monitoring”, 10th AEPSE (2015, Jeju) 22a-C- 3.

[3]H. Matsuura et al,“ Comparison of two inverse heat conduction modelsfor heat flux measurement in the GAMMA 10/PDX,”I. Trans. Plasma Sci. 47,3026 (2019).

P1-03

1 MeV triton orbit analysis in EAST

Kunihiro OGAWA1,2, Guoqiang Zhong3, Ruijie Zhou3,Kai Li3, Mitsutaka Isobe3, Liqun Hu1,2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI3. Institute of Plasma Physics Chinese Academy of Sciences

A fusion burning plasma is sustained by DT born energetic alpha particles. There-fore, energetic alpha particles must be well confined. In deuterium experiment,1 MeV tritons are created by D(d, p)T reactions. 1 MeV triton is regarded assimulated DT born alpha particles because kinetic parameters such as Larmorradius and precession frequency are almost the same. In addition, 1 MeV tritonshave the isotropic birth profile as same as alpha particles. A study of 1 MeVtriton confinement has been widely and intensively performed in fusion devicesfor understanding the alpha particle confinement. They used triton burnup ratio,which is defined as the ratio of the DT neutron yield on DD neutron yield, as anindex of triton confinement capability. Recently, installation of NAS was com-pleted and measurement of DT/DD neutron yield is now initiated by using Si foilsin Experimental Advanced Superconducting Tokamak (EAST). To understand 1MeV triton confinement/loss in EAST, 1 MeV triton orbit analysis is performedin various Ip cases. The plasma equilibrium is given by EFIT code. The birthprofile of 1 MeV triton is given by random number generator based on the 1 MeVtriton emissivity calculated using NUBEAM code [3]. The collisionless orbit of1 MeV tritons are followed using LORBIT code [4]. Here, the velocity of tritonis given using a random number generator. Note that orbit following time is setto be 1 ms which is much less than the Spitzer slowing down time. It is shownthat the number of lost tritons decreases with the increase of Ip. Note that thenumber of lost tritons increases rapidly at 10-6 s, and then are almost saturatedat 10-5 s regardless of Ip. Pitch angle distribution of confined 1 MeV triton showsthat the tritons existing wider pitch angle range can be confined in higher Ip casecompared with lower Ip case. Confinement property of 1 MeV triton orbit willbe discussed in this presentation.

[1] A. Pankin et al., Comp. Phys. Commun. 159 (2004) 157. [2] M. Isobe etal., J. Plasma Fusion Res. SERIES 8 (2009) 330.

P1-04

Numerical Study on Large-Angle-Scattering

Effects Observed in the LHD Deuterium Plasma

Using the Boltzmann-Fokker-Planck Model

Hideaki MATSUURA1, Shota SUGIYAMA1,Kento KIMURA1, Tomoki URAKAWA1,Takeo NISHITANI2, Kunihiro OGAWA2,3,Yasuko KAWAMOTO2, Naoki TAMURA2,3,Mitsutaka ISOBE2,3, Masaki OSAKABE2,3

1. Kyushu University2. NIFS3. SOKENDAI

In a thermonuclear plasma, energetic ions play important roles in variousstages of fusion reactor operation. The energetic ions slow down, deposit theirenergy via Coulombic collisions, and create energetic tails (non-Maxwellian com-ponents) in bulk-ion velocity distribution functions. It has been predicted thatthe slowing-down behavior is affected by the nuclear elastic scattering (NES) [1].NES is a large-angle-scattering process, and a large fraction of the fast-ion energyis transferred to bulk ion in a single scattering event. The large-angle-scatteringeffect would appear always in burning plasma operation and experiment. Some-times it could be an influential phenomenon for reactor-grade plasma [2]. Itis important to grasp the detailed property and validate the related simulationmodel.

We have developed the Boltzmann-Fokker-Planck (BFP) model to analyzethe large-angle-scattering process in a high-temperature plasma [3], and haveattempted to observe a knock-on tail in“deuteron”distribution function createdby beam-injected“protons”in 19th and 20th campaigns on the LHD experiment.We have made efforts to understand the observed phenomena using the BFPmodel. In this paper the large-angle-scattering effects in the burning plasmaand their dependency on the plasma condition are studied on the basis of theBFP model. Discussion is made for the plasma confinement properties and theircorrelation with the large-angle-scattering effects for various plasma parameterranges.

[1] J. J. Devaney and M.L.Stein, Nucl. Sci. Eng., 46 (1971) 323.[2] H. Matsuura, et al. J. Plasma Fusion Res., 91 (2015) 441. (in Japanese)[3] e.g. H. Matsuura, et al., Plasma Phys. Contr. Fusion, 53 (2011) 035023.

P1-05

Measurement of absolute carbon and helium

impurity density profiles based on CXRS

diagnostic on HL-2A tokmak

Liang LIU1

1. Southwestern Institute of Physics

The absolute carbon and helium impurity density profiles are calculated withtwo methods based on the charge-exchange recombination spectroscopy (CXRS)diagnostic during L and H mode on HL-2A tokamak. One method to derivethe impurity concentration is to use the absolutely calibrated active spectrumintensity, the neutral beam density and the CXRS emission rate. The neutralbeam density is calculated with the beam attenuation process, which is deter-mined by the electron density profile, the beam stopping rate, and so on. Both ofthe CXRS emission and beam stopping rates are from the Atomic Data AnalysisStructure (ADAS) database. The other method is to combine the flux of CXRSand beam emission spectroscopy (BES) together, with the beam attenuation cal-culation omitted. In this method, a newly developed tri-band and high spectralresolution spectrometer is applied, which is able to provide the measurementsof He II (468.57 nm), C VI (529.1 nm), and Dα (656.1 nm accompanied byBES) simultaneously. Both of the CXRS and beam emission rates are neededand inquired from the ADAS database in the second method.

P1-07

The stray laser light simulation and optimized

design of beam dump for Thomson scattering

system in HL-2M tokamak

Chunhua LIU1, Yuan Huang1, Zhen Feng1,Yuqin Wang1, Zhipei Hou1

1. Southwestern institute of physics4. Southwestern institute of physics5. Southwestern institute of physics

Laser beam dump is one of the critical components of Thomson scattering (TS)system to decrease stray laser light to an acceptable level. Two compact beamdumps are needed to absorb residual laser energy of HL-2M TS system. One willbe installed at the inner wall of the HL-2M vacuum vessel in low field side, and theother will be set up in the divertor region. Chevron beam dump has a compactstructure, and the lifetime of which manufactured by molybdenum materials,under the harsh thermal and electromagnetic loads could meet the demand of TSsystem in ITER [1], as well as TS system in the central region and divertor regionof HL-2M. But compared with a beam dump that is set up outside the machine,the stray laser light caused by an inside beam dump maybe much stronger toseriously influence the scattered light measurement, and there is no referencesdiscussed the problem before. So it is important to simulate the stray laser lightcaused by the chevron beam dump. In order to analyze its mechanical structureand simulate the stray laser light caused by it, bi-directional reection distributionfunction (BRDF) of molybdenum is measured by scatterometer (Fraunhofer IOF)in 532 nm. Then the surface properties of Mo material are loaded into the ray-tracing software to simulate the stray laser light emitting from the beam dump.The light source parameters are the same as that of the TS system on HL-2M.The light absorbing threshold is 1e-10. According to the beam dump structureand ray-tracing results, it is obvious that the stray laser light emits in the coneangleα 1 of about 65 degrees. Simulation results show that increasing the heightof the entrance port could not decrease effectively the cone angle α. But aftera reflecting cover is added in front of the beam dump, the stray laser light emitsin the cone angle α 2=23 degree. In this condition, the strength of stray laserlight directly entering into collection widow becomes acceptable. So the optimizedchevron beam dump could be utilized to absorb laser energy on HL-2M TS system.

[1] E. Yatsuka, et al., Review of Scientific Instruments 84, 103503 (2013).

P1-08

Measurement of Electron Temperature Profile

and Fluctuation with ECE Radiometer System in

Heliotron J

Maoyuan LUO1, Kazunobu NAGASAKI2, Gavin WEIR3,

Hiroyuki OKADA2, Takashi MINAMI2,Shinichiro KADO2, Shinji KOBAYASHI2,

Satoshi YAMAMOTO4, Shinsuke OHSHIMA2,Tohru MIZUUCHI2, Shigeru KONOSHIMA2,Yuji NAKAMURA1, Akihiro ISHIZAWA1

1. Graduate School of Energy Science, Kyoto University2. Institute of Advanced Energy, Kyoto University3. Max-Planck Institute Greifswald4. Naka Fusion Institute, National Institute and Radiological Science and

Technology

A multi-channel electron cyclotron emission (ECE) radiometer system is beingdeveloped in Heliotron J to measure electron temperature profile and evaluate Tefluctuation. The conventional ECE part of this system includes 16 channels andcan measure electron temperature profile from 58 GHz to 74 GHz, correspondingto 2nd harmonic X-mode. The correlation ECE part is composed of“ CECE-RF”and“CECE-IF”systems. The CECE-RF side allows flexible RF frequency,ranging from 56 GHz to 88 GHz, and the CECE-IF side has 4 frequency-fixedchannels, corresponding to 60 GHz, 64 GHz, 68 GHz and 72 GHz. The frequencycovers the core region to the edge one. The signals from these two sides sharea same source of electron cyclotron emission and thus are possible to estimateelectron temperature fluctuation through correlation analysis. An ECE profilehas been obtained and compared with data from Thomson scattering system inan electron cyclotron heated plasma, and we also estimated the core Te fluctuationlevel through correlation analysis to eliminate contribution from thermal noise.

P1-09

Macroscopic and microscopic instabilities in

detached plasma of LHD

Toshiki KINOSHITA1, Kenji TANAKA1,2,Masahiro KOBAYASHI2,3, Yuki TAKEMURA2,3,

LHD experiment group

1. Interdisciplinary Graduate School of Engineering Science, Kyushu Univer-sity, Kasuga 816-8580, Japan

2. National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292,Japan

3. SOKENDAI, 322-6 Oroshi, Toki, Gifu 509-5292, Japan

A plasma detachment is essential to reduce the heat load on the divertor platein a tokamak and a stellarator/heliotron. In LHD, detachment was achieved inhigh density regime (line averaged density is 5 ∼ 10 × 1019[m−3] ) with Reso-nant Magnetic Perturbation (RMP) at outwardly shifted configuration [1]. Inthe recent LHD deuterium experimental campaign, a new regime of detachmentwas found. After transition from the attached phase to the detached phase, sta-tionary detachment was sustained. Then, strong instabilities appeared suddenlymaintaining detachment. In order to investigate the change of the turbulencedriven transport, the spatial structure of the instability was measured by usingan 80ch CO2 laser imaging heterodyne interferometer [2]. Also, microturbulencewas measured by using a two dimensional phase contrast imaging [3]. The rolesof mcroscopic and microscopic instabilities on detachment will be discussed.

References

[1] M. Kobayashi, et al., Nucl. Fusion, 53, 093032 (2013)

[2] K. Tanaka, et al, Rev. Sci. Instrum., 75, 3429 (2004)

[3] K. Tanaka, et al, Rev. Sci. Instrum., 79, 10E702 (2008)

P1-10

DESIGN OF COMPACT MULTI-PATH

THOMSON SCATTERING DIAGNOSTIC

WITH SIGNAL DELAY SYSTEM IN

HELIOTRON J

Dechuan QIU1, Takashi Minami2, Takuya Nishide1,

Masahiro Miyoshi1, Yuta Yamanaka1, Shinichiro Kado2,Hiroyuki Okada2, Kazunobu Nagasaki2,

Shinji Kobayashi2, Shinsuke Ohshima2, Tohru Mizuuchi2,Shigeru Konoshima2, Ryo Yasuhara3

1. Graduate School of Energy Science, Kyoto University, Gokasho, Uji, Kyoto,Japan

2. Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto,Japan

3. National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan

Suprathermal electrons existing in the electron cyclotron heating (ECH) plasmacan be investigated by measuring anisotropic electron velocity distribution (EVD).Thomson scattering system has a potential of obtaining EVD if the signal tonoise ratio is sufficient. In order to increase the scattered photon number perlaser pulse, multi-path Thomson scattering (MPTS) is applied and it can alsoprovide a pulse train in single laser shot. However, because there’s limited roomprovided for setting optical path in Heliotron J, scattered light signals producedby normal multi-path Thomson scattering system with a single Pockels cell areoverlapping and cannot be separated, due to a very short time interval of thepulse train. In this study, a design of double Pockels cells inserted into normalMPTS based on polarization control technique is proposed to extend the intervaltime. By switching on the second Pockels cell, the polarization direction of laserlight is changed and laser light is guided into another path called a signal delaysystem (SDS) and confined in it during several reflection periods. The intervalbetween signals can be controlled by the reflection number in SDS. The relation-ship between the intensity of scattered light signal and the reflection number oflight confined in SDS was estimated by assuming the transmission rate of Pockelscell and other optical components existing in SDS as 98 percent and 99 percent,respectively. The result showed that the intensity of the signal pulse train wasdecreased by 23 percent after two reciprocation in SDS with 100 ns time delay.

P1-12

Measurement of Electron Density Fluctuations

by Using the O-mode Microwave Imaging

Reflectometry (O-MIR) in TST-2 Spherical

Tokamak

Yoshio NAGAYAMA1, Akira EJIRI2, Yuichi TAKASE2,Hideya NAKANISHI3, Masaki OHSUNA3,

Hayato TSUCHIYA3, Soichiro YAMAGUCHI4

1. College of Science and Technology, Nihon University2. Graduate School of Frontier Sciences, University of Tokyo3. National Institute for Fusion Science4. Faculty of Engineering Science, Kansai University

Micro instabilities, MHD instabilities and Turbulences are revealed in the electrondensity fluctuation. The Microwave Imaging Reflectometry (MIR) is potentiallya diagnostic tool to observe those instabilities and turbulence, because it is oneof the imaging diagnostics of the electron density fluctuations. Thus, MIR hasbeen intensively developed in Large Helical Device (LHD) and some interestingphenomena had been observed [1]. The O-mode MIR (O-MIR) system has beeninstalled in a Spherical Tokamak (ST) named TST-2. The reflection surface ofthe O-mode wave should be an equi-density surface, which is nearly flat andis located near the MIR device. This helps interpretation of the obtained MIRimage. In TST-2, two frequencies which are separated by 1.4 GHz are illuminatedsimultaneously so that the radial wavenumber of the density fluctuation can bemeasured. The illumination wave frequency is 23 - 35 GHz. So, the observabledensity is 0.5−1.5×1019m−3. Suppose the central density is 1.2×1019m−3 and thedensity profile is parabolic, the region of 0 < r/a < 0.7 can be measured by theO-MIR. The microwave image from the scattered wave is made on the imagingdetector by the imaging optics. The microwave imaging detector is Horn-antennaMillimeter Imaging Detector (HMID)[2]. The detected channels are 6 (poloidal)× 6 (toroidal) × 2 (radial) channels. Phase and amplitude of MIR signals andsome ofTST-2 signals are sampled every 0.5 µsec by using 224 digitizers, of whichdata are taken and stored by the LABCOM system at NIFS via SNET. Observedfrequency spectrum of MIR signals extends to higher frequencies.

This work is supported by NIFS (NIFS17KLEP022), and by KAKENHI (No.17K18772).[1] Y. Nagayama, et al: Rev. Sci. Instrum. 83, 10E305 (2012).[2] Y. Nagayama, et al: Rev. Sci. Instrum. 88, 044703 (2017).

P1-13

Evaluation of Quasi-Optical Mirror in Millimeter-

and Submillimeter-Wave Range Using

Three-Dimensional Printer Technique

Masaharu FUKUYAMA1, Hiroshi IDEI2,Ryuichi ASHIDA1, Daichi OGATA1, Michihiro KUDO1,

Ryouya KATO1, Ryuya IKEZOE2

1. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity

2. Research Institute for Applied Mechanics, Kyushu University

Millimeter and submillimeter waves are widely used for plasma heating and diag-nostics. Typical applications of the waves are electron cyclotron heating and cur-rent drive (ECHCD), radiometry, reflectometry, interferometer and some scatter-ing measurements. Quai-optical mirrors required for the applications are broadlydesigned using Gaussian optics. Moreover, Kirchhoff-integral approaches are ef-fective to treat the wave propagation from and to a waveguide aperture. Thefocusing launcher-mirrors of the ECHCD and interferometer systems on QUEST(Q-shu University Experiment with steady-state Spherical Tokamak) were de-signed by the Kirchhoff integral code and Gaussian optics [1-2]. A phase-reversalquasi-optical (QO) mirror of the HE11 mode was also designed with Kirchhoffintegral code for the QO polarizer in the transmission line of the ECHCD system[3]. The mirror designs are based on the phase-matching condition without tak-ing amplitude asymmetry on the mirror-surface into consideration properly. Toconfirm the effect of amplitude asymmetry, the designed mirror performance hasbeen verified with full wave simulation or low power tests with fabricated mirrors.The full-wave simulations using finite element (FLM) and finite-difference time-domain (FDTD) methods are available and powerful for the mirror evaluation,However, the simulations of higher operating frequency waves, such as millimeterand sub-millimeter waves, need more computing resources and are difficult in par-ticular to the wave propagation. In addition, a metal-mirror fabrication is costlyand takes a lot of time for precise machining required for the higher frequencyapplications. A simple evaluation method should be considered to assess the QOmirror performance.

In this presentation, a new evaluation method of the QO mirror using there-dimensional (3D) printer technique is proposed. The printer can precisely fab-ricate rather complex surface as requested at the reasonable cost. Since the 3Dprinting material is normally poly-lactic-acid, silver coated copper powder wassprayed on the poly-lactic-acid surface of the 3D-printed mirror to reflect thewave. The performance of the 3D-printed mirror was tested at low power facili-

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ties. The designed mirror surface is discussed with the low-power test results andthe full-wave simulations.[1] H. Idei, et al.,“ 28-GHz ECHCD system with beam focusing launcher on theQUEST spherical tokamak”, Fusion Eng. Des., in press.[2] M. Yunoki, et al., “ Prototype of a Quasi-optical Launcher System of 4mm Round-trip Interferometer for the QUEST Spherical Tokamak Experiments”,Plasma and Fusion Res., in press.[3] M. Fukuyama, et al.,“ Quasi-optical Polarizer System for ECHCD Experi-ments in the QUEST”, Fusion Eng. Des., in press.

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Evaluation of X-ray penetration pass through

shield gap of a hard X-ray measurement system

Osamu Watanabe1, Hibiki Yamazaki1, Yuki Aoi1,Akira Ejiri1, Yongtae Ko1, Kotaro Iwasaki1,James Rice1, Kyohei Matsuzaki1, Peng Yi1,Yuki Osawa1, Naoto Tsujii1, Yuichi Takase1

1. The University of Tokyo

On the TST-2 spherical tokamak (ST) device, formation and sustainment ofST plasma have been studied using 2.45 GHz, 21 MHz [1] and 200 MHz RFpower. Toroidal current is driven by the RF power, and closed magnetic surfacesare formed, leading to an ST configuration. In order to establish a stable andreliable start-up and sustainment scenario, the physics of the current drive andaccompanying phenomena should be investigated, not only in TST-2, but also inother ST devices. In TST-2, two straps antenna for high harmonic fast wave (21MHz) was used, and a dielectric-loaded waveguide array antenna was used forlower hybrid wave (LHW: 200 MHz) in the past. After that, a combline antennaare developed to increase the plasma current driven by LHW. Currently, TST-2has two combline antennas: one on the outboard side on the mid-plane and theother on the top side of the plasma. The plasma current started-up by using theoutboard side antenna can be ramped-up by using the topside antenna, and theresultant maximum plasma current is 25 kA. The wave physics is numericallyinvestigated by COMSOL [2]: a finite element code and is experimentally studiedby magnetic probes. In order to obtain the wave power absorption profile and tospeculate the driven current profile, soft and hard X-rays from the LHW sustainedplasma are measured by scintillators made of the sodium iodide (NaI) or Ceriumdoped Lutetium Yttrium Orthosilicate (LYSO). A pinhole camera configuration(with a tangential viewing sight, with a thin scintillator plate) is useful to identifythe line of sight and to form a multi-spatial channel system. We use tungstenheavy alloy and lead to form a pinhole configuration, in which the scintillatorsshould be shielded from X-rays which do not pass through the pinhole. However,due to the limited space and due to technical reasons, the parts of the shield havemany interstices and thin sections, and the shielding of such unwanted X-raysmay not be perfect, and quantitative evaluation (i.e., simulation) of such X-raysis necessary. The X-ray penetration pass through the shield gap is computed byPHITS code Ver. 3.11 [3]. Assuming simplified configurations with various shieldinterstices and thickness, the effects of the unwanted X-rays are evaluated. Then,the effects in the actual configuration used in TST-2 is evaluated. It was foundthat the signal (i.e., the sensitivity) is not uniform on the scintillator surface,although all parts and light source are right and left (i.e., toroidally) symmetric

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except for the vacuum vessel of the TST-2. The vessel asymmetry arises from thetangential viewing sight configuration, and care must be taken in the analysis.

[1] Oosako T., et al. ”Parametric decay instability during high harmonic fastwave heating experiments on the TST-2 spherical tokamak” Nuclear Fusion 49.6(2009): 065020.

[2] Yajima S., et al. ”Current Drive Experiment Using Top/Outboard SideLower Hybrid Wave Injection on TST-2 Spherical Tokamak” Plasma and FusionResearch 13 (2018): 3402114.

[3] Sato T., et al. ”Particle and heavy ion transport code system, PHITS,version 2.52” Journal of Nuclear Science and Technology 50.9 (2013): 913-923.

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Particle Transport Analysis by Pellet Injection

for Characterization of Isotope Effect in LHD

N. Imagawa1, H. Yamada1,2, K. Miyamae1,

T. Yokoyama1, K. Ida2,3, R. Sakamoto2,3, K. Fujii4,M. Yoshinuma2, G. Motojima2,3

1. The university of Tokyo2. National Institute for Fusion Science/NIFS3. SOKENDAI4. Kyoto University

Particle transport in magnetically confined plasmas has been analyzed bymeans of transient response to pellet injection on LHD. This study aims at char-acterization of isotope effect on particle transport.

The most promising fusion reaction for a fusion reactor is the reaction ofdeuterium and tritium. Both deuterium and tritium are isotopes of hydrogen,and the control of their concentration of 50/50 is required to maximize fusionpower output. Nonetheless, difference and similarity of particle transport of theseisotopes have not been identified experimentally yet.

Recently, deuterium plasma experiment has begun in LHD, and also measure-ment separating hydrogen (H) and deuterium (D) density profiles has becomeavailable by bulk charge exchange recombination spectroscopy. Then, the elab-orated experimental study on particle transport of H and D has been enabledtogether with a multi-barrel solid hydrogen pellet injector. The transport of Hand D injected as a pellet into the NBI-heated plasma has been investigated bythe temporal change in density profiles through diffusion after pellet injection.The diffusion coefficient D and the pinch velocity V are obtained by the followingequation.

Γ=D∇ n+nV

∇ n is a density gradient, and Γ is a particle flux. It should be empha-sized that particle transport of electron, H, and D can be examined separatelyby Thomson scattering and bulk charge exchange recombination spectroscopy.In particular, the case of hydrogen pellet and the case of deuterium pellet arecompared. The presence or absence of isotope effects in the particle transport isdiscussed based on this result.

This work is supported by the National Institute for Fusion Science grantadministrative budgets NIFS18KLPP051 and JSPS KAKENHI Grant NumbersJP17H01368.

P1-16

The response of RMP on interchange instability

in LHD

Shu ITO1, Kiyomasa WATANABE1,2,Satoru SAKAKIBARA2, Yuki TAKEMURA2,

Sadao MASAMUNE3

1. Nagoya Univ.2. NIFS3. Chubu Univ.

MHD instability degrades the plasma confinement in the fusion reactor. TheRMP is considered as an effective knob to suppress the MHD instabilities. Inthis study, we apply the RMP by the external coils to the plasmas of the LHD(Large Helical Devise) with the MHD instability m/n=1/1 and m/n=2/1 andinvestigate the response of the plasmas. Here m and n are the poloidal andtoroidal mode number. And in the LHD, the resonant surface of m/n=1/1 islocated in the peripheral region (the normalized minor radius, r/a 0.9), and thatof m/n=2/1 is in the bulk region (r/a 0.5). In both cases, the observed resonantmagnetic fluctuation decreases with the external RMP field. On the contrary,in the m/n=1/1 case, the fluctuation gradually decreases as the RMP increases,and in the m/n=2/1 case, the fluctuation gradually decreases up to a the RMPfield and is suddenly suppressed as the RMP increases. The operation magneticfield strength is different between the m/n=1/1 and the m/n=2/1 cases. In thefuture, we should confirm the reason why the behavior of the plasma responsewith m/n=1/1 and the m/n=2/1 are different due to the operation magnetic fieldstrength and/or the mode structure of the fluctuation.

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Proton Beam Production for Divertor Plasma

Simulation Experiments in DT-ALPHA Device

Peerapat BOONYARITTIPONG1, Hiroyuki TAKAHASHI1,

Atsushi Okamoto2, Kenta Ogasawara1,Takeshi SAIKYO1, Tomohiro Seini1, Kenji Tobita1,

Sumio Kitajima1

1. Department of Quantum Science and Energy Engineering, Tohoku Univer-sity, Japan

2. Department of Applied Energy, Nagoya University, Japan

Crucial problem for divertor plates is the high heat flux from core plasma. Apromising method to reduce heat flux incoming to the divertor plate is to useplasma detachment by inducing volumetric recombination. However, it has beenindicated that the Rydberg atom produced by volumetric recombination may bere-ionized due to energetic electrons released by Edge Localized Modes [1]. On theother hand, the interaction between recombining plasma and energetic ion is notyet clearly understood. The DT-ALPHA device uses radio-frequency dischargeto produce plasma, which allows superimposition of ion beam to the recombiningplasma. The experiment using helium plasma indicates that recombination rateis reduced due to collision with energetic ion [2]. However, hydrogen ion beamis required to investigate the energetic ion interaction with molecular assistedrecombination process, and injection of multiple hydrogen ion species may resultin ambiguous understanding of results due to different velocity of each ion. Thus,it is necessary to inject ion beam with one ion specie to simplify the results. Weintend to utilize Wien filter to separate hydrogen ion beam. We have investigatedthe hydrogen ion separation efficiency of Wien filter, and the effect of Wien filterto both helium and hydrogen ion beam trajectory in DT-ALPHA device usingsimulation program [3]. Currently, the preparation for the installation of Wienfilter is ongoing. Based on equations in Ref. 4, we have predicted the proton beamratio from ion source geometry and magnetic configuration. After the Wien filteris installed, the ratio and amount of each extracted hydrogen ion beam will beinvestigated. The experiment results of proton beam production compared withthe prediction results by calculation will be presented.

The work was partly supported by the Japan Society for the Promotion ofScience (JSPS) Grant-in-Aid for Young Scientists (B) 17K14895

[1] N. Ohno, et al., Phys. Plasmas 6, 2486 (1999).[2] H. Takahashi et al., Physics of Plasmas 23, 112510 (2016).[3] P. Boonyarittipong et al., Plasma Fusion Res. 13, 3402102 (2018).[4] Y. Okumura et al., Rev. Sci. Instrum. 55, 1 (1984).

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A scheme for ICRF heating of high-density core

plasma in LHD

Kenji SAITO1,2, Ryosuke Seki1,2, Shuji Kamio1,

Hiroshi Kasahara1, Tetsuo Seki1

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies), Departmentof Fusion Science

In the large helical device (LHD), ion cyclotron range of frequencies (ICRF)heating has been conducted with various heating methods, for example minorityion heating, mode conversion heating and second harmonic heating [1]. Here,we introduce high-density plasma heating of deuterium plasma in LHD. In toohigh-density plasma, it is difficult to inject the neutral beam deeply into theplasma core and cut-off exists in the electron cyclotron heating (ECH). How-ever, an advantage of ICRF heating is that there occurs no cut-off even if densityis very high. When the ion cyclotron resonance layer exists around the saddlepoint of the magnetic field strength, high heating efficiency in minority ion heat-ing was achieved because the gradient of the strength of the magnetic field issmall [2]. In this configuration, it was shown that the high power-absorptionoccurs around the magnetic axis with the second harmonic heating in the case ofhigh-density deuterium plasma by the calculation with a simple model of ICRFheating. Wave number perpendicular to the static magnetic field increases withthe plasma density, and it enhances the finite Larmor radius effect in the secondharmonic heating. Enhanced finite Larmor radius effect and the large amountof resonant ions enable the intense absorption. In the case of the standard mag-netic configuration of Rax=3.6 m, the peak of power deposition profile locatesin off-axis region near the normalized minor radius of 0.4 [2]. However, the coreheating is possible by locating the Shafranov shifted magnetic axis around thesaddle point. By increasing the frequency, third harmonic heating will be alsopossible. Though the intensity of absorption decreases, more localized heatingon axis is possible because the finite Larmor radius effect is more essential in thethird harmonic heating.

References[1] K. Saito, et al., Fusion Sci. Technol. 58, 515 (2010).[2] K. Saito, et al., Nucl. Fusion 41, 1021 (2001).

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MEASUREMENT OF NEUTRON YIELD FOR

STRICT OBSERVANCE OF ITS BUDGET IN

LHD

Mitsutaka ISOBE1,2, Kunihiro OGAWA1,2,Takashi KOBUCHI1, Hitoshi MIYAKE1, Takuya SAZE1,

Makoto KOBAYASHI1,2, Takeo NISHITANI1,Masaki OSAKABE1,2, LHD Experiment Group1

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies)

The Large Helical Device (LHD) deuterium operation has been conducted sinceMarch 2017 [1,2]. With the start of deuterium operation, the LHD is registeredas a radiation generating device in the Nuclear Regulation Authority of Japan,and as a result, weekly, three month, and annual neutron yield budgets wereset from the standpoint of radiation safety. Therefore, management of neutronyield is essentially required in proceeding the LHD deuterium experiment. In theLHD, neutron emission rate and yield have been measured in all of the dischargeswith in-situ calibrated ex-vessel neutron flux monitor (NFM) characterized bywide dynamic range capability over 109 in cps [3,4]. In addition to neutronyield management, the NFM plays an important role in assessing tritium yield indeuterium plasmas. Tritium yield must be managed as well in accordance withthe agreement of local governments. As an aim of neutron yield management,prior to submission of experiment proposal, all proposers, including domestic andinternational collaborators, are requested to evaluate expected maximum neutronyield for his/her proposed machine time by use of the FIT3D-DD code, whichis based on the steady-state Fokker-Planck model. Each proposer must submitthe proposal together with expected maximum neutron yield for his/her ownpurpose. Based on this, the entire experimental plan in each campaign is drawnup so that the expected maximum neutron yield in the campaign is less than 60%of the annual neutron budget. In terms of hardware measure toward observance ofneutron yield budget, the NFM in LHD is equipped with strict interlock system.The system alerts, and terminates the LHD plasma once neutron emission ratein a shot and/or neutron yield integrated over ongoing campaign exceed the setpoint which is set to be 80% of the permitted value for neutron emission rate andyield. Also, once a serious fault occurs in NFM, the LHD cannot be operated.The performance of the LHD NFM is regularly inspected before the start of thecampaign. In this paper, our way of thinking and manner toward strict observanceof neutron yield budget and experiences accumulated through the first and the

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second deuterium campaigns for neutron emission rate and yield in the LHD aredescribed.

[1] Y. Takeiri et al., IEEE Trans. Plasma Sci. 46 (2018) 2348. [2] M. Osakabeet al., IEEE Trans. Plasma Sci. 46 (2018) 2324. [3] M. Isobe et al., Rev. Sci.Instrum. 85 (2014) 11E114. [4] M. Isobe et al., IEEE Trans. Plasma Sci. 46(2018) 2050.

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Equilibrium analysis of tokamak plasma including

the eddy current effects in TOKASTAR-2

Ryohei IKEDA1, Takaaki FUJITA1,

Atsushi OKAMOTO1, Hideki ARIMOTO1,Kouhei YASUDA1, Sora KIMATA1, Keitaro KADO1

1. Graduate School of Engineering, Nagoya University

TOKASTAR-2 is a plasma experimental device which is able to generate toka-mak and helical magnetic field independently. One of the main objectives ofTOKASTAR-2 is finding the effect of applying helical magnetic field to tokamakplasma. In order to examine the effect, we have to analyze tokamak plasma profilelike a plasma current distribution with high accuracy. We are able to calculate aplasma current centroid and the last closed flux surface using the filament currentapproximation (FCA) method [1], but have not yet done MHD equilibrium anal-ysis by which plasma current profiles would be obtained. One of the difficulties isthe fact that a large eddy current is driven in the TOKASTAR-2 vessel and hencewe have to consider an eddy magnetic field carefully. In this paper, we reportMHD equilibrium analysis using the TOSCA code including the eddy current.TOKASTAR-2 has seven coils. Among them, three-block Ohmic Heating (OH)

coils, a pair of Pulsed Vertical Field (PVF) coils, a pair of Shape Control (SC)coils and eight Toroidal Field (TF) coils are used for tokamak operation. Pulsedcurrents are driven in these coils with capacitor banks. The toroidal field strengthis 0.1 T and pre-discharge is generated with electron cyclotron resonance heating(ECRH) matched with this toroidal field. The OH coils induce a plasma current(2.2 kA, 0.5 msec) and the PVF coils and the SC coils form the equilibrium field.The eddy current is analyzed by the TOSCA code, where the passive con-

ductors including the vacuum vessel are modeled as axisymmetric blocks. Inorder to consider the coil installation error and complex vacuum vessel shapewhich have a great effect on the magnetic field, we scanned many parame-ters in the model and decide optimal ones such that difference between thecalculated values of vacuum magnetic field and the measured values is mini-mized. In equilibrium analysis, the OH coil current and the plasma currentwere fixed to measured values while the PVF coil current are adjusted to haveequilibria with specified plasma major radius and plasma shape, which are ob-tained by FCA method. As a result, we obtained the convergence solution butwith some difference between the calculated PVF coil current and the measuredone. In the conference, we will discuss the comparison between the calcula-tion and the measurement on the poloidal magnetic field and the poloidal flux.[1]K.Yasuda et al., Plasma Fusion Res. 13 3402072 (2018)

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Inspection of Arc Trails Formed in

Stellarator/Heliotron Devices W7-X and LHD

Dogyun HWANGBO1, Shin KAJITA2,

Chandra Prakash DHARD3, Masayuki TOKITANI4,Marco KRAUSE3, Dirk NAUJOKS3,

Suguru MASUZAKI4, Noriyasu OHNO1, Soren KLOSE3,The W7-X team3

1. Graduate School of Engineering, Nagoya University, Nagoya, Japan2. Institute of Materials and Systems for Sustainability, Nagoya University,

Nagoya, Japan3. Max-Planck Institute for Plasma Physics, Greifswald, Germany4. National Institute for Fusion Science, Toki, Japan

Arcing is an important plasma material interaction issue in fusion devices. Arcingerodes wall materials as macroparticles, which possibly result in significant plasmacooling. However, because there is not enough data existing for arcing now,we cannot accurately assess material erosion by arcing in future fusion devicesincluding ITER. This paper is about the inspection of in-vessel components andobserved arc trails formed inside of Wendelstein 7-X (W7-X) and the Large HelicalDevice (LHD), which both have experienced multiple operation phases.

In-vessel inspection, which was performed in LHD using a digital camera,identified several arc trails formed on Stainless Steel (SS) first walls, ElectronCyclotron Heating (ECH) mirrors and graphite divertor tiles. The arc trails hadlinear shapes. A tungsten sample with fiberform nanostructured surfaces wasinstalled on the sample stage and exposed to the LHD plasmas. Arcs ignited andformed clear linear trails along −j×B direction indicating that the arcs ignitedin the presence of magnetic field.

After the first Operational Phase OP1.1 of W7-X, severe damages probablydue to arcings were found on few carbon Langmuir probe tips installed in graphitelimiter tiles and also on surrounding SS sleeves, where no direct plasma heat loadis probable. After the second operational phase (OP1.2), the whole plasma vesselsof W7-X were inspected using a handy digital microscope and a digital camera,including first SS wall panels, ports, graphite divertor tiles, heat shields, heatsinks of baffles made of copper and inboard-side vacuum vessel surfaces behindthe wall panels. 59 out of 197 SS wall panels were taken out of the vessel andinspected for arc trails. For the SS wall panels, no linear shape trail has beenfound, indicating that the arcing was initiated during glow discharge cleaningphase rather than the main discharge phase. Many of the arc trails formed at theedge, both on the front side and backside, seemingly bifurcated while running.

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Evaluation of Translation Velocity Control by

Auxiliary Coils for Collisional Merging Formation

of FRCs by 2-D Resistive MHD Simulation

Daichi KOBAYASHI1, Tomohiko ASAI1,Tsutomu TAKAHASHI1, Junichi SEKIGUCHI1,

Hiroshi GOTA2, Sean DETTRICK2, Yung MOK2,Michl BINDERBAUER2, Toshiki TAJIMA2,3

1. College of Science and Technology, Nihon University, Tokyo 101-8308,Japan

2. TAE Technologies, Inc., Foothill Ranch, CA 92610, USA3. Department of Physics and Astronomy, UCI, Irvine, CA 92697, USA

A collisional merging formation of field-reversed configurations (FRCs) hasbeen developed to generate high performance FRCs [1]. In this technique, twomagnetized plasmoids formed by the field-reversed theta-pinch (FRTP) methodare translated at super Alfvenic velocity toward each other so that they collideand merge into a single FRC state. A drastic increase of the excluded flux andtemperature has been observed in the merged FRC generated with this technique[2], and these phenomena strongly depend on the translation velocity of the indi-vidual plasmoids [1]. However, the dependence of the merged-FRC performanceon the translation velocity has not been studied in detail.

In this paper, the effect of auxiliary coils for controlling the translation ve-locity has been studied and evaluated by Lamy Ridge code[3]. The code canbe used to simulate the process of formation, translation and collisional merg-ing of FRCs because it consists of the two-dimensional resistive MHD equationssupplemented with an energy equation to handle non-adiabatic effects, and fluidequations coupled with the MHD equations. The auxiliary coils are added inthe FRTP formation region of the FAT-CM device. The coils are triggered whenthe formed plasmoid is near the coils during the translation, at which the peakcurrent is several tens of kA. The plasmoid is pushed out by the applied strongmagnetic field, and it is possible to control the translation velocity in the orderof several km/s.[1] H. Guo, M. Binderbauer, D. Barnes, S. Putvinski, N. Rostoker et al., Phys.Plasmas 18, 056110 (2011).[2] T. Asai, Ts. Takahashi, J. Sekiguchi, D. Kobayashi, S. Okada et al., Nucl.Fusion 59, 056024 (2019).[3] Y. Mok, D. Barnes, and S. Dettrick, Bull. Am. Phys. Soc. 55, GP9.97(2010).

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Impurity Transport of Externally

Injected Metal impurities in

the LHD Plasma with Impurity Hole

Hisamichi FUNABA1, Naoki TAMURA1,2, Chihiro SUZUKI1,Mikirou YOSHINUMA1, Katsumi IDA1, Kenji TANAKA1,Satoru SAKAKIBARA1, and LHD Experiment Group1

1. National Institute for Fusion Science, National Institutes of NaturalSciences, Toki, Japan

2. Graduate University for Advanced Studies (SOKENDAI), Toki, Japan

The impurity hole has been observed in the plasmas with high ion tempera-ture gradient in the Large Helical Device (LHD) [1-3]. In the plasmas with theimpurity hole, the carbon density measured by the charge exchange spectroscopy(CXS) shows very hollow shape. In order to observe the behavior of externally-injected impurities with higher Z than carbon, metal impurities are injected bythe laser blow-off [4] and the tracer-encapsulated solid pellet (TESPEL) [5] intosuch plasmas. These experiments also can provide an insight into the impact ofthe impurity hole on the impurities breaking into the plasma from outside.

After the titanium was injected by the laser blow-off or TESPEL, the tempo-ral development of two spectral lines of Ti XX were observed by the VUV spec-troscopy. The lines from Ti XX disappeared within almost 0.25 s in the plasmawith the impurity hole, while they remained more than 0.5 s in the plasma with-out the impurity hole. These results are compared with the calculation results byan impurity transport code, STRAHL[6], with the diffusivity, D, and the convec-tion velocity, v, of impurities. In order to reproduce the behaviour of Ti XX inthe impurity hole plasmas, outward v is required in the region of 0.6 < ρ < 0.8,where ρ is a normalized minor radius. The injection of molybdenum is also madeby TESPEL. From the comparison between the results of the laser blow-off andTESPEL, it is found that the strong outward v is not required in the peripheralregion of ρ > 0.9.

1. K. Ida, et al., Phys. Plasmas 16 056111 (2009).2. M. Yoshinuma, et al., Nucl. Fusion 49 062002 (2009).3. K. Mukai, et al., Plasma Phys. Control. Fusion 60 074005 (2018).4. E.S. Marmar, et al., Rev. Sci. Instrum. 46 1149 (1975).5. S. Sudo and N. Tamura, Rev. Sci. Instrum. 83 023503 (2012).6. R. Dux, STRAHL Manual Report IPP 10/30, Max-Planck-Institut fur

Plasmaphysik, Garching (2006).

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Application to Plasma Immersion Ion

Implantation by Inertial Electrostatic

Confinement (IEC ) Device

Hiroshi OKAWA 1, Koki NARIMATSU 1,Tomoki KANEKO1, Daigo KASHIWABARA1,

Shiki SATO1, Yawara SHIINA1, Tetsuya AKITSU2

1. Happy Science University2. University of Yamanashi

The inertial electrostatic confinement (IEC) fusion device is known as a compactneutron generator. In this study, we developed a linear IEC device which consistsof glass tube section separating ground and high voltage walls. First of all, weconfirmed the presence of high energy hydrogen ions and neutral particles fromBalmer-α emission spectra Doppler shift. Subsequently, we observed a relation-ship between the cathode voltage and the nitrogen implantation, in the plasmaimmersion ion implantation in the IEC device, on a stainless steel (SUS304) targetthat had a strong oxide film on the surface. Ion nitriding in plasma immersion ionimplantation is a promising glow discharge surface modification technique, whichcan be primarily used to increase the fatigue strength, wear, corrosion resistanceand surface hardness of steels.

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Void-free Fuel Solidification in a Foam Shell

FIERX Target

Akifumi Iwamoto1, Takeshi Fujimura2,Takayoshi Norimatsu2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. Institute of Laser Engineering, Osaka University

We study fuel layering for Fast Ignition Realization Experiment (FIREX) cryo-genic targets. One of the strategies is a foam shell method. The challenge onfoam shell is that residual voids should be reduced by less than 1 percent in thesolid fuel impregnated within the foam. We have demonstrated the residual voidfraction of 1 percent in a foam wedge. ANSYS simulations have shown that thetechnique will be applicable to a FIREX target. We examined each step in thesimulated process using a dummy foam shell target and succeeded in forming anice layer with a reduced void fraction.

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Renewal of large current dc power system for

fusion plasma experiment facility LHD

H. CHIKARAISHI1, S. TAKAMI1, TANOUE1,H, MIZUNO2, K. TANAKA3

1. National Institute for Fusion Science2. Aichi Electrix co., ltd.3. Emertech. co., ltd.

The Large Helical Device (LHD) is a fusion plasma experiment facility and ithas been operated from 1999. This facility has six superconducting coil set tomake magnetic field to confine the plasma and it has six dc power supplies toexcite these coils. To control these power supplies, a computer system was builtat the same time and used. The base system that are steady state power suppliesand main part of the computer system were constructed from 1994 to 1995. In2007, additional high voltage power supplies are installed for two power suppliesto enhance the current response[?], and the motor driven polarity exchange gearswere built in 2009. A new control system, high voltage equipments and capacitorbank of dc filter were constructed and started of operation from 2017. In thispaper, the renewal of devices and the operation result is introduced.

To build and install a new components, there is a limitation that the LHDoperation cannot be interrupt. With this condition, we fix a concept of replace-ment as follows. For the computer system, we construct a new computer systemwithin 3 years and replace old system with new one in last year end. The othercomponents such as high voltage equipment or local controller is designed that ithas compatible interface with old devices, and they are replaced year by year.

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Radiation Resistant Camera System For

Monitoring Deuterium Plasma Discharge In The

Large Helical Device

Mamoru SHOJI1, LHD Experiment Group1

1. National Institute for Fusion Science

The first deuterium (D-D) experimental campaign in the Large Helical Device(LHD) has begun on March 2017, and the third D-D campaign will be startedfrom October in this year (2019). During the experimental campaign, neutronswith an energy of 2.45 MeV and 14.1 MeV are yielded by D-D and D-T (tri-tium) reactions in the LHD plasma, and gamma-rays are emitted by neutroncapture in mechanical structures in the device and the experimental hall. It hasbeen found that CCD cameras, which have been effectively applied for monitoringLHD plasms, are subject to suffer from the radiation (neutrons and gamma-rays).A radiation resistant camera system was newly developed for monitoring the plas-mas in the D-D campaigns. Radiation shield boxes were installed at positionsbeing away from the LHD torus on diagnostic stages in the experimental hall.The cameras are protected by the shield boxes consisting of 10 percent boratedpolyethylene blocks (100 mm in thickness) and lead plates (15 mm in thickness).The images of the plasma in the vacuum vessel are transferred to commercial un-cooled CCD cameras installed in the shield boxes via radiation resistant bundledfibers (having 30 thousand pixels) with a length of about 20 meters. The camerasystem was successfully operated under a plasma discharge condition with themaximum neutron emission rate. Although no serious functional and electricalproblems have occurred in the experimental campaigns owing to the radiationshield box, it has been found that the number of permanent (unrecoverable)specks of images on the CCD sensor, which are pixels damaged by the radia-tion, increases with the accumulated neutron yield. The relationship betweenthe number of the specks on the censor and the fluence of the radiation on theCCD cameras in the shield box is investigated using the MCNP Monte Carloneutronics code (MCNP-6) with a detailed three-dimensional model of LHD andthe experimental hall for this code. In future nuclear fusion reactors, it is nec-essary to use the camera system appropriately protected with the shield box formonitoring the plasmas, the vacuum vessel and the torus hall, etc. In this pa-per, the temporal transition of the observations of the plasma images in D-Dexperimental campaigns are presented. In addition, the first investigation on theinfluence of the radiation on the shielded cameras in the nuclear fusion device(D-D experiments in LHD) is reported. The results will contribute to finding anoptimal application of the shielded CCD camera systems to monitoring nuclearfusion electric power plants in the future.

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Comparative Studies on the Control Algorithm

for the High-density Ignition Regime in FFHR-d1

Osamu MITARAI1, Shota SIGIYAMA2,Nagato YANAGI3, Yoshiro NARUSHIMA3,Ryuichi SAKAMOTO3, Takuya GOTO3,Hideaki MATSUURA2, Akio SAGARA3

1. Institute for Advanced Fusion and Physics Education, Kumamoto, Japan2. Department of Applied Quantum Physics and Nuclear Engineering, Kyushu

University, Fukuoka , Japan3. National Institute for Fusion Science, Toki, Japan

In the helical reactor FFHR-d1 (R = 15.6m, a = 2.5m, Bo = 4.5T, ⟨β⟩ ∼ 5% andfusion power of 3 GW), the high density and low temperature operation regime(n(0) ∼ 9 × 1020m−3 and T (0) ∼ 7 keV) is attractive because alpha particleloss would be reduced and the bootstrap current is expected to be low, fuelingby pellet penetration would be improved than the high temperature operationregime, and the divertor heat load is reduced owing to the better confinement andlarger bremsstrahlung radiation. Although the practical control method proposedfor the high density and low temperature regime can be achieved by fueling,one possible drawback is its control failure due to the malfunction of the pelletinjection system. Because as it is operated in the thermally unstable regime, thethermal runaway would take place when the fueling control is failed.

On the other hand, impurity injection method could be used for the low tem-perature and high density operation. As this method works in the thermallystable regime, a failure of the impurity injection system does not cause any trou-ble. However, by comparing these two operational control algorithms we haveconcluded that fueling control in the thermally unstable regime is much betterfor ignited operation due to the wider operation regime, such as low temperaturedown to 6.8 keV.

Furthermore, the thermal runaway behavior is analyzed when the controlfailure takes place during fueling. When the thermal runaway takes place, thefusion power and the beta value are increased to ∼15 GW and ∼9.2 % in a shorttime, which may be resulted in the disruptive nature. However, the plasma shiftsby the sudden increase in the beta value during the thermal runaway, and thesubsequent plasma volume reduction cause the confinement degradation, limitingthe fusion power surge. We thus show that the thermally unstable operation hasan inherently safe function, which should be experimentally confirmed. We alsoshow that the alpha particle loss would be reduced when the operation density isvery high due to the shorter slowing down time of the alpha particles in FFHR-d1.

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Study on System Dynamics of Fuel and Burning

in DD Start-up Scenario of a Fusion Reactor

Kento MIYAMAE1, Hiroshi YAMADA1,2,Ryuta KASADA3, Satoshi KONISHI4

1. Graduate School of Frontier Science, The University of Tokyo2. National Institute for Fusion Science, NIFS3. Institute for Materials Research, Tohoku University4. Institute of Advanced Energy, Kyoto University

The start-up of a fusion reactor without initial tritium loading has been as-sessed with detailed modelling of a burning plasma. System dynamics of DT fuelsand He ash in a fusion reactor plant has been discussed in order to identify theimpact of dilution due to He ash and recycling of DT fuels.

Initial loading of tritium in a fusion reactor is a critical issue because of avail-ability of tritium. The natural abundance of tritium is almost zero, and resourceis limited to several tens of kilogram worldwide. Therefore, the start-up onlyfrom deuterium (DD start-up) has been attracting interests. While the earlierresearch of this DD start-up scenario [1] showed its technical feasibility, advance-ment of the model is inevitable for quantitative assessment of the operationalscenario. Temporal evolution of temperature of a fusion plasma incorporating avariety of temperature dependence of physical parameters, profile of density andtemperature, and dilution effect have been involved in the present study.

It should be noted that plasma temperature changes during the start-up phasedue to the dependence of energy confinement on the fusion power heating as wellas the isotope effect by the build-up tritium concentration. Also, it is predictedthat temperature and density profiles greatly affect the fusion power, and dilutionof fuel decreases the fusion power. In the present model, evolution of temperatureconsistent with power balance including radiation losses and an empirical scalingof an energy confinement time has been integrated with updated plasma model.

Operational parameters are based upon the recent tokamak fusion DEMOdesign by the Joint Special Design Team [2]. It takes around 50 days to reach thesteady state with D/T=50/50. He ash concentration increases from 4.7% to 7.3%as the He particle confinement time changes from 10 s to 20 s. Consequently, thefusion power in steady state degrades from 1.21 GW to 0.94 GW. Build-up oftritium inventory in a plant is also discussed.

This work is supported by the National Institute for Fusion Science grantadministrative budgets NIFS18KLPP051 and JSPS KAKENHI Grant NumbersJP17H01368.

[1] R.Kasada, et al., Fusion Eng. Design 98-99 (2015) 1804.[2] Y.Sakamoto, et al., Fusion Eng. Design 89 (2014) 2440.

P1-30

Simulation of Deuterium Retention in Tungsten

by Repetitive Irradiation with High-Flux

Deuterium Ions

Kaoru OHYA1, Makoto OYA2

1. Professor Emeritus of Tokushima University2. Faculty of Engineering Science, Kyushu University

To study the effect of irradiation cycles on hydrogen isotope retention in plasmafacing walls, thermal processes of implanted low-energy (50 eV) and high-flux( 1022 m−2s−1) deuterium (D) ions were simulated with repetitive pulsed irradi-ation of tungsten. The irradiation was repeated with one and more breaks untilthe total fluence reached 1026 D m−2 in order to compare with continuous irradi-ation for 10,000 s with no break. In the simulation, the diffusion, the trappingand detrapping in the target and the recombination at the surface were takeninto account. The implantation profile was inputted at intervals of 1 s by simu-lating collision processes of the implanted D [1]. The time evolution of the targettemperature during the irradiation and the breaks was taken from a high-flux Dirradiation experiment in the same fluence condition [2].

In spite of low-energy implantation within the depth of 10 nm, the implantedD distributed over 10 µm owing to high-density of solute D atoms and deep Dtrapping by existing trap sites. The breaks between the irradiation decrease thesolute D density. Total D retention dominated by trapped D atoms were muchless influenced by the irradiation cycles than expected from thermal desorptionspectra observed by the experiment [2]. In order to evaluate irradiation damageeffects that produced additional trap sites in a shallow region, thermal desorptionspectra were simulated in succession to the high-fluence D irradiation. By fittingthe calculated spectra to the observation in the experiment, the detrapping en-ergies and the densities of the additional trap sites are estimated as a function ofirradiation cycles. The trap concentration profiles are also studied.

[1] K. Ohya et al., Plasma Fusion Res. 7 (2012) 2403083.[2] M. Oya et al., Nucl. Mater. Energy 12 (2017) 674.

P1-31

Development of a bipolar power supply for

Ohmic heating on QUEST spherical tokamak

Yifan Zhang1, Takumi Onchi2, Kazuo Nakamura2,

Makoto Hasegawa2, Ryuya Ikezoe2, Hiroshi Idei2

1. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity

2. Research Institute for Applied Mechanics, Kyushu University

In QUEST spherical tokamak, high plasma density is required for heating andcurrent drive with electron Bernstein wave. Start-up of high plasma current is aneffective approach to generate such overdense state. A new Ohmic heating (OH)system consisting of two high current sources, positive supply (PS) and negativesupply (NS) is designed to drive high plasma current Ip. IGBT stacks installedbetween the current sources and Central Solenoid (CS) can work as the switchesto change the polarity of OH primary current.

There are two steps mainly for operation of the OH system. In the first stagethat PS power supply and IGBT stack 1 with positive circuit are turned on, CScurrent Ics reaches 8 kA and then decreases to 0 A. The plasma current Ip canincrease and reach about 90 kA. In the second stage, IGBT stack 1 is turned offto stop the PS current. Then, NS and IGBT stack 2 are turned on and drive Ipfurther more. This proposed bipolar OH system with the double flux swing inthe CS coil is a promising method to achieve high Ip, which is expected to be180 kA, in the ideal plasma circuit. Each IGBT stack consists of 2x2 array toeliminate the risk of overcurrent and overheating in IGBT stacks. The current isevenly distributed by using share resistors between each IGBT and bus.

Safety must be guaranteed for the upgraded OH system. The situation thatboth PS and NS are on-state simultaneously is prohibited when the direction ofIcs changes. An interlock system controlled through FPGA is designed.

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Temperature dependence on co-deposition layer

formed by sputtered tungsten and helium plasma

K. ASAI1, N. Yoshida2, N. Ohno1, S. Kajita3,H. Tanaka1, M. Yajima4

1. Grad. School of Eng., Nagoya University2. RIAM, Kyushu University3. IMaSS, Nagoya University4. National Institute for Fusion Science

Tungsten (W) is one of the candidates of plasma-facing materials (PFM) in fusionreactor, and its surface structure and properties are modified by helium (He) ionbombardment [1]. Sputtered W and plasma particles can form co-depositionlayers on the PFM surface, which are expected to have an effect on gas retentionproperty of the PFM. Previous work investigating the retention used bulk W[2] and W deposition layer formed by physical vapor deposition (PVD) [3]. Inthis study, we have formed W deposition layer with He plasma and investigatedsample temperature dependence on its characteristics.

W samples were exposed to He and W mixture plasma in the linear plasmadevice CO-NAGDIS. He plasma was generated by a DC discharge with a direct-current heated LaB6 cathode and a hollow cupper anode. A W wire was installed5 mm in front of the sample and, was negatively biased to provide sputtered Watoms into the plasma to form a deposition layer on the sample.

He-W co-deposition layer formed at 150˜500℃ had He bubbles and consistedof small grains with a size of a few tens of nm. Surface and cross-section imagesusing SEM and TEM showed that He bubble size became bigger, and blister, crackand void were observed more frequently as the sample temperature increased.Regarding TDS measurement, He-W co-deposition layer had a unique retentioncapability of He compared to previous study using He irradiated W [4]. Althoughthe He irradiated W showed He desorption peak at around 1000℃, the He-Wco-deposition layer had only the peak at low temperature (< 600℃). Porousstructure of co-deposition layer due to He bubble, blister, crack, and void may beable to make the invading of He easier to release. The He-W co-deposition layerhas possibility to determine the hydrogen isotope retention behavior in futurereactor.

[1] S. Takamura et al., Plasma Fusion Res. 1 (2006) 051.[2] M. Miyamoto et al., Journal of Nuclear Materials 463 (2015) 333-336.[3] G. De Temmerman et al., Journal of Nuclear Materials 389 (2009) 479-483.[4] T. Hino et al., Journal of Nuclear Materials 266-269 (1999) 538-541.

P1-33

ENGINEERING ANALYSIS FOR VACUUM

VESSEL OF CFQS QUASI-AXISYMMETRIC

STELLARATOR

Sho Nakagawa1, Takanori Murase1, Akihiro Shimizu1,

Shigeyoshi Kinoshita1, Mitsutaka Isobe1,3,Shoichi Okamura1, Yuhong Xu2, Haifeng Liu2,

Guozhen Xiong2, Hai Liu2, Dapeng Yin4, Yi Wan4,and CFQS team1,2,3,4

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences, Toki 509-5292, Japan

2. Institute of Fusion Science, School of Physical Science and Technology,Southwest Jiaotong University, Chengdu 610031, People’s Republic of China

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

4. Hefei Keye Electro Physical Equipment Manufacturing Co., Ltd, Hefei230000, People’s Republic of China

The world’s first quasi-axisymmetric stellarator CFQS is constructed as the jointproject of National Institute for Fusion Science (NIFS) in Japan and SouthwestJiaotong University (SWJTU) in China [1]. Physics design of CFQS plasma wascompleted[2,3], and numerous efforts are now being made to finalize engineeringdesign of CFQS.

Compared to tokamak fusion devices, the CFQS vacuum vessel is fairly com-plicated in shape because it is designed so as to follow the shape of the plasma[4].The vacuum vessel will be made of a thin plate of SUS316 having thickness of6 mm and needs to be capable of withstanding an atmospheric pressure, bak-ing temperature, and gravitational force. Therefore, the structural reliability ofvacuum vessel under these three loads is of our great concern.

To evaluate the mechanical behavior of the vacuum vessel, the finite elementmethod software ANSYS is employed in this work. We consider three loadsabove mentioned. Each effect can be evaluated independently. A fine mesh isadequately chosen to obtain sufficiently accurate results about stress distributionand deformation for each case. Analyses indicate that the maximum stress anddeformation on the vacuum vessel are 126 MPa and 3 mm, respectively even if weconsider all three loads simultaneously, and these are within an acceptable range.

Reference:[1] M. Isobe et al., Plasma Fusion Res. 14 (2019) 3402074.[2] H. Liu et al., Plasma Fusion Res. 13 (2018) 3405067.

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[3] A. Shimizu et al., Plasma Fusion Res. 13 (2018) 3403123.[4] S. Kinoshita et al., Plasma Fusion Res. 14 (2019) 3405097.

P1-34

Influence of neutron irradiation on LHD

Thomson scattering system

Ichihiro Yamada1, Hisamichi Funaba1,Kazumichi Narihara1, Jong-ha Lee2, Yuan Huang3,

Chunhua Liu3

1. National Institute for Fusion Science, Toki 509-5292, Japan2. National Fusion Research Institute, Daejeon 34133, Korea3. Southwestern Institute of Physics, P.O. Box 432, Chengdu, 610041, China

In LHD, deuterium plasma experiments (DD experiments) have been carried outsince 2017. In the DD experiments, neutrons with specific energy of 2.45 MeV aregenerated through deuterium thermonuclear fusion reactions, and the influenceon the devices located near LHD are concerned.

Concerning the LHD Thomson scattering system, a view window, light col-lection mirror and optical fibers have been installed near LHD. If the changes oftransmittances of the window and fiber and reflectance of the mirror occurs, theymay cause serious problem in the results from the Thomson scattering diagnostic,electron temperature and density.

The view window and the optical fiber are made of fused quartz and syntheticquartz, respectively, and the light collection mirror is gold coated. We carefullycalibrated the window transmittance, the fiber transmittance and the mirrorreflectance, and compared them to those measured prior to DD experiments. Ofthe three components, the optical fiber is considered to be most affected becausethe light transmission length is much longer (46 m) than the thickness of thewindow (5 cm). Therefore, we developed a transmission monitor system for theoptical fiber, and measured the transmittance during LHD DD experiments. Wewill discuss the neutron influences on the LHD Thomson scattering diagnosticsat the conference.

P1-35

The investigation of pretreatment methods for

liquid scintillation measurement of environmental

water-samples using ion exchange resin

Shunya NAKASONE1, Sumi YOKOYAMA2,

Tomoyuki TAKAHASHI3, Masakazu OTA4,Hideki KAKIUCHI5, Shinji SUGIHARA6,

Shigekazu HIRAO7, Noriyuki MOMOSHIMA8,Toshiya TAMARI8, Nagayoshi SHIMA8,

Mariko ATARASHI-ANDOH4, Satoshi FUKUTANI3,Akinobu ISHIMINE1, Masahide FURUKAWA1,Masahiro TANAKA9,10, Naofumi AKATA11

1. University of the Ryukyus2. Fujita Health University3. Kyoto University4. Japan Atomic Energy Agency5. Institute for Environmental Sciences6. Kyushu University7. Fukushima University8. Kyushu Environmental Evaluation Association9. National Institute for Fusion Science10. SOKENDAI (The Graduate University for Advanced Studies)11. Hirosaki University

To reduce chemical quenching for high precision analysis in the LSC methods, wemust remove the impurities from the water samples. The treatment method of or-dinary pressure distillation removes the impurities such as organic matter and theion. In this study, we report an examination result of the tritium analysis acceler-ation by the ion exchange resin method (batch method and the column method).In this study, we suggest the sample treatment method by the ion exchange resinto pretreatment plural samples easily and inspect it about the application possi-bility. In addition, we performed a comparison between the column method andthe batch method experiment. In this study, we assumed rainfall of Hokkaido andused a simulation sample (1.09 ± 0.07 Bq/L) which diluted D2O standard sample(103.24 ± 0.13 Bq/L) 100 times in a rainfall sample. The sample was mixed withcation- and anion-exchange resin, activated carbon. The ion exchange resin toremove ion used Eichrom resin and Powdex resin. In addition, we added an acti-vated carbon to remove organic matter of the sample water. In other hands, The

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column method used solid-phase extraction.The sample was mixed with cation-and anion- exchange resin, activated carbon, 0.2 g and 0.2 g, 0.4 g, respectively inthe batch method. From the result of an electric conductivity and pH, we couldconfirm the removal of impurities of the sample water and confirmed that theremoval of impurities was possible for a short time in (30 minutes). In addition,we measured the absorption spectrum of the ultraviolet domain using an absorp-tiometer (ASUV-3100PC) to confirm whether organic matter was included afterpretreatment. Absorption was accepted to around 300nm in all sample water.

P1-36

DESIGN OF POWER SUPPLY SYSTEM FOR

MAGNET SYSTEMS OF THAILAND

TOKAMAK -1

Nopporn POOLYARAT1,

Suebsak SUKSAENGPANOMRUNG2,Wutthichok SANGWANG1

1. Center of Advanced Nuclear Technology, Thailand Institute of NuclearTechnology

2. Electrical Engineering Department, King Mongut’s University of Technol-ogy Thonburi

Thailand Tokamak -1 is a small size tokamak, formerly known as HT-6M. SinceInstitute of Plasma Physics, Chinese Academy of Science (ASIPP, China) andThailand Institute of Nuclear Technology (TINT, Thailand) agreed and signed aMemorandum of Understanding for scientific cooperation to help Thailand on fu-sion technology. Under this MOU, ASIPP has donated the HT-6M to TINT. TheHT-6M will undergo a reconstruction on many supporting system, i.e. magnetpower supply system, vacuum system, data acquisition system, etc., and will benamed TT-1 after finished. For the power supply for magnet system, new designtogether with new technology will be used and for the reconstruction. Simulationson circuit design will also be employed for the optimized operating condition ofthe first phase to TT-1.

P1-37

Numerical Study of Supersonic Molecular Beam

Injection System in Future Thailand Tokamak

Jiraporn PROMPING1, Apiwat WISITSORASAK2,

Boonyarit CHATTHONG3, Kewalee NILGUMHANG 1,Dusit NGAMRUNGROJ4, Wiwat WONGKOKUA5

1. Thailand Institute of Nuclear Technology (Public Organization), Bangkok,Thailand

2. Department of Physics, Faculty of Science, King Mongkut’s University ofTechnology Thonburi, Bangkok, Thailand

3. Department of Physics, Faculty of Science, Prince of Songkla University,Songkhla, Thailand

4. King Mongkut’s University of Technology North Bangkok, Bangkok, Thai-land

5. Department of Physics, Faculty of Science, Kasetsart University, Bangkok,Thailand

Thailand Institute of Nuclear Technology (TINT) is planning to develop ThailandTokamak from HT-6M tokamak. In the first phase of operation, the device will beequipped with two fuelling systems: gas puffing (GP) and supersonic molecularbeam injection (SMBI). Since the SMBI system has never been used with thisdevice, this work, therefore, numerical studies the effect of the SMBI on theplasma of Thailand Tokamak. The interaction of the supersonic molecule withthe background plasma has been computed by the transport simulation codeTASK/TR that is coupled with a six-field fluid model of SMBI. The preliminaryresult demonstrated that the SMBI is an efficient method for fuelling the plasmaand the beam can be delivered to the center of the core plasma.

P1-38

Design of VFT and Multipactor Test Chamber

for High Power Helicon Current Drive System in

KSTAR

Kwangho Jang1, Sonjong Wang1, Hyunho Wi1,

Kenji Saito2, Hyungyong Lee1, Jeehyun Kim1,Jonggu Kwak1

1. National Fusion Research Institute (NFRI)2. National Institute for Fusion Science (NIFS)

A helicon wave was confirmed to be coupled at low power and a high powercoupling has been tried in KSTAR. When the RF is applied to the antenna sys-tem, the reflected power gradually increases by the timescale of sub-milliseconds.Rather slower process than usual arcing suggests that the reflection is caused by amulti-pactor discharge. In order to apply a high power helicon wave to the toka-mak plasmas, it is important to mitigate or eliminate the multi-pactor discharge.A new VFT is fabricated following the design which focuses on the reduced totalRF electric field and zero axial electric field on the TiN coated alumina surface.The disc type alumina window does not protrude in to the coaxial conductors soas to eliminate axial electric field. The unmatched impedance caused by the 1 cmthick alumina is compensated by the series matching element positioned in thepressurized section. For further understanding multi-pactor discharge at the he-licon antenna system the multi-pactor test chamber (MTC) was developed. Thehigh electric field is generated by the series unmaching-matching structures at theboth sides of sample holder. The MTC is mainly used to evaluate muti-pactoravoiding technique and to develop conditioning processes.

P1-39

Propagation of lower-hybrid surface waves in

semibounded warm plasmas

Myoung-Jae LEE1, Young-Dae JUNG2

1. Department of Physics, Hanyang University, Seoul 04763, South Korea2. Department of Applied Physics, Hanyang University, Ansan, Kyunggi-Do

15588, South Korea

We have derived the dispersion relation for the lower-hybrid surface waves prop-agating perpendicular to the magnetic field in the semi-bounded warm plasma.The mirror reflection boundary condition is adopted to obtain the surface modeof lower-hybrid plasma waves. The effects of magnetic field strength and thefinite ion temperature on the propagation of the lower-hybrid surface wave areinvestigated. The increase of ion temperature significantly increases the wave fre-quency, but an interesting hump of the group velocity appears in the region wherethe wavelength of the lower-hybrid surface wave is much larger than the electronDebye length. For cold ions, the surface wave is resonant near the lower-hybridfrequency as the wave number goes to an infinity.

P1-40

Theoretical investigation of E1 transitions for Pr

II - Gd II ions

Gediminas GAIGALAS1, Laima RADZIUTE1,Pavel RYNKUN1, Daiji KATO2,3, Masaomi TANAKA4

1. Institute of Theoretical Physics and Astronomy, Vilnius University, Lithua-nia

2. National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Japan3. Department of Advanced Energy Engineering Science, Kyushu University,

Fukuoka, Japan4. Astronomical Institute, Tohoku University, Aoba, Sendai, Japan

Observations of a kilonova associated with GW170817 provide a unique opportu-nity to study heavy element synthesis in the universe. However, the atomic dataof r-process elements are not yet complete enough to decipher the light curvesand spectral features of kilonova. We performed extended atomic calculations forsingly ionized elements with Z = 59 − 64, by employing GRASP2018 package[1] and Nd II computation strategies [2]. This research was based on multicon-figuration Dirac-Hartree-Fock and configuration interaction (RCI) methods [3].Energy levels were computed for states of the configurations:

[Xe]4fN{5d6s, 5d6p, 6s6p, 5d2, 4fN6s, 4fN6p, 4fN5d}, where N = 3−8 for Pr-Gd ions, respectively. In addition, calculations of configuration states [Xe]4fN+1

were performed for Sm II and Eu II, whereas for Gd II configuration states[Xe]4fN−16s2 were added.

The configuration space was increased step by step with increasing the numberof layers. The orbitals of previous layers were held fixed and only the orbitals ofthe newest layer were allowed to vary. Sets of virtual orbital were generated by sin-gle (S), double (SD) or triple (SDT) substitutions. SDT substitutions were usedto generate active space for configuration [Xe]4fN+1, SD for [Xe]4fN{6s, 5d, 6p}and S for [Xe]4fN−1{5d6s, 5d6p, 6s6p} and [Xe]4fN−1{6s2, 5d2}. The largest ac-tive space used in the computations was {8s, 8p, 7d, 6f, 5g}. Substitutions weredone only from valence orbitals. Each configuration was computed separately.

The analysis of the results shows that oscillator strength of E1 transitions issuitable for opacity computation of neutron stars merger. This includes data for411 314, 2 955 492, 7 104 005, 4 720 626, 467 724 and 1 383 694 numbers oftransitions for Pr II, Nd II, Pm II, Sm II, Eu II, and Gd II, respectively.

Acknowledgments: this research was funded by a grant (No. S-LJB-18-1) fromthe Research Council of Lithuania.

[1] Froese Fischer C et al 2019 Comput. Phys. Comm. 237 184[2] Gaigalas G et al 2019 ApJS 240 29[3] Froese Fischer C et al 2016 J. Phys. B 49 182004

P1-41

Study of the plasma response on external RMP

field in a small tokamak device HYBTOK-II

Yoshihide SHIBATA1, Masaaki OKAMOTO2,Kiyomasa WATANABE3, Noriyasu OHNO4,

Yusuke KIKUCHI5

1. National Institute of Technology, Gifu college2. National Institute of Technology, Ishikawa college3. National Institute for Fusion Science4. Nagoya University5. University of Hyogo

Investigations of a plasma response on error fields and resonant magnetic pertur-bations (RMPs) by using external coils are important in teams of plasma con-finement. However, mechanisms of amplification and shielding of RMPs arounda resonant surface has not clarified yet. It is important to understand thesemechanisms in detail by using an experimental data. In terms of this, study ofa magnetic shielding has been carried out in many tokamak and helical devices.Decrease of poloidal flow by a magnetic island was observed in LHD and a gen-eration of magnetic island changes poloidal flow [1]. Different plasma responseof radial magnetic field profile was observed by using Rotating Helical MagneticPerturbation (RHMP) coils and stable plasma in HYBTOK-II [2,3]. In electrondiamagnetic rotation of RMP, plasma response was amplification. On the otherhand, plasma response shows shielding in rotating direction of ion diamagnetic.In a previous study in HYBTOK-II, it was also found that the relative velocitybetween EXB velocity at resonant surface and RHMP was important for Br am-plification and shielding phenomena. From former experimental results, plasmaresponse such as B

rprofile has strong dependence of EXB direction at resonant

surface. However, mechanisms of Amplification and shielding of RMPs around aresonant surface has not clarified yet. In this study, we carried out the experi-ment of RHMP coils in HYBTOK-II by using Mach and Br probe arrays with ahigh spatial resolution to investigate the relationship between poloidal flow andplasma response. [1] K. Ida, et. al., Phys Rev Lett. 88(2001) 015002. [2] S.Takamura, et. al., Nucl. Fusion. 43(2003) 393. [3] Y. Kikuchi, et. al., Journal.Nucl. Matt. 313(2003) 1272.

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Experimental studies on ion-ion separation and

their collection for direct energy conversion

K. Hashiguchi1, K. Ichimura1, S. Nakamoto1,

H. Takeno1, J. Miyazawa2, T. Goto2

1. Graduate School of Engineering, Kobe University2. National Institute for Fusion Science

　D-3He fusion is expected as an advanced power generation scheme. The plasmaflowing out of D-3He fusion reactor contains various charged particles such aselectrons, thermal ions, and high energy protons. Since these particles differ incharge polarity and energy, separation of them is necessary to apply direct energyconversion. Momota et al. proposed CuspDEC device for the separation [1], andvarious studies on ion-electron separation have been done. However, there arefew studies on high energy proton and thermal ion (ion-ion) separation.

The authors proposed a new scheme of ion-ion separation. A pair of plane elec-trodes consisting of grounded and biased ones in the upstream and downstream,respectively, are settled at the bottom of CuspDEC. Both electrodes have a hole,and low energy ions incident into the region are reflected by the field leaking fromthe hole while high energy ions can pass through there, thus ion-ion separationcan be achieved.

In this research, an experimental study for the new scheme is performed. As apreparation study, a combined plasma source was constructed [2]. A high energyion source and a thermal plasma source are connected in series to simulate theflowing out plasma from D-3He fusion reactor. Particle orbit calculation due tothe new separation scheme was also performed [3]. In the condition of the existingCuspDEC device, the distance between electrodes and the size of the holes wereexamined for the electrode design in the new experiment. The present researchis based on those results, and the detail will be presented in the conference.

This work was supported in part by a Grant-in-Aid for Scientific Research(B) (16H04317) from JSPS.

References[1] H. Momota et al., Proc. 7th Int. Conf. on Emerging Nucl. Energy Systems,16 (1993).[2] Y. Okamoto et al., 12th Int. Conf. on Open Magnetic Systems for PlasmaConfinement, P16 (2018).[3] Y. Okamoto et al., 27th Int. Toki Conf., P1-48 (2018).

P1-43

A Study on the Detection of Plasma Generated

Radicals in the Degradation of Dibromophenols

in Water Solution

Tetsuya AKITSU1, Shin-Ichiro KOJIMA2,

Keiko Katayama-Hirayama1, Hiroshi OKAWA3

1. University of Yamanashi, 4 Takeda, Kofu, Yamanashi, 400-8511, Japan2. Kyushu University, FUkuoka, Fukuoka 812-8581, Japan3. Happy Science University, 4427-1 Hitotsumatsu-Hei, Chosei, Chiba 299-

4325, Japan

In the previous meeting, we reported a dielectric barrier discharge operated inthe liquid-gas boundary, and applied to the degradation of Dibromophenol. Theadvanced oxidation process by charged species, anion as well as neutral radicalsare presumed. In the interpretation of the observed difference in the resistance,we also presume these radicals. The originality in this session is the detection ofthe generated radical using the excitation-and-light emission by Xe-Cl excimerlamp.

P1-44

Study of reconstruction of microwave

reflectometry image by machine learning

Hayato TSUCHIYA1, Ryo MANABE2,

Ryota TAKENAKA2, Shuji YAMAMOTO3,Naofumi IWAMA1, Mayuko KOGA2,

Soichiro YAMAGUCHI3

1. National Institute for Fusion Science2. University of Hyogo3. Kansai University

The imaging diagnostics for plasma are actively studied in recent years. Thevisible light measurement such as gas puff imaging has the advantage of easy-implementing using commercial CCD cameras or its fast camera. However, asis well known, the reconstruction with rigid spatial resolution is not easy be-cause the obtained image includes the line integral effect. On other hands, themicrowave/millimeter wave diagnostic such as reflectometry has advantage ofeasy-identify the view plane. According to the wave length of microwave, thespatial resolution and number of imaging antennas are limited in general.

In the microwave diagnostics, it is general to obtain the complex amplitude,which can be derived from measured power and phase by an antenna. We, mi-crowave imaging team of NIFS, developed the microwave camera (HMID) whichis 2D-horn antenna array type imaging devise, and operates for plasma experi-ment. To apply the limitation of the number of imaging antenna, recently, we arestudying the reconstruction method from measured complex amplitude withoutfocus optics systems[1].

As one of the reconstruction method, the reconstruction by machine learningis reported in this conference. In general, because the 2D distribution of complexamplitude of reflected microwave from distorted cutoff surface corresponds to theshape of reflection surface, the information of cutoff surface can be reconstructedby proper regularization such as Tikhonov regularization even if the number ofimaging antenna (i.e. number of measured data) is smaller than that of pixel ofreconstructed image. However, because such regularization method requires a lotof computing time for one image, the movie (i.e. animation) analysis would notbe practical. Machine learning method does not need a lot of computing time forreconstruction once it spends a lot of time for learning. The issue such as whatkind of teacher data is appropriate to learning are discussed at the conference.

[1] H.Tsuchiya, et al. Plasma Fusion Res. (2019), To be published.

P1-45

Detection of Non-thermal Electrons in LHD

Plasmas via Fe-line Spectroscopy

Tomoko KAWATE1, Motoshi GOTO2,Izumi MURAKAMI2, Tetsuya WATANABE3

1. Institute of Space and Astronautical Science, Japan Aerospace ExplorationAgency

2. National Institute for Fusion Science, National Institutes of Natural Sci-ences

3. National Astronomical Observatory of Japan, National Institutes of Natu-ral Sciences

This research aims at deriving the electron energy distribution function in theplasmas of the LHD (Large Helical Device) by using soft x-ray Fe lines. We fo-cus on intensity ratios among the dielectronic satellite line (j) from Li-like ions,resonance line (w) and M1 transition line (z) of He-like ions, inner-shell excita-tion line (q) of Li-like ions. Intensity ratios of these lines depend on the electrontemperature, so that they provide information of the electron energy distributionfunction.

We observed Fe+24 lines around 1.85 A emitted from the LHD plasmas byusing the x-ray crystal spectrometer (Morita & Goto 2003). The plasmas wereheated by electron cyclotron heating, and the soft x-ray Fe lines were observedeven without iron pellet injection. The Fe ions are thought to originate in thevacuum vessel wall of LHD, which is made of stainless steel. Based on the electrontemperature and density profiles measured by the Thomson scattering method,we simulate an integrated spectrum over the line of sight under the thermalequilibrium approximation where CHIANTI v8 (Dere et al. 1997, Young et al.2016) is used, and compare the results with the observed ones. As a result,the observed intensity ratio of j and w is consistent with the simulated result,and the typical temperature is evaluated to be 4 keV. On the other hand, theobserved intensity ratio of z and w amounts to 70% of the simulated result, andthe typical temperature is evaluated to be 7 keV. If we assume the electron energydistribution function contains two Maxwell distributions and the temperature ofone component is 10 keV, the population of 10 keV electrons amount to 4–8% ofthe all electrons.

P1-46

High density plasma generation in a thruster

using a helical antenna with variable axial length

Yoshiki MARUKI1, Keisuke KONUMA1,Satoshi NAKAMOTO1, Kazuya ICHIMURA1,

Hiromasa TAKENO1

1. Graduate School of Engineering, Kobe University

Recently, various types of space exploration are performed so a propulsionsystem with independent control of thrust and specific impulse such as VASIMR isrequired [1]. The authors proposed a simple structure with a single helical antennaand a single Radio Frequency (RF) power supply that controls thrust and specificimpulse independently via plasma generation and ion heating, respectively [2]. Ahelical antenna can excite different polarized waves bi-directionally: the right-hand polarized wave is suitable for plasma generation and the left-hand polarizedone is effective for ion heating. As there is a single RF power source, the externalmagnetic field is used to weight between plasma generation and ion heating. TheRF frequency is usually fixed, so corresponding wavelength of the wave changeswith variation of the magnetic field, and thus wave excitation efficiency will bedeteriorated. The authors proposed a Variable Pitch Helical Antenna (VPHA),which was composed of rickrack plates and had a function of variable axial length.The excitation efficiency is expected to be improved by adjusting the axial lengthof the antenna.

According to the results of the VPHA experiment [3], the plasma densityvaries with the axial length of the antenna as well as RF power and externalmagnetic field. However, these parameters does not necessarily satisfy dispersionrelation of the expected wave. Measured spatial variation of the RF magneticfield does not show the wave characteristics. These may be because the plasmadensity is relatively low.

In this research, high density plasma generation is tried by introducing a singlepulse operation of RF and gas supply [4]. Examination of dispersion relationand RF field measurements are planned, and the detail will be presented in theconference.

This work is supported by a Grant-in-Aid for Scientific Research (16K13920)from JSPS.

References[1] F. R. Chang Diaz, Trans. Fusion Sci. Tech., 43 (2003) 3.[2] H. Wakabayashi, et al., 21st Int. Toki Conf. (2011) P1-103.[3] Y. Maruki, et al., 12th Int. Conf. on Open Magnetic Systems for Plasma

Confinement (2018) P18.[4] T. Kawaguchi, et al., Trans. Fusion Sci. Tech., 63 (2013) 307.

P1-47

Improvement of electron collection using a

magnetic field with low mirror ratio in a

secondary electron direct energy converter

simulator

Atsuya KURISUNO1, Satoshi NAKAMOTO1,

Kazuya ICHIMURA1, Hiromasa TAKENO1,Yuichi FURUYAMA2, Akira TANIIKE2

1. Graduate School of Engineering, Kobe University2. Graduate School of Maritime Sciences, Kobe University

As an energy source for next generation, D-3He fusion is expected becauseneutron is not produced in the reaction. Direct energy conversion can be ap-plied, and a traveling wave direct energy converter (TWDEC) was proposed asan energy recovery device for high energy protons [1], however, some protons isnot decelerated and conversion efficiency is limited. Using a Secondary Electron(SE) Direct Energy Converter (SEDEC) was proposed to recover the fast protonspassing through the TWDEC [2]. In SEDEC, protons are irradiated to the metalfoil and is recovered by collecting the generated SEs.

A series of experiments were performed, but most of SEs did not arrive atelectron collectors, and seemed to flow into anteroposterior foil electrodes. Amagnetic field by permanent magnets perpendicular to the proton beam wasintroduced to guide SEs to collector, however, magnetic mirror effect in front ofthe permanent magnets blocked SEs arrival at the collectors.

Some trials to improve SE collection were reported. Recently, an experimentsetting the mirror ratio between the surface of the collector and the SE generationpoint to less than 1 was done, and relatively large amount of SE collection wasobserved [3]. In this study, we further proceed with this scheme. We introduceda new arrangement of the permanent magnets, and realized a magnetic field witha lower mirror ratio. The amount of collected SEs increased as expected. Thefurther experiments are going and detailed results and discussion will be presentedin the conference.

References[1] H. Momota, LA-11808-C, Los Alamos Natl. Lab. (1990) 8.[2] D. Akashi, et al., Trans. Fusion Sci. Tech., 63(1T) (2013) 301.[3] A. Kurisuno, et al., 12th Int. Conf. on Open Magnetic Systems for Plasma

Confinement (2018) P19.

P1-48

Plasma generation study on dependence of

magnetic fields ratio between upstream and

downstream for a single helical antenna thruster

Hiromune YAMAZUI1, Satoshi NAKAMOTO1,Kazuya ICHIMURA1, Hiromasa TAKENO1

1. Graduate School of Engineering, Kobe University

Recently, the range of human activity has expanded by development of spaceengineering, and more functional and fuel-efficient rocket engines are required.The VASIMR engine enables independent control of thrust and specific impulseby separating functional parts of plasma generation and ion heating [1]. However,the composition of multiple power sources and antennas results in upsizing of theengine.

The authors proposed a miniaturizing method to simultaneously performplasma generation and ion heating using a single helical antenna [2]. The methodis based on a bi-directional wave excitation characteristic of a helical antenna.The excited waves are right and left hand polarization, respectively, which cor-respond to R and L waves and effective for plasma generation and ion heating,respectively.

These waves have different dispersions, so the suitable magnetic field strengthfor the radiation of one antenna is generally different. Thus, the excitation ef-ficiency can be improved by setting different magnetic field strength in each di-rection. In the previous experiment, plasma generation was examined with anaxially movable helical antenna in a magnetic field region with an axial gradient[3]. The plasma density varied by not only magnetic field strength, but also itsratio between upstream and downstream of the antenna.

In this study, further investigation by changing the axial field distribution willbe performed. Results with a wider variety of magnetic field conditions will beannounced at the meeting.

References[1] F. R. Chang-Diaz, Trans. Fusion Sci. Tech. 43 (2003) 3.[2] H. Wakabayashi, et al., 21st Int. Toki Conf. (2011) P1-103.[3] K. Inui, et al., 27th Int. Toki Conf. (2018) P2-71.

P1-49

The Formation of Cone Structure On Silicon

Irradiated with Low Energy Helium Plasma:

Flux and Sample Temperature Dependences

Quan SHI1, Shin KAJITA2, Noriyasu OHNO1,

Masayuki TOKITANI 3, Nagata DAISUKE3

1. Graduate School of Engineering, Nagoya University, Nagoya 464-8603,Japan

2. Institute of Materials and Systems for Sustainability, Nagoya University,Nagoya 464-8603, Japan

3. National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292,Japan

Black silicon is a prominent material in solar cell applications. One of the con-ventional methods to fabricate black silicon is reactive ion etching that involveshigh energy of several different species with hundreds of electron volts by usingradiofrequency bias which may cause surface damages [1]. Other approaches toform black silicon using such as neutral beam etching are also complicate [2]. Re-cently, by using noble gas (He) plasma irradiation on doped silicon with relativelylow ion energy around 50 eV, black silicon was obtained due to the formation ofcone structure on the surface. Sputtering is considered as the main mechanismfor cone formation by comparing the exposed area with pristine silicon [3, 4].However, the complete evolution of cone structure is not clear.

In this article, similar experiments have been done by the linear plasma device,Co-NAGDIS. With the assistance of airflow cooling system and electron heating,we investigate the influence of ion flux, ion energy, sample temperature, etc.,on the feature of surface morphology. The experimental environment to get theoptimal black silicon is found. Moreover, a flux gradient experiment is done bypartially cover the silicon sample with small space above it. It shows that theion flux is a crucial factor to form different morphology of the silicon surface.According to these results, the mechanism of the cone structure generation isdiscussed

[1] X. Liu, P.R. Coxon, M. Peters, B. Hoex, J.M. Cole, and D. Fray, EnergyEnviron. Sci. 7 (2014) 3223-3263.

[2] S. Samukawa, Jpn. J. Appl. Phys. 45 (2006) 2395-2407.[3] S. Takamura, Y. Kikuchi, K. Yamada, S. Maenaka, K. Fujita, Y. Uesugi,

Jpn. J. Appl. Phy. 55 (2016) 12301[4] S. Takamura, T. Aota, H. Iwata, S. Maenaka, K. Fujita, Y. Kikuchi, Y.

Uesugi, Appl. Surf. Sci. 487 (2019) 755-765

P1-50

Simulation experiment of thermal load reduction

by direct energy conversion using an ion beam

injector

Hiromasa TAKENO1, Hirofumi KANNO1,Kazuya ICHIMURA1, Satoshi NAKAMOTO1,Hiroto MATSUURA2, Kazuo OOKAWA3,Yousuke NAKASHIMA3, Mafumi HIRATA3

1. Graduate School of Engineering, Kobe University2. Radiation Research Center, Osaka Prefecture University3. Plasma Research Center, University of Tsukuba

Reduction of divertor thermal load is one of the key issues to realize fusion re-actor. The authors proposed an application of direct energy conversion (DEC)technique to thermal load reduction, the principle of which is charge separationby CuspDEC and deceleration by retarding field [1]. We reported a few resultsof experiments [2,3], but the objective plasma in these experiments is rather lowdensity. The effectiveness of the scheme should be confirmed for higher densityplasma. The present research aims the flux level in the middle range between thepast experiments and GAMMA 10/PDX, and the research is positioned as thepreliminary stage of the experiment in GAMMA 10/PDX.

In the actual experiment, we employ an ion beam injector developed forplasma-wall interaction investigation in GAMMA 10/PDX [4]. It can provide5A hydrogen beam with over 10 kV, but proper energy for the present study ismuch lower. We adjusted operation parameter to lower energy, and 0.2 A with2 kV is realized at present. The device has a beam guide space in front of theion source, where a target plate and measurement devices can be placed. In thepresent study, a beam energy analyzer and a calorimeter system developed bythe authors are settled. As an initial experiment of the research, incident heatquantity and temperature variation will be observed. The detail will be shown inthe conference.

References[1] H. Takeno, et al., Trans. Fusion Sci. Tech. 63 (2013) 131.[2] K. Ichimura, et al., Fusion Eng. Des. 136 (2018) 381-385.[3] Y. Nonda, et al., Plasma and Fusion Res. 13 (2018) 3405050.[4] K. Fukui, et al., 32nd Ann. Meeting of JSPF (2015) 24aD38P.

P1-51

Mixer Comb-frequency Generator for Differential

Phase Reflectometry

Daichi OGATA1, Ryuichi ASHIDA1,Masaharu FUKUYAMA1, Ryoya KATO1,

Michihiro KUDO1, Ryuya IKEZOE2, Hiroshi IDEI2

1. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity

2. Research Institute for Applied Mechanics, Kyushu University

　Differential Phase Reflectometry (DPR) has been developed to measure phasedelay for electron density profile diagnostics on the QUEST (Q-shu UniversityExperiments with Steady State Spherical Tokamak) [1]. This technique is basedon phase delay measurements between probing and reflected waves from densitycutoff layers in the plasma. In this measurement system, the differential fre-quency probing waves are excited at a Lower Side Band (LSB) and an UpperSide Band (USB) by mixer operation of up/down conversion around the pumpedLocal Oscillator (LO) wave. While, Comb-frequency oscillators have been usedto excite several frequency waves simultaneously for the temporal density profilediagnostics. A mixer harmonic operation is now considering to generate manyfrequency waves for the comb-frequency DPR. The operating frequency rangewas 7-14 GHz in the reflectometry on the QUEST. Several probing-waves of [7.5/8.5/ 9.5/ 10.5/ 11.5] GHz for density-profile diagnostics were excited with a lowfrequency difference of ± 70 MHz simultaneously by the mixer comb-frequencygenerator. The multiple reflection points at cutoff layers corresponding to eachfrequency component are measured at the same time by this measurement sys-tem.　　 In this paper, a method to generate several probing waves of [(7.5/ 8.5/ 9.5/10.5/ 11.5) ± 0.07 ] GHz was described with the mixer operation of up/downconversion and harmonic waves. Additionally, comb-frequency waves as the LOat the DPR heterodyne detection of [(7.5/ 8.5/ 9.5/ 10.5/ 11.5)] GHz were ex-cited by the mixer up/down converter operation. For this measurement system,three oscillators of [0.07/ 1.00/ 9.50] GHz and four frequency mixers were pre-pared. The phase evolution and reflector positions were detected by receivingantenna in Proof-of-Principle experiments with an aluminum reflector plate atlow power test. 　The frequency derivative method was proposed to remove themultiple-wall-reflection effect in QUEST vacuum vessel [2]. For the frequencyderivative method, low frequency component of 78.125 kHz was prepared. The78.125 kHz was determined by the frequency resolution of measuring instrument.The excited sub-sideband components of [(7.5/ 8.5/ 9.5/ 10.5/ 11.5) ± 0.07 ±0.000078125] GHz around the LSB and USB were also confirmed.　

P1-52

[1]H. Idei, et. al., J. Plasma Fusion Res., SERIES, 9 112 (2010)[2]H. Idei, et. al., Rev. Sci. Instrum., 85 11D842 (2014)

P1-52

NEW IDENTIFICATION OF UV-VISIBLE

EMISSION LINES FROM HIGHLY CHARGED

TUNGSTEN IONS IN COBIT

Shota ERA1, Hiroyuki. A. SAKAUE2,Izumi MURAKAMI2,3, Nobuyuki NAKAMURA4,

Daiji KATO1,2

1. Kyushu University2. National Institute for Fusion Science3. SOKENDAI4. Institute for Laser Science, The University of ElectroCommunications

Tungsten will be used as divertor materials in the International ThermonuclearExperimental Reactor (ITER) because of the high sputtering threshold, the high-est melting point among all the elements, and less tritium retention. Spectro-scopic data of tungsten are very important for experimental studies on edge trans-port of tungsten ions into the core of ITER plasmas. Visible lines are particularlyuseful, because in this wavelength range precise measurements are facilitated andoptical fibers are available. Usage of the visible line emission of tungsten hasbeen exclusively limited to neutral or low charge state ions. Kato et al. [1] ob-served magnetic-dipole (M1) lines of highly charged W27+ and W26+ ions inUV-Visible spectra obtained by tungsten pellet injection at Large Helical Device, and showed potential usefulness of the M1 lines as a novel mean for tungstenmeasurements. However, knowledge on such M1 lines available for the tungstenmeasurements of fusion plasmas are still very limited. In the present work, wemeasured UV-Visible emission lines of highly charged tungsten ions by using acompact electron beam ion trap, called CoBIT [2]. Tungsten is introduced as avapor of W(CO)6 through a gas injector equipped to the CoBIT. Highly chargedions are produced by a magnetically-compressed highdensity electron beam. Theproduced ions are trapped in a radial space charge potential of the electron beamand in an axial potential well of trap electrodes inside the CoBIT. Photon emis-sion from the trapped ions is observed through a view port at the right angle tothe electron beam. UV-Visible spectra are measured with a spectrometer of a1200 gr/mm grating blazed at 400 nm. In this paper, we present newly identifiedlines in 290 420 nm of highly charged tungsten ions of charge states 20 - 29 pro-duced with electron beam energies between 490 and 1440 eV. Wavelengths of thelines are carefully calibrated using standard lamps, and the results are comparedwith theoretical atomic structure calculations.

(1) D. Kato et al., 26th IAEA Fusion Energy Conference(17-22 oct., 2016,Kyoto, Japan), EX/P8-14.

P1-53

(2) N. Nakamura et al., Rev. Sci. Instrum. 79 (2008) 063104.

P1-53

Two-point measurements along the magnetic

field in detached linear plasmas with laser

Thomson scattering method

Hiroki TAKANO1, Shin KAJITA2, Hirohiko TANAKA1,Shogo HATTORI1, Yohei IMAEDA3, Noriyasu OHNO1

1. Graduate School of Engineering, Nagoya University2. Institute of Materials and System for Sustainability, Nagoya University3. Department of Engineering, Nagoya University

In a fusion reactor, divertor is expected to remove helium ash produced in fusionreaction. Furthermore, heat flux of 100MW/m2 flows into the divertor region,while an engineering heat load capacity is 5 − 10MW/m2. Therefore, reductionof heat flux in front of the divertor plate is one of the most important tasks [1,2].

Detached plasma is proposed as a promising method for reduction of heat load[3,4]. In linear plasma devices, increased neutral gas pressure promotes plasma-gas interactions and vanishes the plasma near the end target. In addition, toproduce the detached plasma, high electron density in the upstream and longconnection length of the magnetic field are required.

In order to design future fusion reactor, it is important to evaluate the decayscale length of detached plasma parameters along the magnetic field and theirdependences on upstream plasma condition. In this study, we have performedtwo-point measurements along the magnetic field in the linear plasma deviceNAGDIS-II, which has approximately two meters in length. To obtain accurateplasma parameters in upstream and downstream regions without disturbance,two laser Thomson scattering (TS) measurement systems have been developed[5,6].

We measured the electron density and electron temperature in the upstreamand the downstream of the plasma with the two TS systems. The recombinationfront of the detached plasma was identified from the light emission associated withrecombination processes, and the position was varied by changing the helium gaspressure around the end target. From the TS results, we evaluated the e-foldinglength as changing the discharge condition. It was found that the e-folding length,which was typically on the order of 0.1 m.

[1] A. Herrmann et al., Plasma Phys. Control. Fusion 37, 17 (1995).[2] K. Tokunaga et al., J. Nucl. Mater. 212-215, 1323 (1994).[3] N. Ohno et al., Nucl. Fusion 41, 8 (2001).[4] S. Kajita et al., Phys. Plasmas 24, 073301 (2007).[5] H.Takano et al., Plasma Fus. Res. 14, 2405031 (2019).[6] H. Ohshima et al., Plasma Fusion Res. 13, 1201099 (2018).

P1-54

Measurement of Nitrogen Atom Density

Generated by Spiral Shape Recombining Plasma

Koji ASAOKA1, Noriyasu OHNO1, Shin KAJITA2,

Hirohiko TANAKA1, Ryosuke NISHIO3, Yuki HAYASHI4

1. Graduate School of Engineering, Nagoya University2. Institute of Materials and Systems for Sustainability, Nagoya University3. School of Engineering, Nagoya University4. National Institute for Fusion Science

Nitrided metal has great resistance of abrasion, fatigue, corrosion, and heat.Therefore, various nitriding methods have been actively studied. The gas nitrid-ing [1] and the ion nitriding [2] are utilized for industry because of its simplicity.However, they have low productivity or ion damage. To conquer the disadvan-tages, the radical nitriding method was proposed [3,4]. In this method, the ma-terials are irradiated with nitrogen (N) radicals to form a nitrided surface. Inconventional plasma discharges with an electron temperature T

e< 10 eV, due to

the high binding energy of the N molecule, it is difficult to generate high densityN atoms.

In our previous work [5], we proposed a new method to generate a higherdensity atomic N source based on the dissociate recombining process. First,high density N+

2plasma is generated in the upstream region by utilizing DC

discharge. Then, the produced plasma is transported along the long magneticfield and cooled down below 1 eV. As a result, the ionizing plasma is changed torecombining plasma, where dissociative recombination (N+

2+ e− → N+ N) would

become dominant and generate high density N atoms [6].In this work, to demonstrate the proposed method, we used the simple torus

device NAGDIS-T [5]. It can generate a high density N+

2plasma by DC discharge.

2D Langmuir probe measurement showed that Tedecreased from the upstream

along the long magnetic field enough to produce the dissociative recombiningplasma. The ground state N atom density n

Nestimated with applying coronal

equilibrium to optical emission spectroscopy was ∼ 1017 m−3. In addition, we aretrying to calculate n

Nby applying the actinometry method.

[1] T. Takase, Tetsu-to-hagane 66, 9, 1423 (1980)[2] M. Hudis, J. Appl. Phys. 44, 1489 (1973).[3] H. Shoyama et al., J. Vac. Sci. Technol. 24, 1999 (2006).[4] H. Akatsuka et al., IEEE Trans. Plasma Sci. 42, 3691 (2014).[5] K. Asaoka, et al., Plasma Fusion Res. 14, 3401069 (2019).[6] N. Ohno, Plasma Phys. Controlled Fusion, 59, 3, 034007 (2017).

P1-55

Development and evaluation of ion energy

analyzer for energetic ion measurement in a

linear plasma device NUMBER

R. OCHIAI1, A. Okamoto1, T. Fujita1, H. Arimoto1,H. Hachikubo1, M. Sugimoto1, K. Iizuka1

1. Graduate School of Engineering Nagoya University

In order to simulate alpha particles generated in DT reaction in a magneticconfinement fusion reactor with high energy ions, a new energetic-ion-productionmethod for smaller devices is required. We are developing a new energetic-ion-production method using the linear plasma experiment device‘NUMBER’whichis excellent in controllability and easiness of measurement.

In order to directly measure high energy ions in plasma, we have developedan ion energy analyzer for measuring the ion energy distribution and measuredbulk ions. When we measured in the plasma whose electron density was higherthan1017 m−3, the ratio of the ion saturation current obtained at the collectorelectrode to the electron density was smaller than that at lower electron density.We examined the space charge limitation inside the analyzer for this cause.

The analyzer consists of three mesh grids and a collector. The first grid isleft at floating potential, in order to minimize plasma disturbance due to theelectric field inside the analyzer. We apply a negative voltage to the second gridV2nd to retard electrons (V

2nd = −150V), and sweep voltage to the third gridV3rd to separate ion according to energy (−10V < V

3rd< 30V). We obtain the

ion energy distribution from the dependence of the ion current obtained at thecollector on the sweep voltage. It is considered that the ion current is limited byspace charge between the second and third grids.

We measured the current flowing into each grid and the collector while chang-ing the voltage applied to the third grid for each shot in the plasma whose electrondensity was 4× 1017 m−3. While, V

2nd was kept at -150V, V3rd was varied in the

range of −180V < V3rd

< −10V, which was lower than V3rd = −10V at which

we obtain the ion saturation current in measuring the ion energy distribution. AtV3rd > −20V, the current flowing into the third grid and the collector decreased

and the current flowing into the second grid increased. This phenomenon was notseen when the plasma electron density was about 5 × 1016 m−3. So, it was con-firmed that the decrease in the ion current occurred between the second grid andthe third grid. In addition, we started investigation on space charge limitationinside the analyzer by using a PIC code.

This research is partially supported by JSPS KAKENHI Grant NumbersJP19H01869, JP17H06231.

P1-56

Plasma potential measurements in detached

plasmas by using electrostatic probes

Shogo HATTORI1, Hirohiko TANAKA1, Shin KAJITA2,

Hiroki TAKANO1, Isaya SAEKI1, Noriyasu OHNO1

1. Graduate School of Engineering, Nagoya University2. Institute of Materials and Systems for Sustainability, Nagoya University

In nuclear fusion reactors, the divertor plate is exposed to a huge heat load.Therefore, reducing the head load is an important issue in order to realize the re-actor. The most effective solution to the problem is producing detached plasma.Detached plasma extinguishes the plasma due to volumetric plasma recombina-tion. In the linear plasma device NAGDIS-II [1], detailed behaviors of plasma pa-rameters during the plasma detachment have been clarified. We measured plasmaparameters in detached helium (He) of the NAGDIS-II [1]. The laser Thomsonscattering systems installed in the upstream and downstream of NAGDIS-II wereutilized to measure the radial profiles of the electron temperature T

eand density

ne[2]. Radially and also axially movable Langmuir probe was installed in the

NAGDIS-II to measure two dimensional profiles of neand T

enear the recom-

bination front region [1]. A double probe (DP) method can provide accuratemeasurement of n

eand T

eevaluated by the laser Thomson scattering systems [3].

On the other hand, plasma potential also plays an essential role in the productionand stability of the detached plasmas, however, it has not been well investigated.

In this research, we have measured the plasma potential as well as neand T

e

in NAGDIS-II using an emissive probe (EP), and DP with a single probe (SP).The EP can provide a direct measurement of the plasma potential by heatingthe filament. We improved accuracy of the EP diagnostics by applying switchingcircuit of heating current. The floating potential of EP was measured just afterthe heating current was switched off. On the other hand, DP can measure nrme

and Te, but cannot measure the plasma potential in principle. By measuring

the floating potential with SP located nearby the DP measurement position, theplasma potential can be numerically estimated from the floating potential andTe. The plasma potential measured with EP is in a good agreement with the

DP measurement in the detached plasma. It suggested that EP measurement isapplicable in the detached plasma. Further, it became clear that above a certaingas pressure, the plasma potential decreased as the gas pressure increased. Thedecrease in plasma potential is considered to be due to increase in T

e.

[1] N. Ohno, et al., Nuclear Materials and Energy 19 (2019) 458-462.[2] H. Takano, et al., Plasma Fus. Res. 14, (2019) 2405031.[3] Y. Hayashi, et al., Contri. Plasma Phys. DOI: 10.1002/ctpp.201800088.

P1-57

Large eddy simulations of magnetized plasmas

described by extended MHD model

Hideaki MIURA1, Fujihiro Hamba2

1. National Institute for Fusion Science2. Institute of Industrial Science, the University of Tokyo

Large eddy simulations (LES) of magnetized plasmas described by an extendedMHD model are carried out for the purpose of verifying the applicability of theLES approach for studying growth of short-wave instability and transition to tur-bulence. An anistropic sub-grid-scale (SGS) model developed for this purpose isadopted. The extended MHD model used in this study is equipped with both theHall and gyro-viscous terms. On one hand, this model does not need assumptionsto the Braginskii’s full two-fluid model other than the massless-electron assump-tion so that the applicability of the model is wider than other extended MHDmodels. On the other hand, however, this model suffers from a strong numericalinstability which can arise Hall and/or gyro-viscous terms. We show that theLES approach provides a numerical approach to simulate short-wave instabilityand turbulence by the use of relatively small number of grid points.

While it has been shown that the LES approach can reduce numerical cost ofnonlinear two-fluid simulations of ballooning modes in the Large Helical Device(LHD)[1], the SGS model used for the LES is intrinsically isotropic, causingexcessive damping of the velocity component parallel to the magnetic field line [2].In order to improve this point, an anisotropic SGS model for a magnetized plasmais adopted. We carry out LES of homogeneous extended MHD turbulence withconstant external magnetic field, and verify the applicability of the SGS modelfor the anisotropic turbulence simulations.In the presentation, we will also reportsome suitable combinations of the adjustable parameters of the SGS model, anddiscuss the applicability of the SGS model for LES of two-fluid ballooning modesin fusion plasma.[1] H. Miura, F.Hamba, and A.Ito,”Two-fluid sub-grid-scale viscosity in nonlinearsimulation of ballooning modes in a heliotron device”, Nuclear Fusion 57 076034(2016).[2] H.Miura, K.Araki, and F.Hamba, ”Hall effects and sub-grid-scale modeling inmagnetohydrodynamic turbulence simulations”, J. Comput. Phys. 315 385396(2016).

P1-58

Global effect on collisional transport of tungsten

impurity

R. Kanno1, G. Kawamura1, M. Nunami1, Y. Homma2,A. Hatayama3, K. Hoshino3

1. National Institute for Fusion Science2. National Institutes for Quantum and Radiological Science and Technology3. Keio University

By using two types of drift-kinetic equation solvers based on the “local” and the“global” models [1], the radial particle flux of tungsten impurity affected by thefriction and the thermal forces [2] is evaluated in the edge region of a tokamakplasma core. Here, the plasma including the impurity and the background ionis presupposed to be quasi-steady. It is found by the “global” model that theradial particle flux (and also the solution fZ = fM,Z + δfZ of the drift-kineticequation for the impurity) on a magnetic flux surface is influenced by the valuesof δfZ all over the edge region. Here, fM,Z is the Maxwellian distribution. In thepresent study, this is called the “global effect” on the impurity transport, andthis effect is caused by the magnetic drift term in the drift-kinetic equation. Thefinite orbit widths of impurity guiding centers across the neighboring magneticflux surfaces are generated by the magnetic drift term. The solution of the drift-kinetic equation on an arbitrary magnetic flux surface cannot be independentfrom the friction force and the thermal force on the neighboring magnetic fluxsurfaces because the orbit width is non-zero. Furthermore, this friction forcedepends on the values of δfZ on the neighboring magnetic flux surfaces throughthe parallel flow velocity. The global effect is ignored in the conventional “local”model of neoclassical transport, and has not been taken into account in previousstudies. The simulation result given by using KEATS code [3] based on the“global” model differs visibly from the theoretical estimate given by the “local”model. It is confirmed that the “global effect” on the collisional transport ofthe impurity in the edge region is significant for evaluating the radial particleflux through the comparison between the drift-kinetic simulations based on the“global” and the “local” models.

This work was performed with the support and under the auspices of the NIFSCollaborative Research Programs “NIFS19KNTT052” and “NIFS19KNST142”,and the IFERC-CSC project “DKSOIT”.

[1] S. Matsuoka et al., Phys. Plasmas 22 (2015) 072511.[2] Y. Homma et al., Nucl. Fusion 56 (2016) 036009.[3] R. Kanno et al., Nucl. Fusion 58 (2018) 016033.

P1-59

Improved Linearized Model Collision Operator

for Kinetic Plasma Simulation

H. Sugama1,2, S. Matsuoka1,2, S. Satake1,2,

M. Nunami1,2, T.-H. Watanabe3

1. National Institute for Fusion Science, Toki 509-5292, Japan2. The Graduate University for Advanced Studies (SOKENDAI), Toki 509-

5292, Japan3. Department of Physics, Nagoya University, Nagoya 464-8602, Japan

In magnetically confined toroidal plasmas, Coulomb collisions are a main cause ofthe neoclassical transport, which is investigated by using the drift kinetic equa-tions. On the other hand, the turbulent transport driven by plasma microinsta-bilities is studied by the gyrokinetic equations, which still need collisional dissi-pation for realizing the steady turbulent state as well as for correctly describingbehaviors of microinstabilities and zonal flows. Therefore, it is desirable to use agood collision model in the kinetic equations, which is easy to treat analyticallyor numerically and satisfies physically correct constraints such as conservationlaws of particles, momentum, and energy. In our previous work [H. Sugama,et al., Phys. Plasmas 16, 112503 (2009)], a linearized model collision operatorfor multiple ion species plasmas is presented that conserves particles, momen-tum, and energy, and satisfies adjointness relations and Boltzmann’s H-theoremeven for collisions between different particle species with unequal temperatures.In this linearized model operator, the field particle part is simplified from thatof the linearized Landau collision operator while it accurately keeps momentumand energy conservation properties. The model operator has been successfullyapplied to studies of neoclassical and turbulent transport in relatively low colli-sional regimes although the difference between its field particle part and that ofthe Landau model is anticipated to increase in a highly collisional regime. In theITER plasmas, impurity ions of tungsten are considered to remain in the Pfirsch-Schlter regime even though bulk ions and electrons are in the banana regime.For such a case, it is necessary to use a collision model which is accurate in allcollisional regimes. This work presents the improved linearized model collisionoperator, which is properly applicable to all cases from low to high collisionality.The improved model is constructed so as to give exactly the same friction-flowrelations as those derived from the linearized Landau operator, and accordinglyit can be used to accurately evaluate neoclassical transport fluxes in all collisionalregimes. In addition, the improved collision operator for gyrokinetic equations isderived by taking the gyrophase average with the finite gyroradius effect takeninto account.

P1-60

Study on Effects of Zonal Flows and Trapped

Electrons on Turbulent Transport by Reduced

Models for Helical Plasmas

S TODA1,2, H SUGAMA1,2

1. National Institute for Fusion Science, Toki 509-5292, Japan2. The Graduate University for Advanced Studies, Toki 509-5292, Japan

In order to realize fusion energy, the turbulent transport in toroidal plasmas isone of the most critical issues. Since it is known that zonal flows can regulatethe turbulent transport, many studies are done to investigate the role of zonalflows to improve plasma confinement. In these studies, nonlinear gyrokinetic sim-ulations have been performed to accurately determine the relation between theturbulent transport level and the zonal flow amplitude. However, such nonlin-ear simulations require a huge computational cost for the parameter scan in thespace of magnetic field configurations and plasma profiles. To reduce the com-putational cost, the reduced models are proposed, which can quickly predict thenonlinear simulation results by the linear simulation results in helical plasmasunder the conditions of adiabatic electrons [1] and kinetic electrons [2,3]. Thesereduced models are presented for the representative plasmas for the ion temper-ature gradient modes in the Large Helical Device (LHD). To evaluate turbulentelectron particle and heat transport fluxes, kinetic electrons need to be treatedin gyrokinetic simulations. Especially for helical plasmas, trapped electrons showcomplicated drift motions and it is a challenge to quantitatively clarify how theyaffect instabilities, zonal flows, and turbulent transport. In this study, effects oftrapped electrons on zonal flows and turbulent transport in the LHD are studied.The residue zonal flow level for the case of kinetic electrons is compared withthat in the adiabatic electron condition by the linear simulation. The depen-dence of the linear simulation results on the radial wavelength is studied. Next,evaluating the quantity related with the mixing length estimate and the zonalflow decay time [4] by the linear gyrokinetic simulations, the saturation levels ofturbulence and zonal flows are predicted from the reduced models with the differ-ent field configurations and plasma profiles. The effects of trapped electrons onzonal flows and plasma confinement are investigated. Thus, the reduced modelspresent useful means to evaluate the turbulent transport levels and the zonal flowamplitudes under the various conditions which appear in the experimental dataanalyses.[1] M. Nunami et al, Phys. Plasmas 20, 092307 (2013)[2] S. Toda et al., Phys. Plasmas 26, 012510 (2019)[3] S. Toda et al., Plasma and Fusion Research 14, 3403061 (2019)[4] H. Sugama and T. -H. Watanabe, Phys. Plasmas 13, 012501 (2006)

P1-61

Quasioptical modeling of the electron cyclotron

resonance heating

Kota YANAGIHARA1, Ilya DODIN2, Shin KUBO1,3,Toru TSUJIMURA3

1. Nagoya University2. Princeton Plasma Physics Laboratory3. National Institute for Fusion Science

Electron-cyclotron resonance heating and current drive with mm waves requiremodeling of the power deposition with high precision. Multi-dimensional full-wave simulations are prohibitively expensive at these wavelengths. Ray- andbeam-tracing techniques just requires small computational resources but are notsufficiently accurate, especially near the resonance conditions due to the ne-glect of either diffraction or transverse inhomogeneity of dissipation. We reportquasioptical modeling of the power deposition by mm-wave beams using a newcode PARADE (PAraxial RAy DEscription), which was recently presented in[arXiv:1901.00268, arXiv:1903.01357, arXiv:1903.01364; to appear in Phys. Plas-mas 26 (2019)]. The PARADE simulations have accounted for beam refraction,diffraction, and mode conversion, such as the X–O mode conversion at the fusionplasma edge caused by the magnetic shear. Here, the transverse inhomogeneityof dissipation is newly introduced in the code, and is calculated from first princi-ples to capture the inhomogeneity accurately. Strongly diffracted and additionalbended beam dynamics after its passage through the resonance conditions, andmore realistic power deposition profiles in fusion plasma are presented.

P1-62

Kinetic full wave analysis of electron cyclotron

waves in a tokamak plasma using integral

operator method

S. A. KHAN1, A. FUKUYAMA2, H. IGAMI3,H. IDEI4

1. National Centre for Physics at QAU, Islamabad Pakistan2. Department of Nuclear Engineering, Kyoto University, Kyoto Japan3. National Institute for Fusion Science, Toki Japan4. Research Institute for Applied Mechanics, Kyushu Univ., Kasuga Japan

Externally excited electromagnetic (EM) waves are vital in auxiliary heatingapplications in tokamaks. Since ordinary (O) or extraordinary (X) waves sufferwave cutoffs in high-density regions, quantitative description of mode conversionto the electron Bernstein waves (EBW) has been desired for electron cyclotron(EC) heating and current drive in an overdense hot plasma. For optimum parallelwave number, the O-X-B mode conversion leads to EBW which is absorbed atthe cyclotron fundamental or harmonic resonance. Description and analyses ofwaves by conventional geometrical optics approaches are difficult when the wave isevanescent or wavelength is comparable to or larger than the inhomogeneity scalelength. In this situation, the full wave analysis with integral operator approachprovides a reliable method to include the hot plasma effects in the dielectrictensor, where the Maxwell’s equation with integral form of dielectric tensor isnumerically solved as a boundary-value problem by the finite element method.

We have implemented the integral formulation to study the EC wave modeconversion in a tokamak plasma. The one-dimensional full wave code TASK/W1was developed [1] and applied to the O-X-B mode conversion in tokamak config-uration. Mode conversion to the EBW near the upper hybrid resonance (UHR)layer and the EC resonance absorption in overdense plasma was successfully de-scribed and confirmed. Dependence of the EBW absorption on various parametersand consistency of the mode conversion efficiency to the analytical estimates havebeen discussed. Waveguide excitation on the wall was implemented to confirmthe pure O-mode excitation.

The TASK/W1 code is extended to perform two-dimensional (2D) full waveanalysis with the integral form of dielectric tensor. Preliminary results of 2Danalyses on an equatorial plane in tokamak plasmas are obtained and the wavestructure after the mode conversions is compared with ray tracing results.

References[1] S. A. Khan, A. Fukuyama, H. Igami and H. Idei, PFR 11, 2403070 (2016).

P1-63

Simulation study of detached helium plasma in

NAGDIS-II by using the fluid code LINDA

I. SAEKI1, H. TANAKA1, N. OHNO1, H. TAKANO1,S. KAJITA2, K. SAWADA3, A. HATAYAMA4,

K. HOSHINO4, M. GOTO5

1. Graduate School of Engineering, Nagoya University2. Institute of Materials and Systems for Sustainability, Nagoya University3. Department of Applied Physics, Faculty of Engineering, Shinshu University4. Faculty of Science and Technology, Keio University5. National Institute for Fusion Science

To realize a thermonuclear fusion reactor requires the reduction of divertor heatflux less than engineering limited value. Producing detached plasma is the mostpromising method to solve the divertor heat-load problem [1]. Researchers havea keen interest in numerical simulation of edge/divertor plasmas and have devel-oped many codes. For example, large-scale numerical simulation codes such asSOLPS [2] and SONIC [3] are being utilized. However, their numerical simula-tions cannot completely reproduce the experimental observation in the detachedplasma. Several physical processes should be improved for accurate simulation ofthe detached plasma. Although detailed verification of these large-scale numericalsimulation codes is important, development and calculation costs are high.

We have simulated detached helium (He) plasmas in the linear plasma deviceNAGDIS-II [1] by using the fluid code LINDA [4] in this research. The linearplasma device has advantages of producing steady-state detached plasmas witha simple magnetic configuration, and controlling plasma (gas) parameters moreflexibly. Detailed comparison with the experimental observation contributes to adeeper understanding of underlying physics in the detached plasma.

In this presentation, we will report two topics focusing on the influence of elec-tron heating associated with three-body recombination process on the detachedplasma, and a comparison with plasma parameters measured with laser Thom-son scattering diagnostics [5] in the NAGDIS-II. Effective electron cooling ratecoefficients due to the three-body recombination evaluated by CR (Collisional-Radiative) model [6] were considered in the LINDA code. The simulation resultsshow the strong effect of the electron heating on the plasma profiles.

[1] N. Ohno, Plasma Phys. Control. Fusion 59 (2017) 034007.[2] R. Schneider et al., Contribution Plasma Phys. 46 (2006) 3.[3] H. Kawashima et al., Plasma Fusion Res. 1 (2006) 031.[4] M.S. Islam et al., Plasma Phys. Control. Fusion 59 (2017) 125010.[5] H. Takano et al., Plasma Fusion Res. 14 (2019) 2405031.[6] M. Goto, J. Quant. Spectrosc. Radiat. Transf. 76 (2003) 331-344.

P1-64

3-D equilibrium reconstruction under resonant

magnetic perturbations on EAST

Jie HUANG1, Yasuhiro SUZUKI1,2

1. National Institute for Fusion Science2. SOKENDAI, The Graduate University for Advanced Studies

The Magnetohydrodynamics (MHD) equilibrium, namely the internal balancebetween the pressure of the plasma and the electromagnetic forces, is a basis ofboth theoretical considerations and physics interpretation of experimental resultsfor magnetically confined fusion. With the toroidal symmetric configuration,the 2-dimentianal (2-D) EFIT code is most widely used for tokamak equilibriumreconstruction. Nevertheless, designed for MHD instabilities control and trans-port, the Resonant Magnetic Perturbations (RMP) will induced the 3-D magnetictopology and the tokamak equilibrium turns to be 3-D equation. Recently Ex-perimental Advanced Superconducting Tokamak (EAST) has been equipped withRMP system, which requires a more accurate 3-D equilibrium reconstruction tounder the RMP mechanism more clearly. In this work, we use the nonlinear re-sistive 3-D HINT code to solve the 3-D equilibrium reconstruction under RMPon EAST. During the RMP penetration phase in low rotation plasma, the 3-Dmagnetic field calculated by HINT shows clearly magnetic island in the corre-sponding resonant surface and stochastic magnetic topology in the plasma edgeregion. And the pressure profile given by HINT is redistributed with magnetictopology changing when RMP switched on. Especially in the pedestal region, thepressure gradient will obviously decrease corresponding to significant magnetictopology modification by RMP. As a result, the Edge-Localized Modes (ELMs)can be suppressed or strong mitigated with respect to peeling-ballooning mode.In addition, the plasma rotation effects on the final 3-D equilibrium is also takeninto account in the calculation, which could induce the plasma screening effectsleading to shield the RMP corresponding to experimental weak ELMs mitigationcase. To suppress and mitigate ELMs strongly, the deep penetration of RMP fieldis a key point, or the RMP configuration and its amplitude and in particular, theplasma rotation play the important roles from our modeling.

P1-65

Equivalent-Circuit Model for Superconducting

Linear Acceleration System: Improvement of

Acceleration Performance

Takazumi YAMAGUCHI1, Teruou TAKAYAMA2,Atsushi KAMITANI2, Hiroaki OHTANI1,3

1. SOKENDAI (The Graduate University for advanced Studies)2. Yamagata University3. National Institute for Fusion Science

As an alternative pellet injection system for the helical type fusion system, asuperconducting linear acceleration (SLA) system has been proposed recently [1].The SLA system is composed of an electromagnet and a pellet container to whicha high-temperature superconducting (HTS) film is attached. In the SLA system,the container is accelerated by an interaction between a shielding current densityin the HTS film and a magnetic flux density generated by the electromagnet.According to the rough estimation, the SLA system can accelerate the containerup to 5–10 km/s. However, the acceleration performance of the SLA system doesnot become clear because the SLA experiment is not performed.

In the present study, we assume that an axisymmetric HTS film is exposed toan axisymmetric magnetic flux density generated by a cylindrical electromagnet.Under the assumption, a distribution of a shielding current density is approx-imated as a set of multiple current loops which is arranged concentrically. Byusing the above approximation, we obtain an equivalent-circuit model (ECM)which is composed of the electromagnet and the multiple current loops. In theECM, a circuit equation is obtained from Faraday ’s law. By solving an initial-value problem of the circuit equation and Newton’s equation of motion, we candetermine both a time evolution of the shielding current density and the dynamicmotion of the pellet container [2].

The purpose of the present study is to propose a way of improving the ac-celeration performance of the SLA system by using the result obtained by theECM. Moreover, a distribution of an electric current flowing in the electromagnetis optimized by using the topology optimization.

References

[1] N. Yanagi, G. Motojima: private communication, 2017.

[2] T. Yamaguchi et al.: IEEE Trans. Magn., 55 (2019) 7204305.

P1-66

Hybrid Method Incorporated with Meshless

approach for Electromagnetic Wave Simulation

Ayumu SAITOH1, Takazumi YAMAGUCHI2,

Atsushi KAMITANI1, Hiroaki NAKAMURA3

1. Yamagata University2. SOKENDAI3. NIFS

As is well known, the finite difference time domain (FDTD) method is widelyadopted for electromagnetic wave simulations [1]. In the standard FDTD method,a target domain needs to be divided into uniform orthogonal meshes. Therefore,it is difficult to apply the FDTD method to the problem in which a target domainhas the complex structure.

As another method of electromagnetic wave simulations, the hybrid method isused [2]. In this method, the target domain must be divided into a set of elements.Therefore, the mesh generation must be executed even in the hybrid method.Recently, some numerical approaches which requires only node information areproposed. If the meshless approach can incorporate into the hybrid method, wemight resolve the above disadvantage.

The purpose of the present study is to develop the hybrid method incorporatedwith meshless approach for the electromagnetic wave simulation and to investigateits performance. The numerical results will be shown in the conference.

[1] K. Yee, IEEE Trans. Antennas Propag., 14(3), pp. 302-307, May 1966.[2] Z. Xiang et al., Prog. Electromagnetics Res., 22, pp. 107-129, 1999.

P1-67

Molecular Dynamics Simulation on Structural

Change of Tritium-substituted Polyethylene

Using Reactive Force Field

Haolun LI1, Susumu FUJIWARA1,Hiroaki NAKAMURA2,3, Tomoko MIZUGUCHI1,

Takao OTSUKA4, Shinji SAITO5

1. Graduate School of Science and Technology, Kyoto Institute of Technology2. Graduate School of Engineering, Nagoya University3. National Institute for Fusion Science, National Institutes of Natural Sci-

ences4. RIKEN5. Institute for Molecular Science, National Institutes of Natural Sciences

Tritium is an isotope of hydrogen, which has two more neutrons in its nucleus.Rather than generated naturally in the earth, tritium is mostly produced bynuclear reaction from human activity. As tritium is radioactive, it can have abeta-decay to helium-3 ion with +1 charge, an electron particle (a beta-ray) andan antineutrino, with a half-life of 12.32 years. When polymers are exposedto tritiated water, polymers will be attacked by beta-rays. Furthermore, thehydrogen atoms on the polymers will be substituted by tritium atoms and thenthe substituted tritium atoms decay which causes further damage to the polymers.To make the mechanism of the damage by substituted tritium on polymers clearin the atomic scale, molecular dynamics simulations were performed to treat apolyethylene chain which has a simplest chemical structure among all polymers.

Reactive force field (ReaxFF), different from traditional force fields used inclassical MD simulations, is able to model chemical reactions with treating thebreaking and formation of bonds. This particular characteristic is supposed use-ful and accurate to treat the structural change and bonds change of a polymerchain after beta-decays of substituted tritium. At this stage, it is difficult tosimulate beta-decays of tritium within MD simulation method. Thus, to get aninsight into the decay effect, we removed a certain amount of hydrogen atomsin the prepared orientationally-ordered structure of polyethylene model and per-formed MD simulations using ReaxFF force field. Then appropriate factors werecalculated to analyze the simulation results.

This work was partially supported by the Joint Studies Program (2017-2019)of the Institute for Molecular Science, the Joint Research by the National Insti-tutes of Natural Sciences (NINS) (NINS program No.01111708), and was partially

P1-68

performed with the support and under the auspices of the National Institute forFusion Science (NIFS) Collaboration Research program (NIFS17KNTS050).

P1-68

Optimization study of heliotron devices

Shinsuke SATAKE1,2, Takuya GOTO1,2,Hiroyuki YAMAGUCHI1, Jose Luis VELASCO3,

Ivan CALVO3, Hisanori TAKAMARU4

1. National Institute for Fusion Science2. The Graduate University for Advanced Studies (SOKENDAI)3. Centro de Investigaciones Energticas, Medioambientales y Tecnolgicas

(CIEMAT)4. Chubu University

Optimization of heliotron-type devices, which are characterized by a pair of con-tinuous helical coils, is studied using several new numerical methods. The pa-rameters to specify the shape of the helical coils are varied starting from existingLHD configuration, and the neoclassical transport level is evaluated using a newneoclassical code, KNOSOS[1], that includes the effect of the magnetic shear andof the component of the magnetic drift tangent to flux surfaces, both of which areconsidered to be important in the evaluation of neoclassical transport in LHD-likedevices. We also adopt the machine-learning method to find out an optimizedconfiguration.

[1] I. Calvo, J. L. Velasco et al., Journal of Plasma Physics Vol.84, 905840407

P1-69

High-speed analysis of heating and current drive

with neutral beam injection in tokamak plasma

Ryoya FUNABASHI1, Takaaki FUJITA1,

Atsushi OKAMOTO1

1. Graduate School of Engineering, Nagoya University

In steady-state operation of a tokamak fusion reactor, it is required to drive thetotal plasma current by a spontaneous current (bootstrap current) and an exter-nally driven current without an inductive current. Neutral beam (NB) injectionis one of the heating and current driving methods that can meet this requirement.In order to establish plasma operation scenarios with NB injection, it is necessaryto carry out a time-dependent analysis considering transport of heat and poloidalmagnetic field. So, we newly introduced analysis modules for neutral beam ion-ization positions and for fast-ion distribution function in the integrated transportcode TOTAL. The introduced fast ion solver is high-speed and sufficiently accu-rate.

We adopted the Monte Carlo method for calculating ionization positions ofinjected neutral beam particles using the ionization cross section developed byShingo Suzuki et al.[1]. The source distribution S(v, ξ, ψ) of fast ions is deter-mined from the equilibrium magnetic field at the ionization position, where vis the velocity, ξ is the pitch angle cosine ξ = v∥/v at the minimum magneticfield on the flux surface, and ψ is the poloidal flux function of the flux surface.The analysis module solves the bounce averaged Fokker-Planck equation for thefast ions ∂f/∂t = C(f) + S, where f is the bounce averaged fast ion distribu-tion function f(ξ, v, t, ψ) and C(f) is the bounce averaged collision term. In thismodule, f is expanded by the eigenfunctions Cn(ξ, ψ) of the pitch angle scatter-ing part of C(f): f(ξ, v, t, ψ) =

∑nCn(ξ, ψ)an(v, t, ψ). The velocity dependent

coefficient functions an(v, t, ψ) are calculated by a finite difference method. Fromthe derived distribution function, we calculate the toroidal fast ion current andthen obtain the NB driven current subtracting the electron shield current fromthe fast ion current. It has been confirmed that its value in steady state agreeswell with the value by the ACCOME code [2].

We can also calculate the heating power to ions and electrons from distributionfunction. We plan to investigate response of the current and pressure profiles tothe NB injection with various conditions.

reference[1]S. Suzuki et al , Plasma Phys. Control. Fusion 40 (1998) 2097-2111.[2]K. Tani, M. Azumi and R. S. Devto, J. Comp. Physics 98 (1992) 332.

P1-70

Numerical Study of Net Toroidal Current Effects

on MHD Stability of LHD Plasmas

Katsuji ICHIGUCHI1,2, Yasuhiro Suzuki1,2,

Yasushi TODO1, Masahiko SATO1, Katsumi IDA1,2,Satoru SAKAKIBARA1,2, Satoshi OHDACHI1,

Yoshiro NARUSHIMA1,2, Benjamin A. CARRERAS3,Linda E. SUGIYAMA4

1. National Institute for Fusion Science, Toki, Japan2. SOKENDAI (The Graduate University for Advanced Studies), Toki, Japan.3. BACV Sol. Inc., Oak Ridge, TN, USA4. Massachusetts Institute of Technology, Cambridge MA, USA

In some recent experiments with net toroidal current in Large Helical Device(LHD), partial collapses of the electron temperature are observed [1,2]. Thus,we systematically study the magnetohydrodynamic (MHD) stability of the LHDplasmas in the change of the net toroidal current by means of the numericalsimulations. In the analysis, we mainly use the HINT code [3] for the equilib-rium calculation and the MIPS code [4] for the linear stability and the nonlineardynamics. We change the total amount and the radial profile of the toroidal cur-rent. In the collapse occurrences in the experiments, the current is driven by theneutral beam injection in the co-direction, which increases the rotational trans-form. Then, the dangerous resonant surface with ι/2π = 1 is shifted inwardly tothe region where the Mercier stability is degraded. Also, the magnetic shear isdecreased at the surface. These effects enhances the evolution of the interchangemodes. Furthermore, in the experiments, the collapses occur just after the plasmarotation stops [1,2]. The recent numerical stability study shows that the globalflows have a stabilizing contribution to the interchange modes [5]. Therefore, wealso discuss the toroidal current effects on the stability with the change of theglobal flow. The influence of the net toroidal current on other types of instabilitywill also be discussed.Acknowledgements: This work is supported by the NIFS budget NIFS19KNST141and JSPS KAKENHI 15k06651. The computers Plasma Simulator in NIFS andJFRS1 in QST are utilized for the numerical calculations.References[1] S. Sakakibara, et al., 2015 Nuclear Fusion, 55 083020.[2] Y. Takemura et al 2019 Nucl. Fusion 59 066036.[3] Y. Suzuki, et al., 2006 Nuclear Fusion, 46 pp.L19[4] Y. Todo, et al., 2010 Plasma and Fusion Res. 5 pp.S2062.[5] K. Ichiguchi et al., 2016, Plasma and Fusion Res. 11, pp.2403035.

P1-71

Neutral effects on the structure of minimum

enstrophy flows

Daisuke AOKI1, Yusuke KOSUGA2,3

1. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity

2. research institute for applied mechanics, Kyushu University3. research center for plasma turbulence, Kyushu University

Turbulence evolves in astronomical plasma and fusion plasma, and non-linearphysics of the phenomena and effect of transportation becomes significant issue.For example, drift-wave turbulence is driven in fusion plasma and two-dimensionalflow structure like zonal flow and streamer is driven. Formation of the flow struc-ture is also reported in weakly ionized plasma. In this system, it is reportedthat not only turbulent plasma but also neutral plays an important role for theformation of flow structure. Therefore, to establish excitation of flow structureand control method focused on neutral is important. In the previous study[1],a quasi-two-dimensional plasma turbulence model including the neutral effect isproposed. Based on this model, the velocity distribution is obtained as a steady-state solution. On the other hand, in this study, we discuss the analysis of flowstructures by focusing on quasi-two-dimensional properties of turbulence. In thissystem, enstrophy, or a mean squared vorticity plays important roles in dynam-ical evolution. Since enstrophy cascades forward small scales and is dissipatedstrongly, we expected the formation of minimum enstrophy flows. Therefore, wediscuss the structure of minimum enstrophy flows under the condition of energyconservation. Here we report the relationship between plasma velocity and neu-tral parameter. Flow profile is obtained and its spatial scale is determined as afunction of neutral flows. Thus neutrals impact flow structures, not only throughdamping, but also through modifying minimal enstrophy flows.

[1] J. Vranjes, A. Okamoto, S. Yoshimura, S. Poedts, M. Kono, and M. Y.Tanaka, Phys. Rev. Lett. 89, 265002 (2002)

P1-72

Isotope Effect on Plasma Coherent Structure

Propagation

Hiroki HASEGAWA1,2, Seiji ISHIGURO1,2

1. Department of Helical Plasma Research, National Institute for FusionScience, National Institutes of Natural Sciences, 322-6 Oroshi-cho, Toki 509-5292Japan

2. Department of Fusion Science, SOKENDAI (The Graduate University forAdvanced Studies), 322-6 Oroshi-cho, Toki 509-5292 Japan

Recently, it is reported that intermittent filamentary coherent plasma structureswhich are called “blob” or “hole” play an important role in the radial transport inboundary layer plasmas by the experiments in various magnetic confinement de-vices. Also, many authors have studied theoretically or numerically the filamentpropagation dynamics on the basis of two-dimensional reduced fluid models [1, 2].On the other hand, we have developed the three-dimensional (3D) electrostaticparticle-in-cell (PIC) simulation code called “p3bd” code [3-5] in order to investi-gate the kinetic dynamics on blob and hole phenomena. The 3D-PIC simulationhas revealed the self-consistent current system [6] in a blob, the temperaturestructure in a blob [6], and the ion temperature effect on the blob dynamics [7,8]. Furthermore, the dynamics between a filament and impurity ions have beenstudied [9, 10]. In our previous work, the influence of ion mass on the blob propa-gation dynamics has been studied with the 3D-PIC simulation, and three isotopeeffects, i.e., the sheath, the polarization drift, and the gyro motion effects, havebeen evaluated. We have shown that the sheath effect is cancelled out by the po-larization drift effect and that the radial propagation speed of a blob is reducedby the gyro motion effect [11]. In this paper, we will investigate the isotope effectson hole propagation dynamics and will analyze the dependence of isotope effectson some parameters, i.e., the “blob regime”.

[1] S. I. Krasheninnikov, D. A. D’Ippolito, and J. R. Myra, J. Plasma Phys. 74(2008) 679 and references therein.[2] D. A. D ’Ippolito, J. R. Myra, and S. J. Zweben, Phys. Plasmas 18 (2011)060501 and references therein.[3] S. Ishiguro and H. Hasegawa, J. Plasma Phys. 72 (2006) 1233.[4] H. Hasegawa and S. Ishiguro, Plasma Fusion Res. 7 (2012) 2401060.[5] H. Hasegawa and S. Ishiguro, Plasma Fusion Res. 12 (2017) 1401044.[6] H. Hasegawa and S. Ishiguro, Phys. Plasmas 22 (2015) 102113.[7] H. Hasegawa and S. Ishiguro, Plasma 1 (2018) 61.[8] H. Hasegawa and S. Ishiguro, Phys. Plasmas 26 (2019) 062104.[9] H. Hasegawa and S. Ishiguro, Nucl. Fusion 57 (2017) 116008.

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[10] H. Hasegawa and S. Ishiguro, Nucl. Mater. Energy 19 (2019) 473.[11] H. Hasegawa and S. Ishiguro, Preprints of 27th IAEA Fusion Energy Con-ference (2018) TH/P7-12.

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Dynamics of resistive ballooning instability in

PLATO tokamak plasma

Shuhei TOMIMATSU1, Naohiro KASUYA1,2,Masahiko SATO3, Atsushi FUKUYAMA4,

Masatoshi YAGI5, Yoshihiko NAGASHIMA1,2,Akihide FUJISAWA1,2

1. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity, Kasuga, Fukuoka 816-8580, Japan

2. Research Institute for Applied Mechanics, Kyushu University, Kasuga,Fukuoka 816-8580, Japan

3. National Institute for Fusion Science, Toki, Gifu 509-5292, Japan4. Department of Nuclear Engineering, Kyoto University, Nishikyo-ku, Kyoto

615-8540, Japan5. National Institute for Quantum and Radiological Science and Technology,

Obuchi, Rokkasho-mura, Aomori 039-3212, Japan

In order to realize a fusion reactor, it is necessary to sustain stable confinementof high-temperature plasmas. However, in a torus plasma, the MagnetoHydro-Dynamics (MHD) instability due to the pressure gradient and the plasma currentcauses plasma collapse phenomena, which degrade the confinement. Therefore,understanding the characteristics of MHD instability is essential for stable controlof plasma performance. PLATO tokamak [1] is now under construction in Kyushuuniversity, and will install many experimental diagnostics to observe global natureof plasma dynamics. In this study, MHD simulations with the PLATO tokamakparameter are carried out by using MIPS code [2]. MIPS is a full MHD codeto calculate plasma dynamics in plasma core and SOL regions. As the initialprofile, the equilibrium is obtained by using integrated transport analysis codeTASK/EQ [3]. The routine to introduce the equilibrium data into MIPS codeis developed, and nonlinear time evolutions of three-dimensional MHD instabili-ties are calculated from various equilibrium conditions. In the first stage of thePLATO experiment, only ohmic heating plasma with rather low beta is expectedto be obtained. With increased plasma beta a pressure driven ballooning modescan become unstable. We analyze nonlinear dynamics of resistive ballooning in-stability, and its dependency on the initial plasma pressure is evaluated. Modespectra in the linear phase and nonlinear mode coupling in the dynamical pressureflattening processes are presented, which are expected in PLATO tokamak.

[1] A. Fujisawa, AIP Conf. Proc. 1993 (2018) 020011. [2] Y. Todoet al,PlasmaFusion Res.5(2010) S2062 [3] http://bpsi.nucleng.kyoto-u.ac.jp/task/

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Reconstruction of Radiation Profile of

Neon-Seeded LHD Plasma from Bolometer

Measurements with the Aid of EMC3-EIRENE

Gakushi Kawamura1, Kiyoshi Mukai1, Byron Peterson1,Yuhe Feng2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences, Toki Japan

2. Max-Planck-Institut fur Plasmaphysik, Greifswald Germany

Impurity seeding is a key technique to reduce heat load on the divertor platesof fusion devices. However, control method of the radiation power should bedeveloped to prevent excessive radiation leading to a plasma collapse and to avoidimpurity accumulation in the core. In order to utilize impurity seeding, physicalinformation such as spatial distribution radiation power is necessary to controlthe puff. Measurement of radiation power is regularly performed in LHD withmultiple bolometer systems including 2D imaging bolometers and multi-channelbolometers. The acquired numerical values are integrated intensity along thelines of sight, and the spatial resolution is restrictive compared with the spatialscale of the plasma.

In this paper, a new reconstruction method with the aid of a physical modelis proposed. The 3D fluid transport code EMC3-EIRENE [1] can provide finespatial distributions of plasma and impurities for a given set of input param-eters. The determination of the parameters always involves uncertainties, andutilization of measurements is necessary. The numerical model can provide asynthetic image or a series of intensity values simulating the optical system ofthe bolometer [2]. Experimental data of the InfraRed imaging Video Bolometer(IRVB) [3] during a neon-seeded discharge can be qualitatively reproduced bysynthetic image based on the EMC3-EIRENE result [2] which can be decom-posed into contributions from hydrogen, carbon, neon, and also the core region.As described above, the amounts of impurity sources are input parameters, andthus the radiation intensity of impurities scales with the sources. In order todetermine them, the least squares fitting of the IRVB data can be a solution ofparameter determination, and hence radiation reconstruction is realized. Detailedprocesses and results for a snapshot and also time evolution of parameters basedon time-resolved measurements will be presented in the paper.

References[1] Y. Feng et al., Contrib. Plasma Phys. 54 (2014) 426.[2] G. Kawamura et al., Plasma Phy. Contr. Fusion 60 (2018) 084005.[3] K. Mukai et al., Plasma Fusion Res. 9 (2014) 3402037.

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MHD Simulation of Plasma Sheet Thinning Due

to Loss of Near-Earth Magnetotail Plasma

Rudolf TRETLER1, Tomo TATSUNO1,Keisuke HOSOKAWA1

1. Department of Communication Engineering and Informatics, University ofElectro-Communications, Tokyo, Japan

The mechanisms behind auroral breakup, a sudden increase of auroral strengthduring magnetospheric substorms, are not yet entirely understood. While it isknown [1] that the three main events in this process are a) magnetotail recon-nection, b) cross tail current reduction, and c) auroral breakup, their exact orderhas not been conclusively determined. There are two main competing models,the Near-Earth Neutral Line (NENL) model [2] and the Current Disruption (CD)model [3].

In the NENL model, a reconnection of the magnetic field lines in the magneto-tail creates jets of plasma that flow earthwards and tailwards. The earthward jetcauses a decrease in the cross tail current, and then enters into the high-latitudeatmosphere, where it causes the auroral breakup.

In the CD model, a current disruption instability reduces the cross tail current,breaking the balance of the near-Earth magnetotail plasma, which enters theatmosphere and causes auroral breakup. This plasma loss induces a rarefactionwave in the plasma sheet, which propagates tailwards, resulting in plasma sheetthinning in the magnetotail, and eventually leads to magnetotail reconnection.

In this research, we consider the CD model. We run a magnetohydrodynamics(MHD) simulation of a rarefaction wave in a 2D model of the plasma sheet, basedon the 1D model by Chao et al. [4], to examine the effects of the magnetic lobeson the dynamics of the system. The rarefaction wave is induced by an initialearthward flow on the near-Earth side of the plasma sheet.

The results show that in the first few moments of the event the rarefactionwave, which is supposed to be a signature of the CD model, is weakened to suchan extent that it essentially disappears. The thinning of the plasma sheet prop-agates independently, though at a reduced velocity dependent on the conditionof the magnetic lobes. This suggests that the actual dynamics may be stronglyinfluenced, or even dominated, by the effects that could not be accounted for inthe 1D model.

[1] K. Schindler, Space Sci. Rev. 17(2-4), pp. 589–614 (1975)[2] A. Nishida, Magnetospheric Physics, pp. 35–44 (1974)[3] A. T. Lui et al., J. Geophys. Res. 82(10), pp. 1547–1565 (1977)[4] J. K. Chao et al., Planet. Space Sci. 25, pp. 703–710 (1977)

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Simulation study of trapped electron effects on

positron acceleration by a shock wave in an

electron-ion-positron plasma

Mieko TOIDA1

1. National Institute for Fusion Scienece

Particle acceleration by a shock wave is an important issue in astrophycical plas-mas. Because positrons are believed to exist around pulsers, we study positronacceleration by a shock wave using particle simulations.

Some electrons can be trapped by a magnetosonic shock wave propagatingobliquely to an external magnetic field and can excite electromagnetic fluctua-tions with finite wavenumbers along the shock front. We study effects of thefluctuations along the shock front on positron acceleration by an oblique shockwave in an electron-ion-positron plasma with a small positron density using a 2D(two space coordinates and three velocities), relativistic, electromagnetic particlesimulation and calucuation of test particles. It is shown that becuase of the fluc-tuations along the shock front, some positrons are accelerated to higher energiesand the accelerated positrons are distributed over a wide region from the up-stream to the downstream of the shock wave. The number of the positrons thatgain energies from the transverse electric field more than the parallel electric fieldis increased due the fluctuations along the shock front. Effects of the acceleratedpositrions on the fluctuations are also discussed.

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Neutron Irradiation Effect on Critical Current

and Critical Magnetic Field of Nb3Sn wire

Arata NISHIMURA1,2, Yoshimitsu HISHINUMA2

1. Technical Institute of Physics and Chemistry, CAS, China2. National Institute for Fusion Science, Japan

The neutron irradiation tests of Nb3Sn wires were carried out at BR2. Two typesof wires were irradiated. One is bronze route and the other is internal tin process.The wires were irradiated up to 1.7E+23 n/m2 (over 0.1 MeV) and the criticalcurrent was measured using 15.5 T superconducting magnet and the variabletemperature insert at Oarai center in Tohoku University. The internal tin processwire showed the same tendency as the bronze route wire. However, the residualradiation of the internal tin process wire was very stronger than the bronze routeone. Generally, the critical current increased once and down to the original valueat around 8.0E+22 n/m2. When the irradiation test continues over 1.0E+23n/m2, the critical current decreased and the critical magnetic field decreaseddrastically. A lot of the neutron irradiation defects were created in the matrixand it resulted in the deterioration of the critical magnetic field. Therefore, thedegradation of the critical current strongly depends on the magnetic field in caseof heavily irradiated wires. The irradiation data of the internal tin process wireand the degradation of the critical magnetic field are the world first observationsand the further investigations are succeeding.

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Optimization of electromagnetic particle

simulation code PASMO

Hiroaki Ohtani1,2, Ritoku Horiuchi1,2, Shunsuke Usami1,3

1. National Institute for Fusion Science, Japan2. SOKENDAI (The Graduate University for Advanced Studies), Japan3. The University of Tokyo, Japan

For performing a particle-in-cell (PIC) simulation code on a distributed memoryand multi-processor computer system with a distributed parallel algorithm, suchas domain decomposition, a load balance problem may appear, if the simulationdomain is decomposed uniformly. For example, in the simulation of magneticreconnection, the number density of plasma is much higher in the central plasmasheet than in the lobe region. There are several methods to avoid the load un-balance. One of them is the dynamic load-balancing algorithm by making eachcomputation node help another heavily loaded node, such as OhHelp [1]. On theother hand, in a thread parallel algorithm, there is other problem which decreasesthe computational efficiency. In PIC algorithm, the particles are treated as La-grangian and the field variables on the grids are treated as Eulerian. When thecharged density and current density on the grid points are calculated from thepositions and velocities of particles, or when the electromagnetic field worked onthe particle is calculated from the electromagnetic field on the grids, the randommemory access takes place because of the randomized particle positions. In orderto avoid the random memory access, the bucket sort algorithm is devised in orderto access the cash memory efficiently. Our three-dimensional electromagnetic PICcode PASMO [2], which decomposes three-dimensionally the simulation domainand introduces the charge conservation scheme [3], adopts the OhHelp library andbucket sort algorithm. Recently, we exchange the particle data array in order toaccess the memory continuously. In this paper, we report the performance of theoptimized PASMO code on the Plasma Simulator (Fujitsu FX100).

This work is performed on“ Plasma Simulator” (FUJITSU FX100) of NIFSwith the support and under the auspices of the NIFS Collaboration Researchprogram (NIFS18KNSS102, NIFS19KNXN378, and NIFS19KNTS058), and ispartially supported by ”Joint Usage/Research Center for Interdisciplinary Large-scale Information Infrastructures” in Japan. Development of some numericalcodes used in this work was supported in part by the ”Code development supportprogram” of Numerical Simulation Reactor research Project (NSRP), NIFS.

[1] H.Nakashima, Y.Miyake, H.Usui, and Y.Omura: Proc. Intl. Conf. Su-percomputing, (2009), 90-99. [2] H.Ohtani and R.Horiuchi: Plasma and FusionResearch, 4 (2009) 024. [3] T.Zh.Esirkepov: Computer Physics Communications,135, (2001) 144-153.

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Impacts of RMP fields on ELMs in a JT-60SA

plasma

Yasuhiro SUZUKI1,2, Go Matsunaga3, Hajime URANO3,Nobuyuki AIBA3, Shunsuke IDE3

1. National Institute for Fusion Science, National Institute of Natural Sicences2. SOKENDAI, The Graduate University for Advanced Studies3. National Institutes for Quantum and Radiological Science and Technology

In the JT-60SA experiments, the understanding and mitigation/suppression ofthe ELM (Edge Localized Mode) are important aims, because the energy loss andenergy fluxes by the ELM crash onto the divertor plate cause the melting andsevere damages to the material. In the JT-60SA, some scenarios to mitigate orsuppress the ELM are proposed. The Resonant Magnetic Perturbation (RMP) byEFCCs (Error Field Correction Coils) is one candidate to mitigate or suppress theELM in the JT-60SA. However, by using the RMP, how much the energy loss andenergy fluxes can be reduced is not considered qualitatively and quantitativelyyet.

In this study, the energy loss and energy fluxes for a scenario of the JT-60SA plasma, which is so-called the ITER-like scenario, is studied systematicallyby MIPS code, which is a three-dimensional (3D) simulation code based on thelinearized or nonlinear MHD equations. At first, patterns of RMP fields withdifferent toroidal mode numbers are considered. In JT-60SA, total 18 EFCCs of6 columns by 3 rows are installed. In that coil configuration, n=1 and 3 RMPfields can be produced with different phases. Since the penetration of RMP fieldsmight be depended on the toroidal mode number, n, and its phase, the patternand amplitude of RMP fields should be carefully studeid. An then, impacts ofRMP fields to ELMs in linear and nonlinear evolutions are studied. The MIPScode can include naturally the change of the magnetic topology by the RMP.We discuss the linear and nonlinear mode structures of ELMs on the stronglystochastic plasma boundary and the reduction of energy loss and energy fluxonto the divertor plate.

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Energy Partition and Temperature Anisotropy inMerging Processes of two

Spherical-Tokamak-type Plasmoids

Ritoku HORIUCHI1, Toseo MORITAKA1,Shunsuke USAMI1

1. National Institute for Fusion Science

The merging processes of spherical-tokamak-type plasmoids (STs), which areconfined in a rectangular conducting vessel, is investigated by means of two-dimensional PIC simulation [1,2]. A series of simulation runs with different massratios clarifies that a starting time of the ST merging is nearly given by a transittime for an ion sound wave to travel from an inner edge of each ST in an initialprofile to a reconnection point and a part of poloidal magnetic energy is trans-ferred to the ion and electron thermal energies at the approximate rate of 2:1during the ST merging process, which is almost independent of the mass ratioexcept for the smallest mass ratio case of Mi /Me=100. This transfer processleads to the increases in a parallel component of electron temperature and a per-pendicular component of ion temperature while keeping the other componentsalmost constant. This is because two-component electron distribution functionwith different velocity shifts along a toroidal magnetic field is formed around areconnection point when two STs merge. On the other hand, an ion distributionfunction, consisting of three components with different velocity shifts perpendic-ular to the toroidal magnetic field, is formed around the reconnection point in themerging phase. It is also found that a sharp peak appears impulsively in the elec-tron parallel temperature profile in the merging phase, which is consistent withthe MAST merging experiments [3]. The detailed mechanism will be discussedin the presentation.

Reference [1] R. Horiuchi, et al., Plasma Fusion Res., vol. 13, 3403035 (2018);DOI: 10.1585/pfr.13.3403035 [2] R. Horiuchi, et al., submitted to Phys. Plasmas,(2019). [3] H. Tanabe, et al., Nucl. Fusion 57, 056037 (2017).

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Day2 - November 6, 2019 (Wed)

09:30-10:20 PL2: Wolfgang STAUTNERSuperconducting technology and cryogenics for fusion reactors – Status and Trends

10:20-10:50 I2-01: Nagato YANAGIDesign and Development of HTS Magnet and Conductors for the Helical Fusion Reactor and Next-Generation Helical Experimental Device

10:50-11:10 Break

11:10-11:40 I2-02: Naoyuki AMEMIYAApplications of HTS magnets to particle accelerators

11:40-12:10 I2-03: Michael J. WOLFHigh Temperature Superconductor Cable in Conduit Conductors for Future Fusion Magnets

12:10-12:40 I2-04: Masaru TOMITANext generation railway systems using superconducting feeder cable

12:40-13:40 Lunch

13:40-15:40 Poster 2

15:40-16:00 O2-01: Weixi CHENDevelopment of Nondestructive Evaluation Technique for Mechanical Lap Joint Fabricated with High-Temperature Superconducting Conductor Using X-ray Microtomography

16:00-16:20 O2-02: Mahmoud BAKRCharacterization and Prospective Applications of a D-D Plasma Fusion from an Ultra-Compact NeutronGenerator based on Inertial Electrostatic Confinement Fusion Device at Kyoto University

16:20-16:50 I2-05: Takayuki WATANABEVisualization and Diagnostics in Thermal Plasma Processing

16:50-17:10 O2-03: Allen Vincent B. CATAPANGOPERATION OF Ar:H 2 O PLASMA IN A REACTIVE, DC MAGNETRON DISCHARGE FOR ZnO FILMDEPOSITION

17:10-17:30 Break

17:30-18:00 I2-06: Tomohiro NOZAKINonthermal plasma enabled catalysis towards sustainable methane conversion

18:00-18:30 I2-07: Akihiro SHIMIZUProgress of physics and engineering design study for quasi-axisymmetric stellarator CFQS

18:30-18:50 O2-04: Takanori MURASEEDDY CURRENT ANALYSIS ON VACUUM VESSEL OF CFQS QUASI-AXISYMMETRIC STELLARATOR

Superconducting technology and cryogenics forfusion reactors – Status and Trends

Wolfgang STAUTNER1

1. GE Global Research

This talk starts with briefly looking back into the history of superconducting mag-net technology development for fusion magnets starting in 1965, before we divedeep into present day technology of large-scale systems like Tore Supra, LCT,TFMC, ITER, LHD, W7-X, KSTAR and other. Superconducting technologyfor fusion magnets unlike MRI and NMR applications, has a range of differentchallenges to face, mainly due to the high operating current beyond 10 kA, thelarge size of the superconducting magnets and its magnetic plasma confinementrequirement at fields well exceeding 10 Tesla pushing low temperature supercon-ductor (LTS) technology to its operating limits. Besides, the magnet coil sizethis requires designing and manufacturing for high coil forces during cooldown,ramp and operation, significantly contributing to the overall system cost. A con-cise overview of appropriate conductor cooling scenarios and technologies is given(Overview) that have been developed for operating temperatures in the 1.8 to4 K range. Milestones highlight the remarkable achievement of superconduct-ing technology for fusion magnets. Superconducting low temperature technologyaspects from magnet design to manufacture as well as enhanced conductor tech-nology and material aspects are being discussed. We then leap into future fusionapplications where we see a paradigm change in magnetic fusion technology usinghigh temperature superconductors that would enable us to go well beyond 20 T(Overview of current HTS fusion projects). Typical HTS cabled concepts andconfigurations for fusion magnets are shown and their feasibility discussed. Con-sequently, future fusion systems will be more compact, modular, easier to cooland increase the efficiency of the fusion power plant.

PL2

Design and Development of HTS Magnet and

Conductors for the Helical Fusion Reactor and

Next-Generation Helical Experimental Device

Nagato YANAGI1,2, Toshiyuki MITO1,2,

Junichi MIYAZAWA1,2, Yuta ONODERA1,Naoki HIRANO1, Yoshiro NARUSHIMA1,2,Shinnosuke MATSUNAGA2, Satoshi ITO3,Hitoshi TAMURA1, Shinji HAMAGUCHI1,

Hidetoshi HASHIZUME3, Kazuya TAKAHATA1,2

1. National Institute for Fusion Science2. SOKENDAI (The Graduate University for Advanced Studies)3. Tohoku University

At NIFS, the Large Helical Device (LHD) has been successfully operated for morethan twenty years. Deuterium experiments have achieved the ion temperature of10 keV. The LHD-type helical fusion reactor FFHR-d1 has also been designedbased on the plasma confinement characteristics confirmed in LHD. Presently,extension of the LHD project is being proposed, and in parallel, discussion forthe post-LHD project has also been initiated. One of the candidates is to builda new device by employing the similar but more optimized heliotron magneticconfiguration, and by employing the High-Temperature Superconductor (HTS).For the large-current capacity HTS conductor, three candidates are now beingdeveloped. The STARS conductor, originally developed for FFHR-d1, has RE-BCO tapes simply stacked, and imbedded into a copper stabilizer and stainlesssteel (SS) reinforcement jacket. A 3-m long 100-kA-class conductor sample wasformerly tested successfully. A bridge-type mechanical lap joint technique withlow joint resistance has also been developed to make the joint-winding feasible.A 20-kA-class conductor is now being developed by selecting a suitable weldingmethod for the SS jacket. The FAIR conductor has a stack of REBCO tapesimbedded in a circular aluminum-alloy jacket. The stack of tapes is twisted to-gether with the aluminum-alloy jacket, which is welded by Friction Stir Welding(FSW). Short samples of 1-m length having different pitch length of twistingare being tested in liquid nitrogen for examining the basic features. The WISEconductor is formed by inserting a stack of REBCO tapes into a flexible metaltube and this conductor is wound into a coil. Then, the coil winding packageis impregnated by filling a low-melting temperature metal. The advantage ofthis conductor is its flexibility during the winding process. Two prototype coilsamples were fabricated and tested in liquid nitrogen.

I2-01

Applications of HTS Magnets to ParticleAccelerators

Naoyuki AMEMIYA1, Yusuke SOGABE1

1. Kyoto University

Particle Accelerators together with magnetic confinements of fusion plasma havebeen killer applications of superconducting magnets, because high field genera-tion in a large volume is only possible by a superconducting magnet. One ofthe front lines in magnet research is the applications of high Tc superconduc-tors (HTSs). The benefits of HTS magnets to particle accelerators include veryhigh field generation that is impossible by low Tc superconductor, high toleranceagainst thermal disturbance, and high efficiency if operated at high temperatureregion. Here, we focus on two examples of HTS magnet applications to particleaccelerators: magnets for accelerator systems for hadron therapy and magnets forrapid-cycling synchrotron for accelerator driven transmutation system. In mag-nets for medical applications such as hadron therapy, their reliabilities are veryimportant issues: a quench leads to the disruption of treatments. In acceleratordriven transmutation system, which treats nuclear waste, efficiency is critical fortheir practical introduction. In this presentation, we report our R&D on HTSmagnets toward these applications.

This work was supported in part by the MEXT under the Innovative Nu-clear Research and Development Program and in part by the JST under theS-Innovation Program.

I2-02

High Temperature Superconductor Cable in

Conduit Conductors for Future Fusion Magnets

Michael J. WOLF1, Walter H. FIETZ1,Reinhard HELLER1, Daniel NICKEL1,

Klaus-Peter WEISS1

1. Karlsruhe Institute of Technology (KIT)

Since their discovery in the late 1980’s, high-temperature superconductors (HTS)are considered as promising materials for the energy-efficient transmission ofelectric power or operation of magnetic coils at higher temperatures or highermagnetic fields compared to existing low-temperature superconductor (LTS) so-lutions. The operation of the magnet system of a future fusion reactor at highmagnetic fields (≥ 13 Tesla) or at high temperatures (≥ 5 Kelvin) compared toconventional LTS with good temperature margin are possible benefits using HTSin a future fusion power plant. In fusion machines, the magnet system relies onso-called cable-in-conduit conductors (CICC), i.e., high current superconductors,which are embedded in a stainless-steel jacket for mechanical support againstLorentz forces and actively cooled by a forced flow of coolant. Since severalyears, different approaches to form high-current CICC from HTS were realizedand the present status of HTS CICC will be introduced. In particular, the devel-opment of the HTS CrossConductor (HTS CroCo) will be presented. HTS CroCois a stacked-tape conductor with a high current density that can be fabricated inlong lengths and which can serve as a strand of such HTS CICC for the windingpack of a future toroidal field magnet. Recent work is focused on the combinedtheoretical and experimental investigation of sub-scale HTS CroCo triplet con-ductors. Investigations include the manufacturing of CICC conductors for fusionmagnets, the modeling and test of the electromechanical performance of individ-ual strands and triplets at high loads and the calculation of the performance ofthese triplet-CICC in case of quench. The present status of these developmentsand an outlook towards next steps will be given.

I2-03

Next generation railway systems usingsuperconducting feeder cable

Masaru TOMITA1

1. Railway Technical Research Institute

In Railway Technical Research Institute (RTRI), the superconducting cable forthe DC railway electrification system has been developed. Railway transportationis safety, convenience, economy and environmental friendliness. Electric railwaysystems are widely used in the world as well as in Japan and DC systems havebeen employed in a number of metropolitan areas. However, they have someproblems, such as cancelled regeneration and energy loss. From the viewpointof saving energy, we have proposed Next Generation DC Railway System us-ing superconducting feeder cables. Introducing superconducting feeder cablesinto railway feeding systems offers a lot of benefits: fewer substations, equaliza-tion of load among substations, reduction of cancelled regeneration and powertransmission loss, and reduction of electrolytic corrosion. Aiming to introducethe superconducting cable for railway systems, we started with fabricating andevaluating superconducting wire materials, and based upon various inquests andexamination results, superconducting feeder cable system had been introducedto the test track. After the system was checked from the aspect of cooling andcurrent properties, the train running test was succeeded.

I2-04

Development of Nondestructive Evaluation

Technique for Mechanical Lap Joint Fabricated

with High-Temperature Superconducting

Conductor Using X-ray Microtomography

Weixi CHEN1, Satoshi ITO1, Noritaka Yusa1,Hidetoshi Hashizume1

1. Graduate Schools of Engineering, Tohoku University

Joint winding of High-Temperature Superconducting (HTS) helical coil with con-ductor segments connected using mechanical joint is considered as a challengingoption for magnet fabrication in heliotoron-type fusion reactor FFHR [1]. StackedTapes Assembled in Rigid Structure (STARS) conductor, where simply-stackedHTS tapes are embedded in copper and stainless-steel jackets, have been pro-posed for use in helical-shaped conductor segments with easy-fabricable resistivelap joints. The mechanical lap joint of 100 kA class STARS conductors whereHTS tapes are pressed together with indium foil inserted was successfully achieved1.8 nΩ at 118 kA, 4.2 K in a previous study [2]. However, the joint resistance isevaluable only after cooling and energizing, which is not preferable for practicalapplication to a fusion reactor. Therefore, a nondestructive evaluation techniqueat room temperature is highly demanded to predict joint resistance at cryogenictemperature.The X-ray Computer Tomography (CT) was introduced to observethe contact interface at the joint on purpose to identify the factor that affectsjoint resistance [3].

The correlation between the contact area observed by X-ray CT and the elec-trical contact resistance at the contact interface of a single-layer-single-row jointwas investigated, and dispersive contact resistivity was found. The reason forthis result was considered to inhomogeneous distribution of fine-structure on thecontact interface. Furthermore, multi-layer-multi-row joint in the actual designlower the discriminability because of the X-ray attenuation. To develop a jointresistance prediction technique, we analyzed fine-structure distribution, whichhas finer scale than X-ray CT resolution, in the cross-section crossing contactinterface, and modified a method to keep discriminability of contact interface inmulti-layer-multi-row joint in this study. The result of analysis will be reportedin the presentation.

Reference[1] N. Yanagi, et al., Nucl. fusion, vol. 55, no. 5, 053021, Apr. 2015.[2] S. Ito, et al., Plasma Fusion res., vol. 9, 3405086, Jun. 2014.[3] S. Ito, et al., IEEE Trans. Appl. Supercond., vol. 26 no 2, March 2016.

O2-01

Characterization and Prospective Applications of

a D-D Plasma Fusion from an Ultra-Compact

Neutron Generator based on Inertial Electrostatic

Confinement Fusion Device at Kyoto University

Mahmoud BAKR1, Kai Masuda 2, Keisuke Mukai1

1. Institute of Advanced Energy, Kyoto University, Japan2. Rokkasho Fusion Institute, Fusion Energy Directorate, National Institutes

for Quantum and Radiological Science and Technology, (QST), Japan

The inertial electrostatic confinement fusion (IECF) is a method for producingnuclear fusion in which ions are confined inside concentric electrodes, which thefirst IECF was configured by Hirsch [1]. Our group has been developing andoperating spherical IECF neutron generators for basic research and applicationsfor more than two decades. The present work focuses on the characterizationand the prospective applications of the recently developed ultra-compact neutrongenerator based on IECF device. The cathode and anode of the IECF deviceare made of molybdenum and titanium in 6 and 17 cm diameters, respectively.The gross weight and hight of the system are 35 kg and 70 cm, respectively [2].The D-D fusion reaction takes place in the IECF device is used for neutronsproduction. So far, the neutron yield from the system is 9x107 n/s by applying5.25 kW (75 kV and 70 mA) [3]. Multedisplanary applications utilizing the IECFneutron generator, such as detection of special nuclear materials, investigation of10B for boron neutron capture therapy (BNCT) in the medical filed, X-ray andneutrons imaging for radiography filed are proposed. Preliminary experimentsusing the neutron generator for those applications are conducted, and the resultsare promising. The dependence of the IECF neutron yield on the applied voltage,current, and deuterium gas pressure will be presented. Also, the prospectiveapplications of the ultra-compact neutron source and the preliminary results willbe discussed in the conference.

References:[1] R.L. Hirsch, J. Appl. Phys. 38, 4522 (1967).[2] M. Bakr, et al., AIP, CAARI 2018, Taxes US, Aug. (2018).[3] M. Bakr, et al., J. Fusion Sci.Technol., V. 75, 6, pp. 479-486, (2019).

O2-02

Visualization and Diagnostics in Thermal Plasma

Processing

Takayuki WATANABE1, Manabu Tanaka1

1. Department of Chemical Engineering, Kyushu University

Thermal plasmas are expected to be utilized for a number of innovative indus-trial applications such as decomposition of harmful materials, recovery of use-ful materials from wastes, and synthesis of high-quality and high-performancenanoparticles. The advantages of thermal plasmas including high enthalpy toenhance reaction kinetics, high chemical reactivity, and oxidation or reductionatmospheres in accordance with required chemical reactions are beneficial forinnovative processing.

The experimental and modeling efforts on thermal plasma characteristics hasbeen devoted to industrial application. However, the thermal plasma character-istics remain to be explored in spite of these efforts. The electrode phenomenaare one of the most considerable issues, because it determines the processing per-formance in thermal plasmas. The objective of the study is to investigate thephysical and chemical phenomena in thermal plasma processing for industrialapplication.

A multiphase AC arc is one of the most attractive thermal plasmas due to itsadvantages such as large plasma volume with low gas velocity, which are favorablefor material processing. Other advantages compared with other thermal plasmasinclude high energy efficiency and low cost. Therefore, the multiphase AC arc hasbeen applied to an innovative material processing such as in-flight glass meltingtechnology and nanomaterial fabrication processes.

The electrode physics in the multiphase AC arc has been investigated based onhigh-speed visualization of the electrode phenomena. An optical system includingthe band-pass filters was combined with a high-speed camera to observe synchro-nized images for different wavelengths. The visualization of dynamic behavior ofthe metal droplet ejection from tungsten-based electrode was conducted by high-speed observation with band-pass filters. Thermal radiation from the dropletscan be observed after removal of strong emission from thermal plasmas throughthe band-pass filters.

The visualization of metal evaporation from the electrode was conducted bythe high-speed visualization system with band-pass filters, which wavelengthswere selected to observe the tungsten vapor in the arc.

The electrode temperature of the multiphase AC arc was estimated by thehigh-speed visualization system with band-pass filters, which wavelengths wereselected to observe the thermal radiation from the electrodes. Two color pyrom-etry was applied to the electrode temperature measurements.

I2-05

OPERATION OF Ar:H2O PLASMA IN A

REACTIVE, DC MAGNETRON DISCHARGE

FOR ZnO FILM DEPOSITION

Allen Vincent B. CATAPANG1, Marlo Nicole R. GILOS1,Magdaleno R. VASQUEZ Jr2, Motoi WADA1

1. Graduate School of Science and Engineering, Doshisha University, Kyotan-abe, Kyoto, Japan

2. Department of Mining, Metallurgical and Materials Engineering, Univer-sity of the Philippines, Diliman, Quezon City, Philippines

Water vapor (H2O) plasma can be utilized to deposit ZnO thin films via reac-tive, DC magnetron sputtering; the presence of hydrogen is expected to improvethe produced film properties. Using this process, we have succeeded to preparetransparent and conductive ZnO films. Control of the stable discharge during thedeposition process, however, is difficult due to the vapor pressure, and the con-densation or freezing of water in vacuum systems. Any feedback control based onthe discharge parameters alone is insufficient to realize the stable plasma process.

In this study, the reactive, DC magnetron sputtering of ZnO thin films wasperformed using a metallic Zn target (99.2%) and varying partial pressure ad-mixtures of argon (Ar) and H2O bled in from a distilled water reservoir. Thedischarge and plasma parameters of the Ar : H2O glow discharge were moni-tored throughout the deposition using a single Langmuir probe. The depositionprocess was carried out in a 215 x 144 mm, rectangular electrode system. Theintroduction of more water vapor into the system caused the electron tempera-ture to increase. The discharge voltage required to sustain the plasma was foundto increase at higher water vapor content, with the temporal decrease of pressureright after ignition of discharge. Water adhesion on the discharge chamber canbe the cause of discharge instability associated with higher water vapor content,and a compact system is being assembled to control the wall temperature to seeif the system realizes a more stable ZnO film plasma based deposition.

O2-03

Nonthermal plasma enabled catalysis towards

sustainable methane conversion

Tomohiro NOZAKI1

1. Tokyo Institute of Technology, Dept of Mechanical Engineering

We investigate CH4 reforming using nonthermal plasma-enhanced catalytic reac-tion. Analogous to electrolysis of water splitting, renewable electricity is used togenerate nonthermal plasma and the electrical energy is converted to the chemicalenergy of syngas via nonthermal plasma-assisted endothermic reactions. Syngaswould be upgraded to H2-rich gas or synthetic fuels depending on the end-usetechnology. Alternatively, liquid fuels such as gasoline and methanol are synthe-sized through well-established C1 chemistry for storage and transport purposes.Carbon-containing liquid fuels are particularly important because energy densityis 10-100 times greater than that of secondary batteries: transport and storagecapability of renewable energy is greatly improved, meanwhile CO2 capture andutilization are strengthened simultaneously. Currently, an electrochemical reac-tion is dominantly studied for renewable-to-chemical energy conversion. Besides,nonthermal plasma provides additional energy and material conversion pathways,contributing to an extended carbon and energy utilization network. This articlefocuses particularly on dielectric barrier discharge (DBD) and catalyst combina-tion for CH4 and CO2 conversion to syngas (CH4 + CO2 = 2CO + 2H2). First,the current status of CH4 conversion technology based on thermal catalysis isoverviewed. Basics of heterogeneous reaction, known as a LangmuirHinshelwoodmechanism, is reviewed for the better understanding of nonthermal plasma andsurface interaction as well as plasma-enhanced heterogeneous reactions. Pulsedreaction spectrometry is introduced as a powerful diagnostic approach of hetero-geneous reaction kinetics under the influence of nonthermal plasma. Moreover,nonthermal plasma-induced synergistic effect, as well as energy efficiency, arediscussed in relation to discharge properties of DBD as well as electron colli-sion kinetics, aiming for deep insight into plasma catalysis of methane reforming.Finally, concluding remark and future outlook are presented.

I2-06

Progress of physics and engineering design study

for quasi-axisymmetric stellarator CFQS

Akihiro SHIMIZU1, Shigeyoshi KINOSHITA1,Haifeng LIU2, Mitsutaka ISOBE1,3, Shoichi OKAMURA1,

Kunihiro OGAWA1,3, Motoki NAKATA1,3,Takanori MURASE1, Sho NAKAGAWA1,

Hiroyuki TANOUE1, Xianqu WANG2, Jie HUANG2,Guozhen XIONG2, Hai LIU2, Yuhong XU2,Changjian TANG4, Dapeng YIN5, Yi WAN5,

CFQS Team1,2,5

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences, Toki 509-5292, Japan

2. Institute of Fusion Science, School of Physical Science and Technology,Southwest Jiaotong University, Chengdu 610031, People ’s Republic of China

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

4. Physics Department, Sichuan University, Chengdu 610041, People ’s Re-public of China

5. Hefei Keye Electro Physical Equipment Manufacturing Co., Ltd, Hefei230000, People ’s Republic of China

The quasi-axisymmetric (QA) stellarator CFQS [1-5] is a new helical device, ofwhich physical and engineering design studies are now being performed as thejoint project of National Institute for Fusion Science in Japan and SouthwestJiaotong University in People ’s Republic of China. The physical design forthe CFQS has been almost completed. Major radius, magnetic field strength,aspect ratio, and toroidal periodic number of the CFQS are 1.0 m, 1.0 T, 4.0,and 2, respectively. The rotational transform profile is designed between 1/3 and2/5 to avoid low-mode rational surfaces resulting in large islands. The magneticwell property is achieved in entire range of minor radius, which leads to stablecharacter in MHD stability.

The engineering design of CFQS has been developed [5]. The modular coilsystem was designed by the NESCOIL code, which is composed of 16 modularcoils. Element conductor for modular coil is 8.5 mm x 8.5 mm sized copperhollow conductor, and total of 72 turns (6 x 12) for each modular coil. Themockup coil is now being constructed by Hefei Keye Co., Ltd. to check feasibilityof manufacturing and the achieved accuracy. The conductor winding processwas already finished. Subsequently, vacuum pressure impregnation (VPI) will be

I2-07

started soon. In this paper, current status of CFQS design study and modularcoil construction will be presented.

[1] H. Liu et al., Plasma Fusion Res. 13, 3405067 (2018).[2] A. Shimizu et al., Plasma Fusion Res. 13, 3403123 (2018).[3] Y. Xu et al., 27th IAEA Fusion Energy Conference, Ahmedabad, India,

EX/P5-23 (2018).[4] M. Isobe et al., Plasma Fusion Res. 14, 3402074 (2019).[5] S. Kinoshita et al., Plasma Fusion Res. 14, 3405097 (2019).

I2-07

EDDY CURRENT ANALYSIS ON VACUUM

VESSEL OF CFQS QUASI-AXISYMMETRIC

STELLARATOR

Takanori Murase1, Sho Nakagawa1,Shigeyoshi Kinoshita1, Akihiro Shimizu1,

Shoichi Okamura1, Mitsutaka Isobe1,2, Guozhen Xiong3,Yuhong Xu3, Haifeng Liu3, Hai Liu3, Dapeng Yin4,

Yi Wan4

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies)3. Institute of Fusion Science, School of Physical Science and Technology,

Southwest Jiaotong University4. Hefei Keye Electro Physical Equipment Manufacturing Co., Ltd

The construction of world’s newest quasi-axisymmetric stellarator CFQS is on-going as a joint project of National Institute for Fusion Science (NIFS) in Japanand Southwest Jiaotong University (SWJTU) in China [1]. In recent year, a lotof efforts have been put into the engineering design [2] on the basis of physicsdesign [3, 4], however the eddy current analysis of the CFQS was not carefullyinvestigated.

The CFQS coils consist of sixteen modular coils in total with four differenttypes, two pairs of the poloidal field coils (PFCs), and the toroidal field coils(TFCs). During the ramp up of the current in PFCs and TFCs, the eddy currentis induced on the CFQS vacuum vessel. In designing fusion devices, eddy currentshave been calculated in order to verify how long the eddy current affects themagnetic field configuration. In particular, the time constant of the eddy currentis evaluated because it determines the life time of eddy current after the end ofthe ramp up phase of PFCs/TFCs currents.

In this work, the finite element method software, ANSYS/Maxwell, was em-ployed to evaluate the time-varying eddy current induced by change of PFCs/TFCscurrent. Here, the eddy current is integrated over the poloidal cross section ofvacuum vessel to calculate the total current in toroidal direction. According tothe resulting eddy current, the time constant of the eddy current due to changeof PFC current is estimated to be 4.4 ms, which is sufficiently shorter than theflat top duration of modular coil current and typical discharge duration ( 100 ms)for the CFQS. Also, the time constant of eddy current caused by change of TFCcurrent is estimated. The estimated time constant is 2.4 ms, which is also muchshorter than the plasma duration. Therefore, it can be reasonably concluded that

O2-04

the effect of the eddy currents due to change of PFC and TFC currents on themagnetic field configuration is not significant during the plasma discharge in theCFQS.

Reference: [1] M. Isobe et al., Plasma Fusion Res. 14 (2019) 3402074. [2] S.Kinoshita et al., Plasma Fusion Res. 14 (2019) 3405097 [3] H. Liu et al., PlasmaFusion Res. 13 (2018) 3405067. [4] A. Shimizu et al., Plasma Fusion Res. 13(2018) 3403123.

O2-04

P2-01 Naohiro KASUYAApplication of numerical diagnostics to fluctuation simulations of torus plasmas

P2-02 James Hamilton Palmer RICEDevelopment of Langmuir Probe Diagnostic for Measurement of Scrape-Off Layer Conditions in RF-Driven Plasmas inTST-2

P2-03 Jacobo VARELAEffect of the tangential NBI current drive on the stability of LHD plasma

P2-04 Yongtae KOObservation of parametric decay instability in the TST-2 LHW driven plasmas

P2-05 Shinji KOBAYASHIFlow Velocity Fluctuation Measurements with Ultra-Fast Charge Exchange Recombination Spectroscopy in Heliotron J

P2-06 Ryohei TAKEHARADemonstration of Helicon Wave Current Drive for Application in Advanced Tokamak Regimes

P2-07 Kotaro IWASAKIMeasurement of the flow of Ohmic plasmas on the TST-2 spherical tokamak

P2-08 Linge ZANGDesign of the analyzer for a new E//B NPA and the neutral count estimation for HL-2A/2M tokamak

P2-09 Shinichiro KOJIMAComparison of parametric decay instability on Bernstein mode conversion between HFS and LFS injections in QUEST

P2-10 Yasuko KAWAMOTOQuantitative valuation of impurity influence on Z eff diagnostic for LHD

P2-11 Kyohei KONDOImprovement of Plasma Shape Reconstruction in UTST Plasma by Normalization of Boundary Integral Equations

P2-12 Arlee TAMMANDiagnostic System for First Phase of Thailand Tokamak 1

P2-13 Ming Lie YAOSimulation and Experimental Test Research on Hydrogen/Deuterium-α Visible Spectra Diagnosis Based on HL-2ATokamak

P2-14 Satoshi OHDACHITomography Reconstruction Method of the SX Emission Profile for the Next Generation Non-Circular Tokamaks

P2-15 Yasuji HAMADAStudy of quasi-steady and burst-like magnetic fluctuations in JIPPT-IIU tokamak high-beta plasmas using a heavy ionbeam probe

P2-16 Takao FUKUYAMADynamic behaviors of ionization waves focused on coherence resonance

P2-17 Qilin YUEMeasurement of Dynamic Retention with Fast Ejecting System of Targeted Sample

P2-18 Haruhisa NAKANOEFFECT OF OFF-AXIS ECRH ON PERFORMANCE OF HIGH ION TEMPERATURE DISCHARGE IN LHD

P2-19 Yasuto KONDOMeasurement of Radial Electric Field Using Doppler Reflectometer in High-Density Plasma of Heliotron J

P2-20 Suguru MASUZAKIInvestigation of the Ratio of Hydrogen Isotopes in Plasms in the Large Helical Device

P2-21 Tetsutarou OISHITemporal Evolution of Emissions from Tungsten Ions in Various Charge States Observed in Impurity Pellet InjectionExperiments in Large Helical Device

P2-22 Yutaka FUJIWARAValidation of Enhanced FIDASIM Using FIDA and Neutron Diagnostics for Fast-Ion Studies in the Large Helical Device

P2-23 Manabu TAKECHISimulation of the fast position control coils for JT-60SA Plasma Control

P2-24 Keigo OTAEvaluation of the momentum dependence of radial diffusion coefficient on density and heat transport in tokamaks

Poster 2 [November 6 (Wednesday) 13:40 - 15:40]

P2-25 Takeo NISHITANINeutronics assessment of a compact D-D neutron generator as a neutron source for the neutron calibration in magneticconfinement fusion devices

P2-26 Takuya GOTOImprovement of the performance of the helical fusion reactor FFHR by the modification of the helical coil winding law

P2-27 Makoto KOBAYASHIDesign of neutron spectrum-shaping assembly around the pneumatic tube-end in the LHD torus hall for the medicalresearch application

P2-28 Masahiro TANAKAAnalysis of hydrocarbons in the exhaust gas of a fusion test device using infrared absorption spectroscopy

P2-29 Junichi MIYAZAWAImproved Design of a Cartridge-Type Helical Blanket System for the Helical Fusion Reactor FFHR-b1

P2-30 Bong Guen HONGImpact of Neutronics and Plasma Physics Constraints on System Parameters of a Tokamak Fusion Reactor

P2-31 Jabir AL SALAMIDEVELOPMENT OF SPH METHOD FOR SIMULATION OF LIQUID METAL DIVERTORS

P2-32 Hitoshi TAMURATopology optimization study for magnet support in helical fusion reactor

P2-33 Takeru OHGO Experimental study on the Plasma Irradiation to the Metal Pebble Flow in the TPDsheet-U

P2-34 Kunqi HUVisualization of the magnetic field lines in a large helical device

P2-35 Shuji KAMIONeutron Effect on the Single Crystal CVD Diamond NPA

P2-36 Naoko ASHIKAWADeuterium retention and permeation of metallic specimens exposed to divertor plasmas in KSTAR

P2-37 Taichi SEKIBehavior of a Tracer-Containing Compact Toroid in a Transverse Magnetic Field

P2-38 Ryota TAKENAKAReconstruction for Microwave Holography using 3D-numerical-calcurated reflection Wave

P2-39 Kazuya ICHIMURAEvaluation of the Gas Pressure in Divertor Simulation Experiments Seeded with Nitrogen-Hydrogen Mixed Radiator

P2-40 Jhoelle Roche M. GUHITSpectroscopic Studies of Magnetized Hydrogen Plasma via Duoplasmatron Ion Source on Near Metal Surfaces

P2-41 Makoto OYAAtomic and Molecular Processes in Plasma Decomposition Method of Hydrocarbon gas

P2-42 Aliena Mari MIRANDALaser desorption measurement of cesium adsorbed on a molybdenum plasma grid of a negative hydrogen ion source

P2-43 Masahiro HASUODynamics Observation of an Atmospheric Pressure and Low Temperature Helium Plasma Jet by Laser Spectroscopy on

the 2 3 S Atoms

P2-44 Takumi MIHARAExperimental investigation of electron acceleration process during high guide field magnetic reconnection in UTST

P2-45 Toshiki HARAEvaluation of spatial characteristics of divertor simulation plasma during impurity seedings in GAMMA 10/PDX

P2-46 Naruyuki UESHIMAFlow Structure Formed by Turbulence in PANTA

P2-47 Masashi KISAKIIon Mass Effect on Beam Deflection Compensation by Aperture Displacement and Application of Ferromagnetic Materialas an Alternative Compensation Technique

P2-48 Yamato TSUNOKAKEConfinement of D-3He fusion product particles in a non-adiabatic trap

P2-49 Yuto SUGIKIQuasi-equilibrium calculation with external magnetic flux decay in field-reversed configuration plasma

P2-50 Kojiro SEKIGUCHIHollow cathode discharge experiment applying magnetic field of permanent magnets

P2-51 Ryo MATSUMOTOParticle simulation for high-density hollow cathode discharge

P2-52 Keito HANADevelopment of Cs-free negative ion source by sheet plasma

P2-53 Hiroshi SHIMIZUNumerical analysis of quantum mechanical E×B drift in non-uniform electric fields

P2-54 Keiji FUIJTAGLOBAL CALCULATION OF IMPURITY TRANSPORT INCLUDING THE VARIATION OF ELECTROSTATICPOTENTIAL ON THE FLUX SURFACE IN HELICAL PLASMAS

P2-55 Shunsuke USAMIStrange Shapes of Ion Velocity Distribution during Magnetic Reconnection in the Presence of a Guide Field

P2-56 Hiroto MATSUURAObservation of Heat Flux Time Evolution Carried by Reentering Fast Ions

P2-57 Trang LEParticle Simulation of Divertor Plasma with Electrical Biasing

P2-58 Atsushi FUKUYAMAComparison of integral-form and differential-form of dielectric tensor in kinetic full-wave analysis of cyclotron waves intokamak plasmas

P2-59 Anggi Budi KURNIAWANEstimating ripple transport of fast tritons by D-D fusion in JT-60SA tokamak

P2-61 Malik IDOUAKASSNumerical Investigation of Energetic Particle Driven Interchange Mode in LHD

P2-62 Atsushi M. ITORefinement of Interatomic Potential for Medium Energy Atomic Collision

P2-63 Keisuke ARAKINonlinear energy transfer between parity reversal invariant subspaces in incompressible Hall magnetohydrodynamicturbulence

P2-64 Atsushi ITOParameter dependence of equilibrium with flow in reduced MHD models

P2-65 Ryosuke SEKIEvaluation of Pressure Anisotropy derived NBI by using the Monte Carlo code

P2-66 Akira MATSUMOTOConvergence-property Improvement of k-skip CG and k-skip MrR

P2-67 Seiki SAITOModel of hydrogen recycling on divertor by molecular dynamics simulation for neutral transport analysis in LHD

P2-68 Yuya MORISHITAIntegrated Transport Simulation of LHD Plasma Applying the Ensemble Kalman Smoother

P2-69 Chio Z. CHENGDiscovery of Alfven-Slow Eigenmodes in Tokamaks

P2-70 Tetsuhiro OBANAInvestigation into quench detection for a mutli-stacked pancake coil wound with Nb3Sn CIC conductors

P2-71 Shinsaku IMAGAWAComprehensive Investigation of 25 Events of Propagation of Normal-zones in the LHD Helical Coils

P2-72 Tomohiro KAWASHIMAAC dissipation current of dielectric material at high electric field in cryogenic

P2-73 Kazuki YAMADAAssessment of electrical insulation performance of cryogenic fluids using partial discharge waveform

P2-74 Hidetoshi OGUROComparison of the mechanical and superconducting properties for various superconducting wires

P2-75 Kyohei YAMADA Tape Shaped Nb3Al Conductor toward Large Helical Coils for Future Fusion System

P2-76 Yoshiro NARUSHIMAInitial studies on large-current high-temperature superconductor WISE and its application to helical fusion devices

P2-77 Naoki HIRANOFeasibility study of HTS coil cooling assist technology by magnetic refrigeration

P2-78 Yuta ONODERANon-destructive inspection of local defect in HTS conductor by using magnetization method

P2-79 Conor PERKSThe Role Of Ionization Versus Transport In Setting Plasma Density Profiles In LAPD

P2-80 Justin Ty COHENThe Use of Divertor End Plates as Diagnostics in the Princeton Field Reversed Configuration II

Application of numerical diagnostics to

fluctuation simulations of torus plasmas

Naohiro KASUYA1,2, Takanori ONOMICHI2,Takeshi IDO3, Masanori NUNAMI3, Shinichiro TODA3,

Masatoshi YAGI4, Yoshihiko NAGASHIMA1,2,Akihide FUJISAWA1,2

1. Research Institute for Applied Mechanics, Kyushu University2. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-

versity3. National Institute for Fusion Science4. National Institute for Quantum and Radiological Science and Technology

Plasma turbulence plays an important role on transport in torus plasmas. Sev-eral kinds of turbulent structures are formed, so one-point measurement is notenough to capture their spatio-temporal features. Three-dimensional simulationsshow global nature of plasma turbulence, and their comparison with experimentscan accelerate understanding of plasma transport mechanisms. We have beencarrying out simulations of turbulence diagnostics in experimental devices [1]. Inthis presentation, the heavy ion beam probe (HIBP) measurement is simulated inPLATO tokamak. PLATO tokamak is now under construction in Kyushu univer-sity, and will install many diagnostics to observe global nature of plasma turbu-lence [2]. Three HIBPs [3] are planned to observe spatio-temporal patterns of thepotential and density fluctuations inside the plasma. For HIBP simulation, notonly the static configuration of the plasma device including the equilibrium butalso dynamics of plasma turbulence are taken into account. Three-dimensionaltime-series data from a 4-field reduced MHD model [4] are used for the numericaldiagnostic. Nonlinear couplings between MHD modes as tearing instability anddrift wave modes, which have different spatial scales, can be calculated. HIBP sig-nals from numerical diagnostics are compared between stronger and weaker fluc-tuation cases. The developed code is flexible to apply to other three-dimensionalturbulence simulations and device configurations.

[1] N. Kasuya, et al., Plasma Sci. Tech. 13 (2011) 326.[2] A. Fujisawa, AIP Conf. Proc. 1993 (2018) 020011.[3] T. Ido, et al., Rev. Sci. Instrum. 77 (2006) 10F523.[4] M. Yagi, et al., Plasma Fusion Res. 2 (2007) 025.

P2-01

Development of Langmuir Probe Diagnostic for

Measurement of Scrape-Off Layer Conditions in

RF-Driven Plasmas in TST-2

James Hamilton Palmer RICE1, Naoto TSUJII1,Yuichi TAKASE1, Akira EJIRI1, Osamu WATANABE1,Hibiki YAMAZAKI1, Yi PENG1, Kotaro IWASAKI1,Yuki AOI1, Yongtae KO1, Kyohei MATSUZAKI1,

Yuki OSAWA1, Tetsuya FUJIMURU1, Hanzheng LI1

1. The University of Tokyo, Kashiwa 277-8561, Japan

A new Langmuir probe has been designed and installed in TST-2 for mea-surement of Scrape-Off Layer conditions (SOL). Non-inductive current drive isconsidered essential for spherical tokamak reactors. It has previously been shownthat a large amount of Lower Hybrid Wave (LHW) power is lost to the SOL byprocesses such as LH absorption [1] and SOL density fluctuations [2]. To betterunderstand how these processes affect the operation of TST-2, a full density pro-file of SOL conditions is necessary. The new probe was designed to have betterdurability of ceramic parts, larger probe area for measurements of low densitySOL plasma, and the possibility of Mach Probe-like analysis of flow velocity inthe SOL. It can also be used as an electrostatic probe during Radio frequency(RF) operation. It has been installed at the bottom side of the plasma, plac-ing the upper and lower sets of probes centred at 6 mm and 10 mm from theplasma edge respectively. Preliminary analyses of electron temperature, Te, showa clear response to RF operation. Both upstream and downstream (with re-spect to plasma current) probes exhibit Te

∼= 20 − 40 eV during RF and ∼ 10eV otherwise. The upstream and downstream probes show typical densities of0.5 − 1.0 × 1013 and 0.1 − 1.0 × 1013 m−3 respectively during LH experiments,reducing almost exponentially from 1.0 × 1013 m−3 to zero as plasma operationends. These measurements show the presence of SOL plasma exclusively belowthe cut-off density for the 200MHz LHW (5 × 1014 m−3) in TST-2, thus LHWcan propagate through the SOL plasma.

[1] N. Tsujii et al., Nulc. Fusion 57, 126032 (2017)[2] E.H. Martin et al., Nucl. Fusion 59 076006 (2019)

P2-02

Effect of the tangential NBI current drive on the

stability of LHD plasma

Jacobo VARELA 1, W. A. Cooper2, K. Nagaoka1,K. Y. Watanabe1, D. A. Spong3, L. Garcia4,

A. Cappa5, A. Azegami6

1. National Institute for Fusion Science, National Institute of Natural Science,Toki, 509-5292, Japan

2. Ecole Polytechnique de Lausanne (EPFL), Centre de Recherches en Physiquedes Plasma (CRPP), CH-1015 Lausanne, Switzerland

3. Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-8071, USA4. Universidad Carlos III de Madrid, 28911 Leganes, Madrid, Spain5. Laboratorio Nacional de Fusion CIEMAT, Madrid, Spain6. Nagoya University

The aim of present study is to analyze the stability of the pressure gradient drivenmodes (PM) and Alfven eigenmodes (AE) in the LHD plasma if the rotationaltransform profile is modified by the current drive of the tangential neutral beaminjectors (NBI). The analysis is performed using the code FAR3d that solves thereduced MHD equations describing the linear evolution of the poloidal flux andthe toroidal component of the vorticity in a full 3D system, coupled with equationsof density and parallel velocity moments for the energetic particle (EP) species,including the effect of the acoustic modes. The Landau damping and resonantdestabilization effects are added via the closure relation. On-axis and off-axis NBIcurrent drive modifies the rotational transform; this becomes strongly distortedas the intensity of the neutral beam current drive (NBCD) increases, leadingto wider continuum gaps and modifying the magnetic shear. The simulationswith on-axis NBI injection show that a ctr-NBCD in inward shifted and defaultconfigurations leads to a lower growth rate of the PM, although strong n=1 and2 AEs can be destabilized. For the outward shifted configurations, a co-NBCDimproves the AEs stability but the PM are further destabilized if the co- NBCDintensity is 30 kA/T. If the NBI injection is off-axis, the plasma stability is notsignificantly improved due to the further destabilization of the AE/EPMs in themiddle and outer plasma region.

P2-03

Observation of parametric decay instability in the

TST-2 LHW driven plasmas

Yongtae KO1, Yuichi TAKASE1, Akira EJIRI1,

Naoto TSUJII1, Osamu WATANABE1,Hibiki YAMAZAKI1, Kotaro IWSAKI1, Peng YI1,Kyohei MATSUZAKI1, Yuki AOI1, James RICE1,

Yuki OSAWA1

1. The University of Tokyo

One of the issues in spherical tokamak study is to establish a non-inductive cur-rent drive method to remove center solenoid. Lower Hybrid Wave (LHW) is anattractive option to achieve non-inductive current drive. In the TST-2 sphericaltokamak, the plasma start-up using lower hybrid waves has been studied. In thisdevice, LHW with a frequency of 200.1 is excited by two capacitively coupledcombline antenna (CCCA) installed on the top side (at toroidal angle φ= -45° ) and on the outboard side (φ= 0 ° ) of the plasma. In order to achieve ahigh current drive efficiency, the Parametric Decay Instability (PDI) is one of theissues to be solved. Radio Frequency Magnetic Probes (RFMPs) are known fora robust and easy method to detect waves in plasma Scrape-off layer legion. 14RFMPs, which are located on the high field side (8-ch.), along radial directionon the bottom side (3-ch.) and on the low field side (3-ch.), were installed on onepoloidal cross section ( toroidallyφ= -60 ° ). The power spectrum shows sideband peaks around the excited frequency (200.1 MHz), and the frequency shiftδ f is in the range of ion cyclotron frequency. The toroidal field dependence ofδ f suggests that PDI occurs at the low field side of plasma when the outboardantenna is used. This is because outboard antenna is located at low field sideand PDI might occur during waves propagating from low field side to high fieldside. On the other hand, when the top launch antenna is used, probes on thehigh field side and low field side also observed PDI. According to the estimationwhere PDI occurs, results show PDI arose high field side and low field side. Highfield side PDI indicate that waves excited by top launch antenna propagate inthe inner side than magnetic axis. Low field side PDI waves passed from top tobottom and from bottom to outer edge. These result suggest that the energy ofwaves lost in outer plasma edge in either antenna launching cases.

P2-04

Flow Velocity Fluctuation Measurements with

Ultra-Fast Charge Exchange Recombination

Spectroscopy in Heliotron J

Shinji KOBAYASHI1, Xiangxun LU2,

Satoshi YAMAMOTO1, Katsumi IDA3,Tatsuya KOBAYASHI3, Mikiro YOSHINUMA3,

Shinichiro KADO1, Shinsuke OHSHIMA1,Hiroyuki OKADA1, Takashi MINAMI1,

Yuji NAKAMURA2, Akihiro ISHIZAWA2,Shin NISHIMURA3, Shigeru KONOSHIMA1,Tohru MIZUUCHI1, Kazunobu NAGASAKI1

1. Institute of Advanced Energy, Kyoto University2. Graduate School of Energy Science, Kyoto University3. National Institute for Fusion Science

For magnetically confined fusion plasmas, many diagnostic techniques to measurefluctuations in density/temperature/flow have been demonstrated to understandfast time-scale phenomena and heat/particle/momentum transports due to MHDand turbulent fluctuations. In this study, we have developed ultra-fast chargeexchange recombination spectroscopy (CXRS) for the flow velocity fluctuationmeasurements [1]. The ultra-fast CXRS system consists of a combination of ahigh-dispersion Echelle grating monochromator with photographic lens (Bunk-oukeiki: SPL-200) and a high speed camera with an 8 × 4 array of avalanchephotodiode (Fusion Instruments Kft: APDCAM). The effective F number of themonochromator is 2.9 and the dispersion is 0.064 nm/mm. Each APD elementhas 1.6 × 1.6 mm size and the maximum quantum efficiency of 87In order toclarify the performance of the system in calculating the fluctuation componentsof the flow velocity, calibration experiment was carried out using two sets ofmonochromator (Nikon: P-250) for the purpose to simulate the fluctuated CXRspectrum. An analytical method using fast Fourier transform has been appliedto evaluate the flow velocity fluctuation. We have demonstrated the flow velocityfluctuation measurements in the Heliotron J plasmas. In this case, the low fre-quency fluctuations were excited by applying resonant magnetic perturbation inthe He plasmas. The fluctuation level of the parallel flow velocity was observedto be several hundred m/s when the low frequency fluctuations were excited.

[1] S. Kobayashi, et al., Proc. 41st EPS conf. Plasma Phys. 41F, P5.108 (2017).

P2-05

Demonstration of Helicon Wave Current Drive

for Application in Advanced Tokamak Regimes

Ryohei TAKEHARA1, Takumi TAMAGAWA1,

Kaho SATAKE1, Takeshi FUKUDA1

1. Graduate School of Engineering, Osaka Univ

Fast waves at frequencies far above the ion cyclotron frequency and approach-ing the lower hybrid frequency, also called helicons, are well known to producehigh density plasmas eciently [1, 2]. Helicon wave propagates into the plasmacore without a density limit, and the weaker electron damping compared to theslow wave implies reduced damping near the plasma boundary. In addition, heli-con wave pattern in torus is dierent from cylinder due to toroidal curvature andmagnetic eld inhomogeneity, and consequently the plasma current is driven byJ × B, as proven experimentally and numerically in a simple torus device [4-6].Therefore, helicon wave is highlighted in this work for non-resonantly driving theplasma current by Landau damping as an alternative to lower hybrid currentdrive. The high frequency causes the whistler-like behavior of the wave powernearly following field lines and provides strong damping in the MHD region. Infact, it has been shown numerically that helicon wave can drive o-axis currentdrive in DIII-D and FNSF high performance discharges with high electron beta(Prater, 2014). However, to date it is not proven in experiments. Therefore,we have performed numerical analyses of electromagnetic field and extensive ex-periments to demonstrate helicon wave current drive (HWCD) in tokamak. As aresult, an increase in HWCD was conrmed numerically when the axial wave num-ber was large for cases m=0, which is equivalent to the toroidal mode number¡ 3 [7]. Therefore, we have prepared a double loop antenna and experimentallyobserved an evidence of HWCD for the first time in tokamak [8].

[1] R. W. Boswell, Plasma Phys. Controlled Fusion 26, 1147 (1984)[2] F. F. Chen and G. Chevalier, J. Vac. Sci. Technol. A 10, 1389 (1992)[3] R. Prater et al., Nucl. Fusion 54 083024 (2014)[4] Manash Kr. Paul and D. Bora, Phys. Plasma 12 062510 (2005); Phys. Plasma 14

082507 (2007); Appl Phys 105 013305(2009)[7] A. Ikeyama, Proc. 33rd Annual Meeting Japan Society of Plasma Science and

Nuclear Fusion Research, Tohoku University, Sendai, 29 Nov.-2 Dec. 2016.[8] M. Inomoto, H. Nozato and T. Fukuda, Proc 22nd Symposium on Plasma

Processing, P3-006, Nagoya, 26-28 Jan. 2005.

P2-06

Measurement of the flow of Ohmic plasmas on

the TST-2 spherical tokamak

Kotaro IWASAKI1, Akira Ejiri1, Yuichi Takase1,Naoto Tsujii1, Osamu Watanabe1, Hibiki Yamazaki1,

Yuki Aoi1, Kyohei Matsuzaki1, Yongtae Ko1,James Rice1, Yuki Osawa1

1. The University of Tokyo

On the TST-2 spherical tokamak, experiments of ohmic plasma are conducted,and Internal reconnection event (IRE)s, which are MHD instabilities specific tospherical tokamak, are often observed. IRE is considered to be the reconnectionevent between internal and external magnetic field lines, and after the reconnec-tion plasma flows outward through the reconnected line. Since rapid variationsin electron density and electron and ion temperatures are often observed duringan IRE, a drastic variation in also in the plasma flow is expected. The objec-tive of this study is the measurement of plasma flows related to IREs and theexamination of the relationship between the flow and the MHD fluctuations.

Ion flow is obtained from the wavelength shift of a line spectrum emittedfrom impurity ions. The emission is measured by a spectrometer, which has a16 channel photomultiplier tube array and its wavelength range is about 0.3 nm.The lines of CIII (464.74 nm), CV (227.09 nm) and OV (278.10 nm) are mainlymeasured. The typical toroidal flow velocity is about 10 km/s and the typicalpoloidal flow velocity is much smaller than toroidal’s, and it is about 1 km/s orless. Magnetic fluctuations are measured by the magnetic probes set on a poloidalcross section encircling the plasma, and the typical frequency of the fluctuationsis 10 kHz.

Magnetic fluctuations grow before IREs. There are two types: (1) fluctuationgrow slowly and (2) fluctuations appear abruptly and grow fast. In some cases,the toroidal flow varies largely (e.g. 5 km/s) just after an IRE. The relationshipbetween magnetic fluctuation and the flow variation is not simple and there seemsto be several types.

P2-07

Design of the analyzer for a new E//B NPA and

the neutral count estimation for HL-2A/2M

tokamak

Linge ZANG1, Yufan QU1, Weiping Lin2

1. Southwestern Institute of Physics, Chengdu, Sichuan, China2. Institute of Nuclear Science and Technology, Sichuan University, Chengdu,

Sichuan, China

An equipment to detect these escaped neutrals and provide the energy or/andmass spectrum is named neutral particle analyzer (NPA) [1]. Nowadays NPA ismainly used for energetic ion physics study [2], and fusion fuel density ratio (T/D)measurement [3]. A new E//B NPA capable of mass resolution and broad energyresolution is under design. The resolvable element is H and D. Energy range is20-200keV. The system is composed of a gas stripping cell, an electromagnetic an-alyzer, and a 2-dimentional detector module. As shown in Figure 1, the analyzeris composed of a permanent magnet and a condenser, to resolve both momentumand energy. Figure 1 (a) is the top view and Figure 1 (b) is the side view. Offsetof the particle trajectory in y direction is: y=y1+y2=meEx1(x1+2x2)/(2p2) mis the ion mass, e is the elementary charge, E is the“ average”electric field ofthe condenser, and p is the momentum of the particle. p is resolved with themagnetic field, and m is resolved with the detectors in y direction. As shown inFigure 1 (b), if the detector at position y detects H atom, then the detector at2y detect D atom. NPA will see the neutral beam to ensure that the countingrating is enough. The neutral count is estimated with analytical method withbeam on for HL-2A/2M tokamak. In the center of the plasma, atom density isdominated by the neutral beam, so the atom estimated from neutral beam data.The geometry factor is estimated with the detector size and distance between thedetector and the measurement point.

P2-08

Comparison of parametric decay instability on

Bernstein mode conversion between HFS and

LFS injections in QUEST

Shinichiro KOJIMA1, Hatem ELSERAFY1,Kazuaki HANADA2, Hiroshi IDEI2, Ryuya IKEZOE2,

Yoshihiko NAGASHIMA2, Makoto HASEGAWA2,Takumi ONCHI2, Kengoh KURODA2,

Kazuo NAKAMURA2, Masaharu FUKUYAMA1,Ryota YONEDA3, Masayuki ONO4, Akira EJIRI5,

Yuichi TAKASE5, Sadayoshi MURAKAMI6

1. Kyushu univ.2. Kyushu univ. RIAM.3. UCLA4. PPPL5. Univ. of Tokyo6. Kyoto univ.

In QUEST, ordinary (O)- extraordinary (X)- Bernstein (B) mode conversion fromlow magnetic field side (LFS) and X-B mode conversion from high magneticfield side (HFS) have been performed [1]. In experiment, X-mode HFS injectionachieved not only higher plasma density than LFS injection, but also over-denseplasma. Moreover, observed radio frequency (RF) leakage by RF monitor waslower in case of HFS injection than that of LFS one. This difference indicates thatHFS injection has higher RF absorption efficiency compared to LFS one. Para-metric decay instability (PDI) is caused by large amplitude electric field aroundupper hybrid resonance (UHR) layer during EBW excitation. Therefore, measur-ing PDI indicates the EBW excitation region in HFS and LFS injections. Twodecay waves exist. One decay wave is high frequency around UHR frequency.Another decay wave is low frequency around lower hybrid resonance (LHR) fre-quency. To investigate PDI experimentally, we prepared a movable Langmuirprobe set which is able to detect the decay wave at the LHR frequency by scan-ning the same plasma radially. This Langmuir probe set had five pins. Fourpins were prepared to measure electron density and temperature. One pin wasequipped to measure floating potential with sampling rate 1.25 GS/s using 300MHz bandwidth oscilloscope for PDI measurement. Spectrum analysis of float-ing potential signal was carried out by fast Fourier transfer. The predicted PDIfrequency was a few hundred MHz depending on plasma density. The spectrumanalysis ’s results showed spectra at around 100 300 MHz in both HFS and

P2-09

LFS. In HFS injection, these spectra had higher amplitude and wider bandwidththan the LFS ones. In both HFS and LFS injections, these spectra showed higheramplitude around UHR layer than other region. Thus, it is concluded that thesespectra are caused by PDI and are considered an experimental confirmation thatHFS injection has higher EBW conversion efficiency.

Reference:[1] H. Elserafy et al, Plasma Fusion Res, 14 1205038 (2019)

P2-09

Quantitative valuation of impurity influence on

Zeff diagnostic for LHD

Yasuko KAWAMOTO1, Motoshi GOTO1,2,Shigeru MORITA1,2, Tetsutarou OISHI1,2,

Masaki OSAKABE1,2, Tomohiro MORISAKI1,2

1. National Institute for Fusion Science2. SOKENDAI (The Graduate University for Advanced Studies)

The effective ion charge (Zeff ) is one of the most important factors for under-standing the plasma state. In many experiments, Zeff value is essential for quan-titative analyses of the data obtained, and a reliable diagnostic method adoptablefor all discharges of various plasma conditions, e.g., even for a discharge accom-panied by the large Shafranov shift, is required. In the Large Helical Device(LHD), studies on radial Zeff profile measurement using the multi-sight opticalspectrometer have been made [1]. In this paper, however, we focus our attentionespecially on the time-dependent changes Zeff measurement for various plasmacondition. The evaluation of Zeff is generally made by taking a ratio of the mea-sured emission intensity of bremsstrahlung to the synthetic one calculated underan assumption of a pure hydrogen plasma, namely, Zeff = 1. In the LHD, amagnetic equilibrium database, the so-called TSMAP [2], is now available for alldischarges. The Zeff measurement for LHD has started to employ TSMAP inderiving the electron temperature Te and density ne values on the line-of-sight ofthe bremsstrahlung observation so that the accuracy of the derived Zeff valueshas been improved. However, in the 19th cycle of LHD experiment, the presentdiagnostic is found to have a problem for discharges with ne lower than 5x1019

m−3, namely, Zeff is skyrocketing with decreasing ne. This would be caused bythe decrement in the bremsstrahlung emission intensity and a relative incrementin the line emission intensities by the edge impurities. In this paper, the methodto correlate the measured emissivity by using impurity radiation of edge plasmais proposed. In the 19th cycle, we employed the interference filter which is thecentral wavelength is 522.4 nm with a FWHM of 3 nm. From the 20th cycle, it isplanned to install the new interference filter which has smaller FWHM comparedto the conventional filter. At the presentation, we will discuss that this methodis effective for Zeff measured with the new filter.

[1] H. Zhou, S. Morita, M. Goto, and M. Chowdhuri, Rev. Sci. Instrum. 79,(2008).

[2] M. Emoto et al., Fusion Engineering and Design 87, 2076 (2012).

P2-10

Improvement of Plasma Shape Reconstruction in

UTST Plasma by Normalization of Boundary

Integral Equations

Kyohei KONDO1, Michiaki INOMOTO2,

Masafumi ITAGAKI3

1. Graduate School of Engineering, The University of Tokyo, Tokyo 113-8656,Japan

2. Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa277-8561, Japan

3. Hokkaido University, Sapporo 060-0808, Japan

Highly accurate estimation of the plasma shape which is identified by LastClosed Flux Surface (LCFS) is needed to reconstruct the current density profilein the toroidal plasmas. It is reported that the Cauchy Condition Surface (CCS)method has higher accuracy and robustness than the Fast Boundary Identification(FBI) method, which is generally incorporated in the EFIT code. In the numericalcalculation using JT-60 equilibrium model with high triangularity and high non-circularity, the test calculation for the study of the performance of the CCS andthe FBI method was conducted in terms of that LCFS was identified or not. Itwas reported that the CCS method completely identified LCFS in two modelswhereas FBI method failed with the probability of 2% for the high triangularitymodel and 20% for the high non-circularity model [1].

The CCS method is usually utilized to estimate the plasma shape in theequilibrium phase, not in the transient phase, e.g. start-up and disruption phases,because non-negligible eddy current is induced on the vacuum vessel wall. In theUniversity of Tokyo Spherical Tokamak (UTST) experiment, a novel sphericaltokamak (ST) start-up method by the merging formation has been developed. Asingle ST is dynamically formed by merging of two initial STs through magneticreconnection. Large eddy current is induced when the two initial STs approacheach other (merging phase) because the magnetic flux profile changes withinshorter period than the resistive decay time of the wall. The Modified CauchyCondition Surface (M-CCS) method demonstrated its applicability to the mergingSTs numerically [2] and is being implemented on the UTST experiment.

The M-CCS method successfully provided the vacuum field profile similarwith that directly measured by the two dimensional internal magnetic probe ar-ray, however, the LCFS position contained significant error of 30-40 mm from thedirectly measured LCFS position. In order to improve the reconstruction accu-racy, the effects of the weighting of boundary integral equations, the eddy current

P2-11

node positions and the sensor positions for the reconstruction accuracy have beeninvestigated numerically. In the UTST plasma, the outboard-side magnetic flux is20-30 times larger than the inboard-side magnetic flux and the solution obtainedby the singular value decomposition utilized in the M-CCS method overfits to thelarger signals. Thus, appropriate weighting is required for boundary equationsof 1) inboard-side flux, 2) outboard-side flux and 3) magnetic field. Test cal-culation using numerical models showed that better reconstruction results wereobtained when the equations are normalized to the average of each measuredvalue in which the contribution from the coil current is deducted. The weightingscheme is applied to the experimental data to develop the optimization methodto minimize the error between the reconstructed LCFS and directly measuredLCFS. The experimental results will be presented in the conference.[1]K. Kurihara, J. Plasma Fusion Res. Vol.91, No.1, 13-22 (2015)[2] T. Ushiki, M. Inomoto, M. Itagaki, Fusion Eng. Des 122, 35-41 (2017)

P2-11

Diagnostic System for First Phase of Thailand

Tokamak 1

Arlee TAMMAN1, Nopparit SOMBOONKITTICHAI2,Peera PONGKITIWANICHAKUL2

1. Thailand Institute of Nuclear Technology (Public Organization), NakhonNayok, 26120 Thailand

2. Department of Physics, Faculty of Science, Kasetsart University, Chatuchak,Bangkok, 10900 Thailand.

The aim of fusion research in Thailand is learning of plasma controlling and fu-sion technology. Thailand Tokamak 1 (TT1) is donated by ASIPP for the firsttokamak device in Thailand for training and improving skill and experience ondesign, development and operating. The first phase of the device is fabricated inASIPP, China. The diagnostic systems for this phase consist of a magnetic probe,HCN interferometer and CCDs diagnostic. The magnetic probes, plasma Currentcoil, OH Current, TF Current, VF Current, Poraidal field magnetic probe, MHDfor M and N modes, diamagnetic Loops, compensation Loops, voltage Loop andsaddle coil, are installed for measuring the basic parameters of the device. HCNinterferometer is mainly objective for plasma density measurement for plasmadensity feedback control. The image of plasma inside the vacuum chamber iscaptured by the high-speed camera with the frame rate up to 2000 fps at 1280 x1024 pixels. The plasma image is used for developing of offline-mode plasma po-sitioning measurement. The plasma position from the magnetic probe and imageprocessing are combined for plasma positioning feedback control that plane forinstalling in phase 2.

P2-12

Simulation and Experimental Test Research on

Hydrogen/Deuterium-α Visible Spectra

Diagnosis Based on HL-2A Tokamak

Ming Lie YAO, Jing WU, Peng CHEN

1. university of electronic science of china2. university of electronic science of china3. university of electronic science of china

As a priority diagnosis in the tokamak fusion device, the hydrogen-α and visiblespectroscopy could measure the distribution of impurities and hydrogen/deuteriumdensities in the divertor and scraping of layers. In the divertor and scraping layerarea, firstly, because the divertor and the wall regions ’strong light reflection,directly and accurately measure the region of impurities (helium, carbon, tung-sten, etc.) and hydrogen/deuterium distribution becomes difficult. Secondly, thestrong magnetic field leads to the existence of Zeeman splitting in the visible spec-trum of this region, and the magnetic field is different in edge locations, results inZeeman splitting different. Based on this, the hydrogen-α and visible spectrumdiagnosis system of HL-2A were established. The visible emission spectra weremeasured and fitted by simulation and experimental researches. The Zeeman Ef-fect was used to distinguish the reflected light information at different positions,thereby accurately impurities, hydrogen/deuterium in the divertor and scrapinglayer could be obtained for spatial and temporal distribution. It is important tostudy the steady-state operation of the fusion device ITER and CFETR by mea-suring the correlation result under different experimental discharge conditions,carrying out the constraint mode conversion, the fuel cycle, the impurity andthe plasma interaction researches. The author would thank Dr. Manfred for thediscussion of the test work on JET and design work for the ITER using Zeemanpaten to resolve the reflection light error on edge Dα diagnostics.

P2-13

Tomography Reconstruction Method of the SX

Emission Profile for the Next Generation

Non-Circular Tokamaks

Satoshi OHDACHI1, Manabu TAKECHI2

1. National Institute for Fusion Science2. National Institutes for Quantum and Radiological Science and Technology

Soft X-ray emission measurement using multi-channel detector is one of the mostbasic diagnostics for the fusion plasmas. Estimation of the equilibrium of plasma,detection of the MHD instabilities and the evaluation of the impurity content canbe made from the soft X-ray (SX) emission profile via tomographic reconstruction.Even in the next generation Tokamak devices, such as JT-60SA , where theradioactive radiation is quite large, this type measurement might be possibleusing scintillator based detector system [1]. Reliable tomographic reconstructionis quite important. In this study two types of newly developed reconstructionmethods are compared with the conventional Phillips-Tikhonov regularizationmethods [2]. First type of the method is the pattern expansion method [3]. Theemission profile is expanded with orthogonal / close to the orthogonal patternsand the coefficients for each pattern are determined by the L1 regularization.The second method is Neural Network Collocation Method (NNCM) [4, 5]. Akind of neuron whose input is the location at a cross section of plasma and whoseoutput is the local emission is introduced. From the learning so that the lineintegration of the local emission matches the experimentally measured integratedvalue, this neuron can be used as the special function giving the local emissionprofile. Learning process using Saito ’s Laplacian Eigen function [6, 7] will bealso introduced.

References:[1] T. Bando et. al., Rev. Sci. Instrum. 90, 013507 (2019).[2] N. Iwama et al., Appl. Phys. Lett. 54, 502 (1989).[3] S. Ohdachi, Plasma Fusion. Res. 14, 340208 (2019).[4] X.F. Ma, M. Fukuhara and T. Takeda, Nucl. Instrum. Methods in Phys.

Res. A 449, 366 (2000).[5] S. Ohdachi, T. F. Ming and M. Takechi,“Investigation of the tomographic

reconstruction methods for magnetically confined plasmas”, poster presentationat Asia Pacific Fusion Association meeting, Nov. 5-8 2013, Gyeongju, Korea.

[6] N. Saito, J. Plasma Fusion Res. 92, 905 (2016).[7] N. Saito, Appl. Comput. Harm. Anal. 25, 68 (2008)

P2-14

Study of quasi-steady and burst-like magnetic

fluctuations in JIPPT-IIU tokamak high-beta

plasmas using a heavy ion beam probe

Yasuji HAMADA1

1. National Institute for Fusion Science

Study of quasi-steady and burst-like magnetic fluctuations in JIPPT-IIU tokamakhigh-beta plasmas using a heavy ion beam probe

Hamada Y.*, Watari T., Nishizawa A., Narihara K., Ida K., Ido T., NakanishiH., and JIPPT-IIU Group

National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan E-mail: hamada.yasuji@toki-fs.jp

Abstract Using a heavy ion beam probe (HIBP) we are now able to mea-sure magnetic, density and potential fluctuations in the core region of JIPPT-IIUtokamak plasma [1,2]. It is performed through the multi-point measurement ofhigh-frequency 2-D (horizontal and vertical) displacement and intensity fluctu-ation of the secondary beam on the detector plates in the energy analyzer of aHIBP. By analyzing the obtained data on the magnetic fluctuations we show firsttime experimentally that a) microtearing mode (MTM) propagating into elec-tron diamagnetic drift direction (EDDD), discussed theoretically and studied bya computer simulation for a longtime, is very common in tokamak plasmas [1],b) kinetic ballooning mode (KBM) which propagates into IDDD begins to be ob-served in higher beta OH phase, in addition to the growth of MTM, c) in stronglyheated (by NBI) phase, the dominant magnetic fluctuations are superseded bythe high-frequency (damped) tearing mode (k=0) [2]. This is also predicted bythe large-scale computer simulation using gyrokinetic equations [3]. We also re-port that a small increase in the amount of gas-puffing to the tokamak plasmasinduces the change of frequent fishbone-like magnetic bursts (frequently observedin low-density NBI heated phase) to a quasi-steady magnetic fluctuations whichhave almost the same high-frequency components up to about 500 kHz as shownin Fig. 1 and 2

Figure 1 Figure 2 1) Y. Hamada et al., Nucl. Fusion 55 (2015) 043. 2) Y.Hamada et al, submitted to Nuclear Fusion. 3) D. R. Hatch et al., PRL 108,235002 (2012).

P2-15

Dynamic behaviors of ionization waves focused

on coherence resonance

Takao FUKUYAMA1, Rina YAMAGUCHI2,Hiroki KANZAKI3

1. Nagasaki University, 1-14 Bunkyo-machi, Nagasaki 852-8521, Japan2. Nagasaki University, 1-14 Bunkyo-machi, Nagasaki 852-8521, Japan3. Nagasaki University, 1-14 Bunkyo-machi, Nagasaki 852-8521, Japan

Dynamic behaviors of ionization waves [1, 2] focused on coherence resonance [3]in a glow discharge plasma are experimentally studied. As our previous study,in ref. [4, 5], dynamic behaviors of ionization waves influenced by feedback ina glow discharge plasma [4] and spatiotemporal structures formed in ionizationwaves [5] have been examined. Ionization waves in a glow discharge has a degreeof freedom with time and space, thus there are a verity of nonlinear studies suchas observation of chaos [6], controlling chaos [7], stochastic resonance [8], andso on. Coherence resonance means a phenomenon that excitable system makesits responses of oscillation most coherent for the external force. In a series ofexperiments, neon plasma is produced by a glow discharge between two electrodesafter the glass tube is evacuated to high vacuum. Spatiotemporal signals for thedata analysis are sampled as uctuations in the light intensity using a line-scancamera and photodiodes. The largest Lyapunov exponents are calculated from thetime series sampled in experiments, based on the algorithm advocated in reference[9]. As results of our studies, coherence resonance is caused by the interaction ofchaotic waves produced in two discharge tubes, application of external force, andinfluence of time-delayed feedback.

[1] M. Novak, Czech. J. Phys. 10, 954 (1960).[2] K. Ohe and S. Takeda, Contrib. Plasma Phys. 14, 55 (1974).[3] A. S. Pikovsky and J. Kurths, Phys. Rev. Lett. 78, 775 (1997).[4] T. Fukuyama, et. al., Plasma Fusion Res. 13, 3401073 (2018).[5] T. Fukuyama, et. al., Plasma Fusion Res. 14, 3401070 (2019).[6] W. X. Ding, et. al., Phys. Rev. E 55, 3769 (1997).[7] Th. Pierre, et. al, Phys. Rev. Lett. 76, 2290 (1996).[8] A. Dinklage, C. Wilke, and T. Klinger, Phys. Plasmas 6, 2968 (1999).[9] A. Wolf, et. al, Physica D 16, 285 (1985).

P2-16

Measurement of Dynamic Retention with Fast

Ejecting System of Targeted Sample

Qilin YUE1, Kazuaki HANADA2, Makoto OYA3,

Shogo MATSUO1, Shinichiro KOJIMA1, Hiroshi IDEI2,Takumi ONCHI2, Kengou KURODA2,

Naoaki YOSHIDA2, Yukai LIU4

1. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity

2. Research Institute for Applied Mechanics, Kyushu University3. Faculty of Engineering Sciences, Kyushu University4. Institute of Plasma Physics, Chinese Academy of Sciences

It is important to fulfill steady operation so as to realize nuclear fusion powergeneration. To reduce the amount of wall-stored tritium that is used as a fuelof nuclear fusion reaction, metals such as tungsten will be adopted for plasmafacing wall (PFW), instead of carbon. Recent ITER-like wall (ILW) experimentspromoted by JET could demonstrate a dramatic reduction of wall-stored fuelparticles with metallic PFWs. While the reduction of wall-stored fuel particlesclearly indicates that dynamic retention greatly enhances and play a significantrole in fuel particle balance.

In QUEST (Q-shu University Experiment with Steady-State Spherical Toka-mak) which has all-metal PFWs and a capability to produce steady-state op-eration (SSO) longer than 1h [1]. It has been found that redeposited layer ofseveral tens of nm thickness caused by the plasma wall interaction (PWI) playsan essential role in fuel particle balance and the wall temperature provides acontrollability of the global dynamic retention and fuel particle balance.

To comprehend the properties of the redeposited layer, each process of plasma-induced fuel particle in the material should be investigated. Fast Ejecting Systemof Targeted Sample called FESTA has been developed to execute this. FESTA isdesigned to pick up a sample using a sample-arm and can expose it to plasmas andrapidly extract it at the targeted timing. The sample is left in a test chamber andtwo gate-valves to isolate the chamber can immediately be closed after ejectionof the sample-arm. A series of operation is programmed and can be regulatedin conformity with experimental requests. Dynamic retention from a plasma-exposed plate made of SUS type 316L was quantitatively measured in the inboardlimiter configuration as increment of H2 partial pressure after closing the testchamber. In the FESTA experiment, some of H2 gas in the QUEST chamberbecome admixed into the measured H2 partial pressure in the test chamber, and

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the same operation without samples is used to monitor the unexpected partialH2 pressure. As a result, the significant difference of the H2 partial pressure inthe test chamber could be obtained and it comes from the dynamically releasedH2 gas from the plasma-exposed sample.

[1] K. Hanada, et al. Investigation of hydrogen recycling in long-durationdischarges and its modification with a hot wall in the spherical tokamak QUEST.Nuclear fusion, 10, 2017.

P2-17

EFFECT OF OFF-AXIS ECRH ON

PERFORMANCE OF HIGH ION

TEMPERATURE DISCHARGE IN LHD

Haruhisa NAKANO1,2, Hiromi TAKAHASHI1,2,Kiyofumi MUKAI1,2, Satoshi OHDACHI1,2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies)

Ion temperature (Ti) of 10 keV with Ion Transport Barrier (ITB) in Large HelicalDevice (LHD) has been achieved. This is one of the milestones toward realizinghelical fusion reactors. Electron heating is dominant in the reactor, where elec-tron temperature (Te) should be high as well as Ti. To obtain high Ti and Tesimultaneously in LHD plasmas, Te has been successfully extended in the highTi discharge with on-axis ECRH heating. However, the reduction of Ti was ob-served during the on-axis ECRH. This seems to be effects of a Te/Ti and/or anECRH on ion thermal confinement. Recently, the ECRH power deposition hasbeen more precisely positioned since microwave ray control of ECRH has beenimproved due to considering periphery LHD plasmas. To investigate the effects ofthe Te/Ti and ECRH, high Ti experiments without ECRH and with on-axis andoff-axis (rho = 0.6) ECRH have been conducted by using only BL1 for tangen-tial NBI. The clear Ti change in the edge region did not observed in the off-axisECRH discharge where the Te profile was characterized by the ECRH depositionposition. It is known that energetic-particle driven resistive-InterChange mode(EIC) events reduce core Ti. It was observed that the off-axis ECRH effectivelysuppressed the event number of the EIC, whose position is expected at aroundrho = 0.6 to 0.8 where iota is approximately one, without core Ti reduction suchas the on-axis ECRH. This means that the off-axis ECRH does not influencesignificantly on the ion heat transport in the core region. Furthermore, higher Tiwas observed in the off-axis ECRH plasma due to smaller number of EIC eventthan no ECRH and the on-axis ECRH plasmas. In this presentation, the effectof the off-axis ECRH in the high Ti plasmas on the density and temperatureprofiles, and ion and electron thermal transport will be discussed.

P2-18

Measurement of Radial Electric Field Using

Doppler Reflectometer in High-Density Plasma of

Heliotron J

Yasuto KONDO1, Shunsuke KENJO2,Takeshi FUKUDA1, Kazunobu NAGASAKI3,

Hiroyuki OKADA3, Takashi MINAMI3,Shinichiro KADO3, Shinji KOBAYASHI3,Satoshi YAMOTO4, Shinsuke OHSHIMA3,Tohru MIZUUCHI3, Shigeru KONOSHIMA3,

Tsuyoshi TOMITA5

1. Graduate School of Engineering, Osaka University2. School of Engineering, Osaka University3. Institute of Advanced Energy, Kyoto University4. National Institutes for Quantum and Radiological Science and Technology5. Graduate School of Energy Science, Kyoto University

In magnetically confined fusion plasmas, it is believed that plasma confinement isgoverned by turbulence, and radial electric field, Er, can affect plasma transportand confinement with the formation mechanism of transport barrier [1][2].In Heliotron J, Er has beenmeasured with a Doppler microwave reflectometer inthe peripheral region of a high-density plasma which is produced by neutral beaminjection heating (NBI) and high intense gas puffing. The ecperimental resultshows that Er is more negative as the energy confinement is improved and thestored energy, Wp, is increased. This shows that the radial electric field is relatedto the energy confinement and may have density dependence in Heliotron J. Thesign and intensity of Er is consistent with one measured with a charge exchangerecombination spectroscopy diagnostic[3]. A similar correlation between Er andWp has been observed in the core region of LHD plasmas based on a toroidalrotation measurement[4]. We will also analyze a plasma with NBI heating andpellet injection where confinement improvement is observed.

[1]A. Fujisawa et al., Phys. Rev. Lett. 82 2669 (1999)[2] K. Cromb et al., Phys. Rev. Lett. 95, 155003 (2005).[3] X. Lu et al., Plasma Fusion Res. 13,1202077 (2018)[4] M. Yoshinuma et al., Plasma Fusion Res. 3, S1014 (2008)

P2-19

Investigation of the Ratio of Hydrogen Isotopes

in Plasms in the Large Helical Device

Suguru MASUZAKI1, Motoshi GOTO1, Gen Motojima1,

the LHD Experiment Group

1. National Institute for Fusion Science

In the Large Helical Device (LHD), deuterium plasma experiment has been con-ducted since 2017 to reveal isotope effects on plasma properties in a three dimen-sional helical plasma. To understand mechanisms of isotope effects, to control theratio of hydrogen isotopes in plasmas is necessary. In this presentation, results ofthe investigation of the ratio of hydrogen isotopes in LHD plasmas are presented.

The ratio of hydrogen isotopes is not decided by only species of working gas,but also species of wall recycling neutrals. Therefore, a wall conditioning suchas a glow discharge which can cause an isotope exchange on surfaces of plasmafacing components is a tool for control of the ratio. In LHD, glow dischargeswith hydrogen gas and deuterium gas have been conducted before conductinghydrogen and deuterium plasma experiments, respectively. In the case of thefirst week of the first deuterium plasma experiment, deuterium glow dischargeswere conducted in nights after experiments. As the result, the ratio of D / (D +H) changed from 0.6 to more than 0.8 in four days.

To investigate the ratio of hydrogen isotopes, a visible spectroscopy and amass analyzer have been utilized. In the spectroscopy case, the ratio is evaluatedwith fitting of spectra in the range of 655 657 nm.

P2-20

Temporal Evolution of Emissions from Tungsten

Ions in Various Charge States Observed in

Impurity Pellet Injection Experiments in Large

Helical Device

Tetsutarou OISHI1,2, Shigeru MORITA1, Daiji KATO1,3,Izumi MURAKAMI1,2, Hiroyuki A. SAKAUE1,Yasuko KAWAMOTO1, Motoshi GOTO1,2,

the LHD Experiment Group1

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. Department of Fusion Science, SOKENDAI (Graduate University for Ad-vanced Studies)

3. Kyushu University

Spectroscopic studies for emissions released from tungsten ions have been in-tensively conducted in the Large Helical Device (LHD) for contribution to thetungsten transport study in tungsten-divertor fusion devices and for the expan-sion of experimental database of tungsten line emissions [1]. Tungsten ions aredistributed in the LHD plasma by injecting a pellet consisting of a small piece oftungsten metal wire enclosed by a carbon or polyethylene tube [2].Recently, line emissions from tungsten ions in low charge states, WIV-VII, havebeen identified in the vacuum ultraviolet (VUV) range of 500 - 1500 A[3]. Alsoin the extreme ultraviolet (EUV) range of 10 - 500 A, tungsten ions in low chargestates, WV-WVIII, as well as high charge states, WXLII-WXLVI, have beenidentified [4]. Based on the progress of the line identification, temporal evolutionof emission from tungsten ions in various charge states are demonstrated in thisstudy. When the pellet is injected, the electron temperature drops rapidly. Inthe latter half of the discharge, the tungsten line intensities from low to highcharge states increased in order because the electron temperature recovered bya continuous neutral beam heating. Their sequential increasing behavior is rea-sonable when considering the relationship between the electron temperature andtheir ionization energies.This work is partially supported by the LHD project financial support (ULPP010,ULPP038) and Grant-in-Aid for Young Scientists (B) (17K14426).1. S. Morita, C. F. Dong, M. Goto et al., AIP Conf. Proc. 1545, 143 (2013).2. X. L. Huang, S. Morita, T. Oishi et al., Rev. Sci. Instrum. 85, 11E818 (2014).3. T. Oishi, S. Morita, X. L. Huang et al., Phys. Scr. 91, 025602 (2016).4. Y. Liu, S. Morita, T. Oishi et al., Plasma Fusion Res. 13, 3402020 (2018).

P2-21

Validation of Enhanced FIDASIM Using FIDA

and Neutron Diagnostics for Fast-Ion Studies in

the Large Helical Device

Yutaka FUJIWARA1, Shuji KAMIO1,Hiroyuki YAMAGUCHI1, Ryosuke SEKI1,2,

Alvin V. GARCIA3, Luke STAGNER4, Hideo NUGA1,Kunihiro OGAWA1,2, Mitsutaka ISOBE1,2,

Masayuki YOKOYAMA1,2, William W. HEIDBRINK3,Masaki OSAKABE1,2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies)3. University of California, Irvine4. Oak Ridge Institute for Science and Education

A magnetic confinement fusion reactor requires the sustainment of plasma byenergetic alpha particles from the fusion reaction. Therefore, it is important tounderstand the behavior of fast-ions in the magnetic confinement device. Toinvestigate the behavior of fast-ions, a Fast-Ion D Alpha (FIDA) diagnostics,neutral particle analyzers, and neutron diagnostics were installed on the LargeHelical Device (LHD) [1-4]. In LHD, we have conducted hydrogen experimentalcampaigns since March 1998. To understand the physics of fast-ions and investi-gate isotope effects, the deuterium experimental campaigns have been conductedin LHD since March 2017. FIDASIM is a synthetic simulation code that wasoriginally developed in the tokamak type [7]. In 2018, FIDASIM was newly mod-ified in order to simulate in a three-dimensional magnetic configuration device.In the FIDA diagnostic, the Doppler-shifted D alpha lights from fast neutralsare utilized as signals of fast-ions [5,6]. An enhanced FIDASIM was applied onLHD to analyze the FIDA and neutron diagnostics. The FIDASIM requires thedistribution function, plasma profiles, magnetic equilibrium, and diagnostic ge-ometry. For LHD, a code is developed to produce the inputs files needed to runthe enhanced FIDASIM. We input the distribution functions which were calcu-lated by GNET, MORH, and MEGA which were the code to simulate the fast-ionbehavior. In order to validate the enhanced FIDASIM, measurement of the ra-dial profile of fast-ions using FIDA and neutron diagnostics were performed inMHD-quiescent plasmas. In the poster, we will describe the current status of theenhanced FIDASIM and, FIDA and neutron diagnostics results on LHD.

References

P2-22

[1] M. Osakabe, et al., Rev. Sci. Instrum. 79 (2008) 10E519.[2] T. Ito, et al., Plasma Fus. Res. 5 (2010) S2099.[3] Y. Fujiwara, et al., Plasma Fus. Res. 14 (2019) 3402129.[4] K. Ogawa, et al., Nucl. Fusion 58 (2019) 076017.[5] W. W. Heidbrink, et al., Plasma Phys. Control. Fusion 46 (2004) 1855.[6] Y. Luo, et al., Rev. Sci. Instrum. 78 (2007) 033505.[7] W.W. Heidbrink, et al., Commun. Comput. Phys. 10 (2011) 716.

P2-22

Simulation of the fast position control coils for

JT-60SA Plasma Control

Manabu TAKECHI

1. National Institutes for Quantum and Radiological Science and Technology,Naka, Ibaraki, 311-0193, Japan

JT-60SA has full super conducting coil system for toroidal field coils and poloidalfield coils. It is difficult to control fast plasma position and to stabilize verticalinstability by using only the super conducting poloidal field coils. Therefore, fastplasma position control coils (FPCCs) will be installed for control of vertical in-stability and plasma position. In this presentation, the fast plasma control systemis shown with simulation of plasma position control. Two FPCCs, each of whichconsists of 24 turns conductor with 120kAT, will be installed. Controllability ofplasma position with FPCC is studied for vertical instability, ELM and mini col-lapse. The feedback plasma position control simulation has been performed withthe linearized Grad-Shafranov equation and 3D FEM analysis [1]. FPCCs areable to control plasma position fast and reduce current and voltage requirementof the superconducting poloidal field coils from several thousand several 10 thou-sand voltage to several hundred voltage for the vertical instability stabilization.Each FPCC coil has individual power supply in order to induce vertical magneticfield for horizontal position control. Simulation of past plasma position controlwith and without horizontal position control has been performed for mini col-lapse. Each FPCC coil has individual power supply for the case with horizontalcontrol. FPCCs are connected in series and have opposite current direction forthe case without horizontal control. Maximum displacement of separatrix withhorizontal control is reduced by about half and recovers to the original positionfaster compared with those without horizontal control. Also Maximum variationof poloidal coil current is found to be reduced by about half of that without hor-izontal control. From these simulation, the specification of the FPCCs and thepower supply of FPCC have been decided. Simulation of the fast plasma positioncontrol with ELM has been performed for AC loss evaluation of superconduct-ing coils. The current variation and voltage of poloidal coils, FPCCs and eddycurrent of the stabilizing plate and the vacuum vessel are evaluated from thissimulation and we estimate the magnetic field variation at PF coils for AC losscalculation.

References [1] I. Senda et al, Nucl. Fusion 4247 568580 (2002).

P2-23

Evaluation of the momentum dependence of

radial diffusion coefficient on density and heat

transport in tokamaks

Keigo OTA1, Atsushi FUKUYAMA,

Sadayoshi MURAKAMI

1. Department of Nuclear Engineering, Kyoto University, Kyoto, Japan

Nowadays, International Thermonuclear Experimental Reactor(ITER) projecthas made progress with the aim of demonstrating the scientific and technicalfeasibility. In ITER project, one of the goals is to achieve the output power 10times higher than the input heating power. For this purpose, effective control ofburning plasma start up is also necessary. Thus, the simulation of the plasmadensity and temperature distribution during the start up phase including thecontribution of energetic particles is essential.

The aim of our research is to develop a tool to simulate the plasma behaviorincluding energetic particles and to establish an effective control scheme in theburning start up phase. In this research, we use a module of the TASK code,TASK/FP, a Fokker-Planck (FP) equation solver in 3D phase space. In the pre-vious analyses, we used the radial diffusion coefficient independent of the particlemomentum. In this case, the particle and the heat diffusion coefficients obtainedby integrating the FP equation are the same value. There are experimental ob-servations, however, that the temperature diffusion coefficient is larger than theparticle diffusion coefficient Therefore, we introduce the momentum dependenceof the radial diffusion coefficient in 3D space and evaluate the momentum depen-dence of the diffusion coefficient by comparing the ratio between the particle andthe heat diffusion coefficients.

We introduced the radial diffusion coefficient with the n-th power of the mo-mentum dependence and the momentum dependence due to quasi-linear diffusionand . As a result, we narrowed the range of n as the ratio between the particleand the temperature diffusion coefficients agreed with experimental observations.

P2-24

Neutronics assessment of a compact D-D neutron

generator as a neutron source for the neutron

calibration in magnetic confinement fusion

devices

Takeo NISHITANI1, Roman Rodionov2,Vitaly Krasilnikov3, Aakanksha Saxena3, Laura Core3,

Luciano Bertalot3

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. Project Center ITER3. ITER Organization

The calibration of neutron flux monitors against the total neutron emission rate inthe whole plasma is one of the most important issues of the neutron measurementon magnetic confinement fusion devices. For the device with deuterium plasmaoperation, a 252Cf neutron source is commonly used because the 252Cf neutronsource has an isotropic neutron emission with the average energy of 2.1 MeV.On the other hand, a compact D-T neutron generator has been used for D-Tplasma operating devices such as TFTR and JET, and will also be used in ITER.A compact D-D neutron generator has a possibility to be used at the neutroncalibration for the deuterium operation phase prior to the D-T operation phase.Because the neutron calibration with the compact D-D neutron generator is agood practice of that with the compact D-T neutron generator for the D-T oper-ation. However, the compact D-D neutron generator has a large anisotropy of theneutron mission due to the anisotropy of the D(d,n) 3He differential cross-sectionand the scattering/absorption by the neutron generator structure itself. By usingthe MCNP code, we evaluate the discrepancy of the detection efficiency obtainedin the neutron calibration using the compact D-D neutron generator from thetrue detection efficiency for the total neutron mission rate in the plasma. It isfound that the discrepancy is so large that we need precise neutronics calculationto derive the true detection efficiency from the detection efficiency obtained onthe neutron calibration using the compact D-D neutron generator.

P2-25

Improvement of the performance of the helical

fusion reactor FFHR by the modification of the

helical coil winding law

Takuya GOTO1, Junichi MIYAZAWA1,

Hitoshi TAMURA1, Nagato YANAGI1, Akio SAGARA1,The FFHR Design Group1

1. National Institute for Fusion Science

The LHD-type helical reactor has several advantages from the viewpoint of reactordesign as well as general advantages as a helical system without the need of thecurrent drive, for example, coil winding with a variation of the curvature, a flexibledivertor design by utilizing the rigid and robust divertor field structure, and ahigh maintainability with large port apertures. In the meantime, conceptualdesign of the LHD-type helical fusion reactor FFHR has shown the feasibilityof the steady-state operation with a fusion gain of 15 can be achieved withinthe physics condition which has been already confirmed in the LHD experiment.Even though this fusion gain is enough for the self-sufficiency of the electricity,further increase in the fusion gain is still desired to increase the operationalmargin and to explore a more attractive design with a larger net electric output.Optimization of the winding law of the helical coils is considered as one of thepromising methods to improve the core plasma performance while keeping theabove-mentioned advantages. The result of parametric analysis of the dependenceof core plasma performance (e.g, MHD stability, energy confinement) on thewinding law of the helical coils will be shown in the presentation.

P2-26

Design of neutron spectrum-shaping assembly

around the pneumatic tube-end in the LHD torus

hall for the medical research application

Makoto KOBAYASHI1,2, Kunihiro OGAWA1,2,Mitsutaka ISOBE1,2, Takeo NISHITANI1,

Teruki NISHIMURA1, Masaki OSAKABE1,2

1. National Institute for Fusion Science2. SOKENDAI (The Graduate University for Advanced Studies)

Deuterium experiment in the Large Helical Device (LHD) produces the neutronfield in the torus hall. The neutron flux in the torus hall near LHD duringplasma operation is estimated to be 109 n cm−2 s−1, which is close to the criteriaof neutron source for the Boron Neutron Capture Therapy (BNCT). Besides, awide neutron energy from thermal neutron (< 0.5 eV) to fast neutron (0.1 ∼

14 MeV) in the torus hall is not suitable for BNCT. Fast and thermal neutronsshould be excluded in BNCT because of the high radiation dose for patient andthe shallow penetration depth of neutron, respectively. Therefore, the neutronspectrum-shaping assembly to generate the neutron field, where epi-thermal neu-tron is dominant, is necessary. For the application of neutron in the LHD torushall to BNCT research, the installation of pneumatic tube is planned. In thisstudy, several designs of the neutron spectrum-shaping assembly for the pneu-matic tube-end were evaluated with respect to the neutron energy control andthe gamma-ray shielding by using MCNP6 (General Mote Carlo N-Particle code).MgF2, LiF, Cd, and Pb were selected as the consisting material of the neutronspectrum-shaping assembly by their characteristics of neuron capture efficiency,mass number, density, and gamma-ray shielding efficiency. The natural isotopicabundances of elements in these materials are considered. The size of the neutronspectrum-shaping assembly is 350× 350× 350 mm3 due to the space limitation inthe torus hall. Cd with the thickness of 1 mm is sufficiently thick to absorb mostof thermal neutron. As the fast-neutron moderator, LiF could moderate fast neu-tron more efficiently compared to MgF2. However, due to the larger cross-sectionof lithium to capture epi-thermal neutron, compared to that of magnesium, theflux of epi-thermal neutron significantly reduced. Therefore, MgF2 is selectedas a dominant moderator material. Pd can work to shield gamma-ray. ThickPb can reduce gamma-ray flux largely, although space for Pb in the neutronspectrum-shaping assembly is limited. In the presentation, the energy, dose, andflux of neutron and gamma-ray will be evaluated for several designs of the neu-tron spectrum-shaping assembly. Then, we will propose the optimized neutronspectrum-shaping assembly for the application to BNCT research in LHD.

P2-27

Analysis of hydrocarbons in the exhaust gas of a

fusion test device using infrared absorption

spectroscopy

Masahiro TANAKA1,2, Hiromi KATO1,Naoyuki SUZUKI1

1. National Institute for Fusion Science2. SOKENDAI (The Graduate University for Advanced Studies)

The understanding of hydrogen isotopes behavior in a fusion test device is im-portant from the viewpoint of the radiation safety and the development of fuelcycle process. The hydrogen plasma in the fusion test devices such as JT-60Uand JET forms the C-H bound by the interaction with carbon materials. Thus asmall amount of hydrocarbons is exhausted from the fusion test devices. In theprevious study, hydrocarbons in the exhaust gas from LHD were not detected bya micro gas chromatograph (detector: TCD, detection limit: 10 ppm). On theother hands, tritiated hydrocarbons could be detected by the use of tracer tritiumproduced by the D-D reaction. In this study, to analyze the low concentrationhydrocarbons in the exhaust gas, we propose the application of infrared absorp-tion spectroscopy. The analytical system consisted of an FT-IR instrument, longoptical path gas cell, and remote gas sampling system. As for the operating con-ditions of FT-IR, the numbers of the scan were 80 times, the optical path lengthwas 16 m, and the resolution of wave number was 1 cm-1. Then, the detectionlimit of hydrocarbons was 0.1 ppm. The analysis of hydrocarbons in the exhaustgas was conducted during the hydrogen and helium glow discharge cleaning op-eration, and the regeneration operation of cryogenic pumps for NBIs and closedhelical divertor. As the results, methane, ethane, ethylene as the hydrocarbonsand carbon monoxide could be detected in the exhaust gas. These hydrocarbonswere observed during the glow discharge cleaning operation. It indicated that thehydrocarbons were formed by the chemical sputtering of hydrogen glow discharge.The infrared absorption spectroscopy gave a positive prospect for the applicationto the exhaust gas analysis.

P2-28

Improved Design of a Cartridge-Type Helical

Blanket System for the Helical Fusion Reactor

FFHR-b1

Junichi MIYAZAWA1,2, Hitoshi TAMURA1,Teruya TANAKA1,2, Yukinori HAMAJI1,

Makoto KOBAYASHI1,2, Takanori MURASE1,Sho NAKAGAWA1, Takuya GOTO1,2,Nagato YANAGI1,2, Akio SAGARA1,2,

and the FFHR Design Group

1. National Institute for Fusion Science2. SOKENDAI

The cartridge-type helical blanket system called the CARDISTRY-B has beenproposed for the helical fusion reactor FFHR-d1. The CARDISTRY-B is aimedat easy construction and maintenance. However, there remain many issues in thedesign. For example, these include a large number of cartridges, no tangentialports, low compatibility with the helical divertor, and others. To solve theseissues, an improved design of the cartridge-type blanket is proposed and namedthe CARDISTRY-B2. This is composed of the tritium breeding blanket (BB)filled with the flowing molten salt and the neutron shielding blanket (SB) filledwith the tungsten-carbide. The number of the BB (or, SB) cartridges in the 1/10section of the full torus are reduced from 32 to 17 (or, from 44 to 17). Tangentialports for the neutral beam injection heating are newly equipped together withthe multi-purpose blanket cartridge. Compatibility with the helical divertor isalso improved. The BB cartridges are fixed on the fixing units set on the top ofthe SB. No other part of the BB is touching the SB except the fixing ribs. The SBcartridges surround the superconducting magnet coils and work as the cryostatinside the vacuum vessel. This makes the large bellows for the large maintenanceports unnecessary. Details of the CARDISTRY-B2 designed for a small helicalfusion reactor FFHR-b1 will be presented.

P2-29

Impact of Neutronics and Plasma Physics

Constraints on System Parameters of a Tokamak

Fusion Reactor

Bong Guen HONG1

1. Chonbuk National University

An optimal radial build and system parameters of a tokamak fusion reactor werefound by utilizing a new simulation method which couples a conventional tokamaksystems analysis and a radiation transport analysis. Neutron impacts on shield-ing and tritium breeding capability were self-consistently incorporated, togetherwith plasma physics and tokamak engineering constraints, which were moderatelyextrapolated from the ITER model. The minimum major radius to produce adesired fusion power was determined not only by the requirements on the shield-ing, but also by the requirements on the tritium breeding and the magnetic fluxdensity at the toroidal field (TF) coil. As the aspect ratio increased, contributionto tritium breeding from the inboard blanket increased and the minimum majorradius and the system size increased as the inboard blanket thickness increasedto meet the requirements for tritium self-sufficiency. The TF coil bore radiusalso increased to meet the requirements for the magnetic flux density at the TFcoil and thus the space for a central solenoid to provide volt-second for a plasmacurrent ramp-up increased. The auxiliary heating and current-drive power forthe given fusion power increased and the fusion energy gain Q decreased.

P2-30

DEVELOPMENT OF SPH METHOD FOR

SIMULATION OF LIQUID METAL

DIVERTORS

Jabir AL SALAMI1, Changhong HU2,Kazuaki HANADA2

1. Interdisciplinary Graduate School of Engineering Sciences,Kyushu University

2. Research Institute for Applied Mechanics, Kyushu University

Concerns regarding the structural integrity of solid plasma facing components(PFCs) at high energy fluxes have motivated research into liquid metals as PFCsfor use in DEMO-scale, and larger, reactors. Liquid PFCs are especially attrac-tive for use in divertors due to their potentially higher heat removal capacity, andtheir inherent immunity to issues such as neutron loading and erosion. To prop-erly fulfill its role, the flow in a liquid metal divertor will be highly turbulent, andthe surface exposed to extreme temperature gradients and strong magnetic fields.Due to the difficulty and cost associated with exclusively relying on experimentalwork for the design of such a device, numerical simulation has the potential tobecome an indispensable tool to expedite and simplify the evaluation and opti-mization of candidate designs. However, the operation of such devices involvesphenomena that span multiple disciplines of physics such as fluid dynamics, elec-tromagnetics, thermodynamics and plasma physics, making it difficult to modeland simulate. Smoothed Particle Hydrodynamics (SPH) is a Lagrangian, mesh-free numerical method with several attractive features that render it suitable foruse in multiphysics applications in general, and free-surface hydrodynamics inparticular. The purpose of this research is the development of an SPH-basedsolver aimed at the simulation of liquid metal divertors. In this presentationrecent progress will be reported. The formulations and algorithms used in thehydrodynamic aspect of the numerical tool developed will be described, includingcorrections and boundary conditions. The development of the capability to simu-late liquid metal magnetohydrodynamics (MHD) using the inductionless, or lowMagnetic Reynold’s number approximation will be discussed. Due to the rela-tively high surface tension coefficient liquid metals exhibit, and the near-vacuumenvironment found inside a reactor, a single-phase surface tension algorithm forSPH is applied. The developed numerical code is validated against standard theo-retical and experimental benchmarks and numerical results will be presented anddiscussed. There is still a long way to go before achieving practical simulationof liquid metal divertors and challenges and future perspectives on this researchwill be discussed.

P2-31

Topology optimization study for magnet support

in helical fusion reactor

Hitoshi TAMURA1, Takuya GOTO1,Junichi MIYAZAWA1, Teruya TANAKA1,

Nagato YANAGI1

1. National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, Japan

The geometrical structure in a helical fusion reactor is very complicated. The elec-tromagnetic force generated by the superconducting magnet system is so strongthat the magnet support structure has to be tough enough to sustain the forceand the magnetic field accuracy. The total weight of the magnet support is esti-mated to be several tens of thousand tons for a GW class reactor. The estimatedweight so far is several times greater than the ideal value which assumes thatthe total weight is proportional to the stored magnetic energy of the system. Inrecent years, the topology optimization method contributes developing a noveldesign that overrides the common designs. In this study, topology optimizationis applied to the magnet system in a helical fusion reactor to evaluate an effectof the optimization.

P2-32

Experimental study on the Plasma Irradiation to

the Metal Pebble Flow in the TPDsheet-U

OHGO Takeru1, GOTO Takuya1,2,MIYAZAWA Junichi1,2, TAKIMOTO Toshikio3,

TONEGAWA Akira3

1. The Graduate University for Advanced Studies2. National Institute for Fusion Science3. TOKAI University

A renewable divertor concept named the REVOLVER-D has been proposed forthe helical reactor FFHR. The REVOLVER-D consists of molten tin shower jetsinserted to the ergodic region of the plasma as a limiter/divertor and expected totolerate a high heat load larger than a few tens of MW per square meter. However,the strong Lorentz force on the molten tin shower jets is a problem. To solve thisproblem, we adopt a tin pebble flow instead of the molten tin flow to cut theelectric current path. In addition, the permissible heat load of the divertor targetcan be increased because of the heat of fusion and the decrease in the initialtemperature. A part of tin pebbles become melted by the heat load from theplasma. Therefore, we decided to melt all tin pebbles in the pool that catchesthe pebbles. Then it also becomes possible to absorb the drop impact of thepebbles. Next to the pool, tin pebbles are manufactured from the molten tin bythe shot tower method using a viscous silicone oil. Then pebbles are transportedmechanically and dropped to the plasma again. This concept is named the FusibleMetal Pebble Divertor (FMPD). The pebble behavior in the plasma is one of theissues on the FMPD. In the previous study on the ceramic pebble divertor, forcesact on the pebble flow (the coulomb force, the Lorentz force, the plasma pressure,and the plasma momentum flux.) have been evaluated. In this presentation, thedetailed scenario of the FMPD and experimental results on the pebble behaviorin the linear plasma device TPDsheet-U will be given.

P2-33

Visualization of the magnetic field lines in a large

helical device

Kunqi HU1, Koji KOYAMADA1, Hiroaki OHTANI2

1. Kyoto University2. National Institute for Fusion Science

Using the magnetic fields to confine the hot fusion fuel in the form of plasma canrealize controlled nuclear fusion, which is one of the non-pollution energy sources.A device called Force Free Helical Reactor (FFHR), which uses helical coils togenerate a magnetic trap, is being designed by the National Institute for FusionScience (NIFS) in Japan. Due to the complicated structure in FFHR, one of theprocesses is to predict the possible area the plasma will reach in this device foravoiding collision between the plasma and components. It is vital for the controlof its operation and also for diagnostic purposes.

General approaches follow the same way by comparing the plasma flux profilewith the sectional drawings of the device in a 2D graph. For lack of efficiencyand integrality, we need an overall perception of the plasma movement in three-dimensional space. In this article, we use the magnetic field lines traced by theVMEC code and MGTRC code to obtain the Poincare plot in each certain degreefrom toroidal orientation. Before the FFHR, there is a completed version, calledthe Large Helical Device (LHD). The existing data was observed from this one.Then, we try using the Alpha Shape algorithm to obtain the outermost profile ofeach Poincare plot and arrange these profiles into an annulus to build a donut-like3D model. The Larmor radius of the plasma will also be considered. Finally, thepolygon data will be transmitted into Shade 3D for further processing. If thisapproach is proved to be effective, it will be applied to the FFHR. This researchis a collaborative research with NIFS. We will continue the optimization of thismodel according to the feedback from NIFS. The model will then be used for thedesign of FFHR or some scientific visualizations.

P2-34

Neutron Effect on the Single Crystal CVD

Diamond NPA

Shuji KAMIO1, Yutaka FUJIWARA1,

Kunihiro OGAWA1,2, Makoto KOBAYASHI1,2,Mitsutaka ISOBE1,2, Ryosuke SEKI1, Hideo NUGA1,

Takeo NISHITANI1, Masaki OSAKABE1,2,Sho TOYAMA3, Misako MIWA3, Shigeo MATSUYAMA3

1. National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292,Japan

2. SOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292, Japan

3. Tohoku University, 6-6 Aoba, Aramaki, Aoba-ku, Sendai 980-8579, Japan

In FY2018, the neutral particle analyzer (NPA) array using single crystal CVD di-amond detectors (DNPA) was newly developed in the Large Helical Device (LHD)deuterium experiments for measuring the helically trapped energetic particles[1].As the diagnostic of NPA array, the signals of the neutron noise should be esti-mated from neutron measurement. In the experiment, the neutral particles wereshut out by closing the gate valve during the plasma discharges. By closing thegate valve, the signals from other than the neutral particles, such as neutron orgamma rays, can be obtained. The gate is made of thin a layer of stainless steel,which does not affect the neutrons. The amount of the signals due to the neutronnoise results in a linear relation with the neutron emission. Therefore, the effectof the neutrons can be estimated using the other available neutron diagnosticseven while the gate valve open. The neutron noise affects a much higher energyregion than the thermal noise because the signals due to neutrons come from theenergy loss by the elastic scattered neutrons inside the diamond. However, thecounting rate due to neutrons is very small even during the neutron emission rateof 1015 n/s which is the highest neutron emission rate in LHD experiments. Inorder to investigate the neutron effect in more detail, irradiation experiments toDNPA have been demonstrated at Fast Neutron Laboratory (FNL) in TohokuUnivesrity. In the experiments, similar peaks of the energy spectrum with theexperiments in LHD were observed, and the peaks were shifted by changing theenergy of irradiating neutrons. Also, in the experiment with the DNPA wideenergy range, the characteristic spectra were observed.

[1] S. Kamio et al., 2019 JINST 14 C08002 (2019)

P2-35

Deuterium retention and permeation of metallic

specimens exposed to divertor plasmas in KSTAR

Naoko ASHIKAWA1,2, Soo-Hyun SON3,Eunmam BANG3, Suk-ho HONG3

1. National Institute for Fusion Science2. SOKENDAI3. National Fusion Research Institute

In DEMO, understandings of permeated hydrogen isotopes are an important is-sue related to control of fuel cycles and safety for tritium leakage. Hence, esti-mations of detained and permeated hydrogen isotopes amounts are required. Ingeneral, hydrogen isotope permeation is measured by a quadrupole mass spec-trometry (QMS), which can detect widely integrated at regions around targetareas. Mainly permeation measurements were done at the first wall, howeverit is not sufficient data evaluated directly the hydrogen permeation amount inplasma facing walls under high heat load plasma irradiation in fusion devices.Metallic specimens by molybdenum (Mo) and tungsten (W) were exposed to di-vertor plasmas using diverter manipulator in KSTAR. After plasma exposures,depth profiles of deuterium and other impurities are measured by glow dischargeoptical emission spectroscopy (GD-OES), and amounts of detained deuterium ismeasured by nuclear reaction analysis (NRA) and elastic recoil detection analysis(ERDA). We find that high amounts of carbon are present at the strike point ofthe divertor plasma. The spatial profiles of the hydrogen isotopes, such as deu-terium and hydrogen differ from that of deposited carbon. The highest amountof energy deposited by the incident divertor flux occurs at the strike point of thedivertor plasma, and D retention at this position is lower than that in peripheralregions. In the poster, influence for retained deuterium at 2nd divertor plasmaexposure and permeated deuterium in the specimens are also discussed.

P2-36

Behavior of a Tracer-Containing Compact Toroid

in a Transverse Magnetic Field

Taichi SEKI1, Tomohiko ASAI1, Daichi KOBAYASHI1,Rika SASAKI1, Asuna MINAMIGI1, Hiroshi GOTA2,

Thomas ROCHE2, Tadafumi MATSUMOTO2,Toshiki TAKAHASHI3, Naoki TAMURA4,

Yoshiro NARUSHIMA4

1. College of Science and Technology, Nihon University, Tokyo 101-8308,Japan

2. TAE Technologies, Inc., Foothill Ranch, CA 92610, USA3. Graduate School of Science and Technology, Gunma University, Kiryu

376-8515, Japan4. National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292,

Japan

The study of impurity behavior in a hot core plasma is one of the impor-tant subjects for fusion reactor developments. Tracer-containing compact toroid(TCCT) has been developed as a new tracer injection method to study the accu-mulation and behavior of impurities in the magnetic confinement fusion plasmas[1]. A TCCT is generated and ejected by using the compact toroid (CT) injectorthat has successfully demonstrated CT injection fueling into a large field-reversedconfiguration (FRC) [2]. The CT injector can eject a spheromak-like magnetizedplasmoid at several hundreds of km/s. Tracer ions are supplied by the indepen-dently controlled tracer source attached on the CT injector. This tracer sourceconsists of a rod electrode made of tracer elements (e.g. tungsten, vanadium,carbon, etc.) and coaxially arranged cylindrical electrode, and the rod electrodeis sputtered by a discharge between those electrodes. The plasmoid containingtracer ions is accelerated and ejected by self-Lorentz force as TCCT. An experi-ment of TCCT injection into a transverse magnetic field, emulating a confinementmagnetic field of FRC plasma, has been conducted to investigate whether tracerions can be injected into the plasma without separation from the TCCT.[1] D. Kobayashi, T. Asai, S. Yamada et al., Rev. Sci. Instrum. 89, 10I111(2018)[2] T. Asai, T. Mastumoto, T. Roche et al., Nucl. Fusion 57, 076018 (2017)

P2-37

Reconstruction for Microwave Holography using

3D-numerical-calcurated reflection Wave

Ryota TAKENAKA1, Hayato TSUCHIYA2,

Ryo MANABE1, Naofumi IWAMA2,Shuji YAMAMOTO3, Soichiro YAMAGUCHI3,

Mayuko KOGA1

1. University of Hyogo2. National Institute for Fusion Science3. Kansai University

In the magnetized plasma confinement study, it is pointed out that the rela-tionship between a good confinement mode (H-mode) and turbulence phenomenais important.Because turbulence has complex structure, the structure should bedescribed in 2D/3D image.

To contribute experimental study, we propose microwave holography, whosemeasurement method is based on a lensless camera principle. A lensless mi-crowave holography can give us three-dimensional data with a wide field of view.The difference with visible light lensless camera is the number of pixels. The re-construction by small number of data is possible by regularization method. How-ever, the number of detector is much smaller than that of CCD due to wavelengthof microwave. So, the further verification of reconstruction in bad condition isnecessary. However, its experimentally development in a test bench is hard dueto its trial cost.

We try to reconstruct the target model from the observation values assumedby the electromagnetic field calculation. It has already been shown that the tar-get model can be reconstructed from the solution of the simple wave propagationequation[1]. Strictly the simple wave propagation equation is not expressed ac-curately because of the effect of quasioptics. The reconstructed image from theobserved values of the electromagnetic field calculation are compared with thereconstructed image from the solution of the wave propagation equation in thiswork. If it can be reconstructed from the observed values of the electromagneticfield calculation by same reconstruction method of simple wave propagation equa-tion, the reconstruction method would be applied to the reconstruction from theexperiment data.

[1] H.Tsuchiya, et al. Plasma Fusion Res. (2019), To be published.

P2-38

Evaluation of the Gas Pressure in Divertor

Simulation Experiments Seeded with

Nitrogen-Hydrogen Mixed Radiator

Kazuya ICHIMURA1, Sotaro YAMASHITA2,

Naomichi EZUMI2, Yousuke NAKASHIMA2,Masakatsu FUKUMOTO3, Mamoru SHOJI4,Mizuki SAKAMOTO2, Takaaki IIJIMA2,Akihiro TERAKADO2, Kunpei NOJIRI2,

Hiromasa TAKENO1

1. Graduate School of Engineering, Kobe University, Kobe 657-8501, Japan2. Plasma Research Center, University of Tsukuba, Tsukuba 305-8577, Japan3. National Institutes for Quantum and Radiological Science and Technology,

Naka 311-0193, Japan4. National Institute for Fusion Science, Toki 509-5292, Japan

To monitor and to control the neutral gas pressure around the divertor region isone of the key issues in radiative divertor. The fast ionization gauge speciallydesigned for the divertor environment is a promising diagnostic tool for the mon-itoring of the divertor gas pressure [1-3]. However, in some species of radiatorgases, it is observed that the sensitivity of the ionization gauge is changed dueto the mixing of different species of gases such as hydrogen and helium [4,5]. Itis reported that the mixture of nitrogen and hydrogen can be a powerful radi-ator gas for the divertor operation [6]. Therefore detailed investigation on themixture-gas effects in the diagnostics of mixture gases with nitrogen is required.In this research, the sensitivity of the fast ionization gauge in the mixture ofnitrogen and hydrogen is investigated both experimentally and numerically. Thesensitivity of the ionization gauge against the mixed gas is investigated by usingthe calibration test chamber. The detailed result of the investigation and theanalysis on the effect of mixing the nitrogen and hydrogen will be shown in thepresentation.

References:[1] G. Haas et al., J. Nucl. Mater. 121, 151 (1984).[2] G. Haas and H.-S. Bosch, Vacuum 51, 39 (1998).[3] K. Ichimura et al., Rev. Sci. Instrum. 87, 11D424 (2016).[4] K. Ichimura et al., Plasma Fusion Research. 13, 3405029 (2018).[5] K. Ichimura et al., Plasma Fusion Research. 14, 3405124 (2019).[6] N. Ezumi et al., Nucl. Fusion 59 066030 (2019).

P2-39

Spectroscopic Studies of Magnetized Hydrogen

Plasma via Duoplasmatron Ion Source on Near

Metal Surfaces

Jhoelle Roche M. GUHIT1, Kenta DOI1, Motoi WADA1

1. Graduate School of Science and Engineering, Doshisha University, Kyotan-abe, Kyoto 610-0394, Japan

Spectroscopic studies of the hydrogen plasma-metallic material interaction areinvestigated using a monochromator with 0.23 A spectral resolution. Suspendedin a linear magnetic field, the hydrogen plasma strikes the surface of the metaltarget at a 45-degree angle of incidence. Currently, this plasma irradiation de-vice (LxWxH: 440mmx225mmx225mm) confines a plasma with 1 A and 70 Vdischarge current and voltage, respectively. The Doppler shift of target atomsreflected at the surface immersed in plasma is studied near the metal target.

The device is being modified to introduce positive ions to probe the surfaceunder plasma irradiation. Duoplasmatron ion source will provide a flux of ionsto the surface of the target metal. As the ions travel through the plasma volume,the space-charge blow-up of the ion beam is mitigated to make transport in theenergy range from 100 eV to 1 kV possible. Both reflection component, as wellas sputtering atoms, will be measured with the high-resolution spectrometer.The effects due to plasma bombardment on the particle emission reflection arediscussed.

P2-40

Atomic and Molecular Processes in Plasma

Decomposition Method of Hydrocarbon gas

Makoto OYA1, Ryosuke IKEDA2,Kazunari KATAYAMA1

1. Faculty of Engineering Science, Kyushu University, Fukuoka 816-8580,Japan

2. Interdisciplinary Graduate School of Engineering Sciences, Kyushu Uni-versity, Fukuoka 816-8580, Japan

For efficient fuel cycle in a fusion reactor, hydrogen isotope recovery from vac-uum exhaust gas is important technique. The exhaust gas from vacuum vesselcan contain hydrocarbon even in all-metal wall devices, i.e. QUEST in KyushuUniversity [1], because carbon (C) atoms dissolve in the metals (tungsten andsteel). Therefore, hydrogen isotopes have to be recovered by decomposition ofthe hydrocarbon gas. In ITER, the hydrocarbon is planned to be decomposedchemically by using catalyst, but the efficiency is difficult to improve because ofthe catalyst degradation and so on.

We have proposed a new method of hydrocarbon decomposition using plasma[2]. A part of hydrogen (H) atoms were recovered by this method. However,byproducts of C atoms were deposited on the inner wall of a vessel, resulting instrong decrease in the H recovery efficiency. In order to improve the efficiency,hydrocarbon decomposition and C deposition processes must be discussed froma viewpoint of atomic and molecular processes. In this study, methane (CH4)decomposition process in plasma, incl. transport, elastic collision, ionization anddissociation, were investigated. Reflection, deposition and sputtering from inner-wall were considered as plasma-wall interaction processes.

We conducted the CH4 decomposition experiments by using radio-frequency(RF) plasma. The experimental results showed that approximately 50% of in-jected CH4 gas was decomposed by the plasma. In addition, in order to revealatomic and molecular processes in the experiments, we developed a Monte Carlosimulation code on CH4 decomposition process. The simulation results showedelectron-collisional dissociations of CHx (x = 1 - 4), i.e. CH4 + e− → CH3 + H +e−, were the most dominant processes on the decomposition. The most efficientplasma parameter for this decomposition method is also presented. Furthermore,C deposition distribution is discussed both experimentally and by simulation.

[1] Z. Wang et al., Review of Scientific Instruments 88 (2017) 093502.[2] K. Katayama et al., Fusion Engineering and Design 82 (2010) 1381-1385.

P2-41

Laser desorption measurement of cesium

adsorbed on a molybdenum plasma grid of a

negative hydrogen ion source

Aliena Mari MIRANDA1, James Edward HERNANDEZ II1,Allen Vincent CATAPANG1, Glynnis Mae SAQUILAYAN2,

Mieko KASHIWAGI2, Motoi WADA1

1. Graduate School of Science and Engineering, Doshisha University, Kyoto,610-0394, Japan

2. National Institutes for Quantum and Radiological Science and Technology(QST), Naka, 311-0193, Japan

In the development of high power negative hydrogen (H-) ion beams for neutralbeam injector systems of fusion reactors, the negative ion sources are operatedwith cesium (Cs) to enhance the surface production of negative ions. The ad-sorption of Cs on the surface lowers the work function and allows the conversionsfrom positive hydrogen ions and atomic hydrogens to negative ions by electrontransfer from the PG surface. As Cs vapor is injected into the source chamber,a thin layer of Cs is deposited on the molybdenum plasma grid (PG) electrodeand the formation of a homogeneous Cs monolayer to efficiently generate nega-tive ions is essential for large H- sources. Diagnostic methods to measure the Csdensity in the source volume are already being utilized, such as surface ionizationdetectors and absorption spectroscopy [1-2]. However, to measure the actual ad-sorbed Cs atoms on the surface, a new method is being developed through thelaser desorption process.

This research proposes a new method to measure Cs spatial distribution in theion source by using laser induced desorption of Cs adsorbed on a Mo surface. Byusing a pulsed laser to gently release the cesium without damaging the surface,a better understanding of Cs adsorption and migration on the surface of theplasma grid can be obtained. A 1064 nm Nd:YAG pulsed laser with a maximumenergy of 400 mJ will be used as the desorption laser. The power density rangeto be scanned will be determined such that the maximum power density preventsdamage of the Mo target due to laser irradiation. The Mo target will be coatedin Cs vapor using a Cs oven. To detect the desorbed Cs, time-of-flight (TOF)measurements of Cs+ ions will be attempted by coupling an intense laser pulse toa Faraday cup ion detector. The correlation between the laser power density andthe Cs flux measurements will be used to analyze the adsorbed Cs properties.

References[1] U. Fantz and C. Wimmer, AIP Conf. Proc., 1390, 348358 (2011).[2] U. Fantz and C. Wimmer, J. Phys. D. Appl. Phys., 44(33) (2011).

P2-42

Dynamics Observation of an Atmospheric

Pressure and Low Temperature Helium Plasma

Jet by Laser Spectroscopy on the 23S Atoms

Masahiro HASUO1, Jiro KAKUTANI1, Shoya TAKAGI1,Keisuke UENO1, Keisuke FUJII1, Taiichi SHIKAMA1

1. Graduate School of Engineering, Kyoto University

Atmospheric pressure and low temperature plasma jets have been attracting con-siderable attention because of high potential for chemical reaction without ther-mal damage on the plasma-irradiated surface. Usually, the spatial distributionof the plasma jet varies in a scale smaller than 1 mm. When we use dielectricbarrier discharge, the plasma jet is generated in a time scale of 1 µs. Here, weobserved spatiotemporal change of such a helium plasma jet with measurementof the 23S-23PJ absorption spectra.

An atmospheric pressure and low temperature helium plasma jet, which spoutedout to open air, was generated at a repetition rate of about 37 kHz with a glasstube (inner diameter of 1.8 mm) and a cold cathode fluorescent lamp invertor.The light source of the absorption spectroscopy was a DFB diode laser. Wescanned the wavelength of the laser light over a period of 100 s and measured thelight intensity transmitted through the plasma jet. We performed this measure-ment with 0.1 mm and 2 mm steps in the radial and axial positions, respectively,in the plasma jet. We read out the measurement data according to the dischargecurrent peak to obtain the absorption spectrum at each time and positon. Fromthe area of the spectrum, we estimated the line integrated 23S atom density, andthen estimated the local density with the inverse Abel transformation. From theDoppler and pressure widths, we estimated the gas temperature.

We estimated the electron plume velocity to be 15 km/s from the temporalchange of the 23S atom density distribution. Since we could not detect clear de-pendence of the spectral shape on the plasma radial position, we estimated thegas temperature assuming a constant temperature along the laser beam. Fromthe observed spatiotemporal change of the temperature, we found that the tem-perature rapidly increases at the time of the laser light absorption peak and thengradually decreases. For the purpose of understanding the observed dynamics, wewill consider 23S atom excitation by the electron plume in the plasma jet and thedecay by collisions with surrounding atoms and molecules. We will also discussthe possibility to observe the spatiotemporal dynamics of the plasma jet usingtwo-photon excitation of helium 23S atoms.

P2-43

Experimental investigation of electron

acceleration process during high guide field

magnetic reconnection in UTST

Takumi MIHARA1, Michiaki Inomoto1, Kyohei Kondo2,Hiraku Kaneko1, Kazu Kusano1, Hiroshi Tanabe2,

Yasushi Ono2

1. Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa277-8561, Japan

2. Graduate School of Engineering, The University of Tokyo, Tokyo 113-8656,Japan

In the UTST experiment, plasma merging method is adopted to form a spher-ical tokamak plasma without use of a center solenoid coil. During merging start-up, magnetic reconnection phenomenon occurs under high guide (toroidal) mag-netic field. Since the reconnection electric field and the magnetic field are parallelnear the X-point of reconnection with a guide field, electron acceleration by theparallel component of electric field is observed in numerical [1] and experimen-tal studies [2]. It is important to elucidate the electron acceleration mechanismbecause the accelerated electrons may contribute to the initial electron heatingduring merging formation of ST. An important feature of the electron accelera-tion process is that the poloidal electric field generated by the charge separationreduces the parallel component of the reconnection electric field, resulting in sup-pression of electron acceleration. Thus, transient evolution of the self-generatedelectric field should be clarified in detail. In the UTST experiment, the poloidalelectric field was derived from the difference between floating potentials measuredat two locations. During merging formation of ST, large axial electric field up to4.7 kV/m was observed. This self-generated electric field is expected to cancel80% of the parallel component of the reconnection electric field. Since the peaktimings of the self-generated electric field and the soft X-ray (SXR) intensityare almost same, it is indicated that the self-generated electric field would affectthe electron acceleration process. Further investigation using two-dimensionalmeasurement of SXR emission in under construction and its initial results willbe presented at the conference together with the observation of poloidal electricfield.

Reference

[1] P.L. Pritchett and F.V. Coroniti, J. Geophys. Res. 109, A01220 (2004).[2] T. Ushiki, et al., Plasma and Fusion Res. 11, 2402100 (2016).

P2-44

Evaluation of spatial characteristics of divertor

simulation plasma during impurity seedings in

GAMMA 10/PDX

T. Hara1, T. Iijima1, N. Ezumi1, K. Nojiri1,A. Terakado1, M.S. Islam2, M. Hirata1,

M. Sakamoto1, Y. Nakashima

1. Plasma Research Center, University of Tsukuba,Tsukuba, Ibaraki 305− 8577, Japan

2. National Institute for Fusion Science, Toki, Gifu 509− 5292, Japan

It is concerned that the divertor is exposed from large particle and head load.For divertor plate protection, one of the method to decrease the large particle andheat load is formation of detached divertor using cooling gases such as impurityand/or fuel gases. Therefore it is important to understand the spatial structureof detached plasma and the influence of the upstream plasma conditions on thedetached plasma.

In this study, we have observed changes of spatial characteristics by injec-tion of radiator gas (H2, Ar, N2, etc) into the divertor simulation experimentalmodule (D-module) installed at the end region of GAMMA 10/PDX [1-2]. Anew measuring instrument with electrostatic probe and spectrometer was madeand installed at the inlet of D-module(upstream). In the present experiment,the ion saturation current, the electron temperature and density are measuredby Langmuir probes and the emission spectra are observed by spectrometers atthe inlet of D-module (upstream) and corner region of V-shaped target plate inthe D-module (downstream). These two points are located on the same magneticfield line. The results from these observations indicate that detached plasma wasformed at the downstream while ionization was dominant at the upstream. Inaddition, the degree of cooling effect and the ratio of density in the formation ofdetached plasma process was different depending on gas species. It is consideredthat the difference is due to impurity transport and radiation loss process.

In this presentation, the details of the energy loss process and comparisonwith upstream and downstream will be discussed.

[1] Y. Nakashima, et al., Nucl. Fusion 57 (2017) 116033.[2] M. Sakamoto, et al., Nucl. Mater. Energy, 12 (2017) 1004.

P2-45

Flow Structure Formed by Turbulence in PANTA

Naruyuki UESHIMA1, Shigeru INAGAKI2,

Kenichiro TERASAKA1, Kotaro YAMASAKI2,Hiroyuki ARAKAWA3, Chanho MOON2,

Yuichi KAWACHI2, Yoshihiko NAGASHIMA2,Akihide FUJISAWA2

1. Interdisciplinary Graduate School of Engineering Science, Kyushu Univer-sity, 6-1 Kasuga-koen, Kasuga 816-8580, JAPAN

2. Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasuga-koen, Kasuga, 816-8580, JAPAN

3. Shimane University, 1060 Nishikawatsu, Matsue, 690-8504, JAPAN

It is well known that turbulence in magnetized plasma interacts with plasma flow,which is driven by turbulence. For understanding such a self-organized structureformation mechanism of the plasma turbulence, it is required to observe theflow structure. Then we perform experiment to excite the plasma turbulence ina linear plasma turbulence device, PANTA (plasma length is 4 m and plasmaradius is 0.05 m). In the PANTA, three turbulent-states is observed; 1) driftwaves excited by density gradient are dominant [1], 2) the D ’Angelo mode isexcited through the Kelvin-Helmholtz type parallel flow share instability [2], 3)other modes are dominate. The flow structure formed by turbulence is predictedto be different depending on the turbulence-state [3]. In this study, we observedparallel and azimuthal flows of the PANTA argon plasma by the laser inducedflorescence. The flow was found to have a spiral structure along the magnetic axis.Differences of direction and pitch of the spiral depending on the turbulence-statewill be reported and they will be compared with theoretical predictions.

[1] T. Yamada, et al., Nature Physics 4, 721725 (2008)[2] N. Dupertuis, et al., Plasma and Fusion Research 12, 1201008 (2017)[3] S. Sasaki, et. al., Phys. Plasmas 24 112103 (2017).

P2-46

Ion Mass Effect on Beam Deflection

Compensation by Aperture Displacement and

Application of Ferromagnetic Material as an

Alternative Compensation Technique

Masashi KISAKI1,2, Katsuyoshi TSUMORI1,2,Haruhisa NAKANO1,2, Katsunori IKEDA1,Yutaka FUJIWARA1, Kenichi NAGAOKA1,3,

Masaki OSAKABE1,2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies)3. Nagoya University

Since first deuterium plasma campaign on the LHD, the beam broadening hasbeen observed in negative-ion-based NBIs (NNBIs). One of the possible causes is amismatching of the beam deflection compensation by the aperture displacementapplied on a steering grid (SG). The apertures on the SG are displaced withrespect to the central axes of the apertures on a plasma grid in order to modifyequipotential lines and steer the negative ion beam perturbed by the magneticfield inside the accelerator. Since the beam deflection angle by the magnetic fielddepends on the ion mass, the deviation of the beamlet axis from the aperturecentre at the SG exit is varied from the designed value, which is optimized forthe hydrogen beam, in the case of the deuterium operation.

In this study, the beamlet trajectory was calculated with 3D beam trajectorycode (OPERA-3d) for both hydrogen and deuterium beams taking into accountthe magnetic field and aperture displacement, and comparison between the sim-ulation and the experiment was performed. It was confirmed that the deflectionof the deuterium beam is over-compensated due to its larger Larmor radius andthis results in broadening of the deuterium beam. An alternative compensationtechnique, where the SG was replaced by the ferromagnetic one, was also studiedin order to realize the cancellation of the beam deflection without dependence onthe ion mass. In the presentation, we discuss the simulation and experiment re-sults in above two compensation techniques performed in LHD-NNBI and testbedin detail.

P2-47

Confinement of D-3He fusion product particles in

a non-adiabatic trap

Yamato TSUNOKAKE1, Takahiro URANO1,Kojiro SEKIGUCHI1, Ryo MATSUMOTO1,Toshiki TAKAHASHI1, Momota HIROMU2

1. Graduate School of Science and Technology, Gunma University2. National Institute for Fusion Science

A non-adiabatic trap is an open confinement concept in which a wide externalweak magnetic field region is formed in the center of the device. Then fusionproduct particles into this region, adiabaticity in the particle motion typified bythe preservation of the magnetic moment breaks in the center weak magneticfield region and the particle motion becomes stochastic. This mean a fusionreactor in which multiple non adiabatic traps are connected can expect extensionthe confinement time. Moreover, since it is a high beta plasma, synchrotronradiation loss is small, and it can be expected as a coreplasma of advanced nuclearfusion like D-3He. In addition, because magnetic field is open-ended, the directenergy converter enables high efficiency power generation. Therefore, in thisstudy, we investigated whether D-3He nuclear fusion particles can be confined bynon-adiabatic trap. By calculating the particle trajectories of 14.7-MeV protons,we consider about the particle effect to the fusion reactor.

P2-48

Quasi-equilibrium calculation with external

magnetic flux decay in field-reversed

configuration plasma

Yuto SUGIKI1, Takahiro URANO1,Toshiki TAKAHASHI1

1. Graduate School of Science and Technology, Gunma University

Although the lifetime of field-reversed configuration (FRC) plasmas formed bya field-reversed theta pinch method is about a several hundred micro-seconds, aquasi-steady state condition can be obtained as a high-beta equilibrium state. Ex-perimentally, the radius of the separatrix that is the boundary separating closedfrom open field regions has been found almost constant[1]. On the other hand,a time evolution of the separatrix radius estimated by a simple resistive decaymodel shows its gradual decrease, which is inconsistent with the experimentalresults. Generally, the coil current for external field is assumed to be constant.However, the temporal change in the coil current is in an attenuated sinusoidalwaveform due to properties of the external field circuit. Effects of the coil currentattenuation have never been estimated.

In the present research, considering the external magnetic flux decay, we as-sumed a quasi-equilibrium state in each calculation timestep. The simulation isdone for the FRC experimental device Nihon University Compact Torus Experi-ment (NUCTE)-III. Calculation results will be compared with experimental dataof NUCTE-III[2]. When the external flux decay is considered, we are able toobtain a quasi-equilibrium condition with an almost constant separatrix radius.

[1] Loren C. Steinhauer, Phys. Plasmas 18, 070501 (2011).[2] Hirotomo Itagaki: A Doctoral Dissertation,“Studies of stabilization tech-

nique for field-reversed configuration by using magnetized plasmoid injection”,The University of Tokyo, December 2013.

P2-49

Hollow cathode discharge experiment applying

magnetic field of permanent magnets

Kojiro SEKIGUCHI1, Ryo MATSUMOTO1,

Takahiro URANO1, Toshiki TAKAHASHI1,Makoto GOTO2, Tomio OKADA3

1. Graduate School of Science and Technology, Gunma University2. Polytechnic Center Gunma3. NPO Wireless Brain Network

A high density plasma source is needed to realize a high current density ion beamfor advanced cancer treatment, such as boron neutron capture therapy (BNCT).By using a hollow cathode discharge, we can obtain relatively high density plasmadue to the reciprocating electron motion between facing cathode electrodes, so-called hollow cathode effect.The purpose of this study is to clarify whether it is possible to further increasethe discharge current by applying a static magnetic field generated by a pairof permanent magnets to a plane-parallel hollow cathode discharge. Permanentmagnets were arranged so as to sandwich the central parallel plate electrodes.The case where the north pole and the south pole of the magnet are arranged toface each other is called N-S here. Also, the arrangement in which the north polesface each other is called N-N. In the case of N-S, the directions of the electricfield and the magnetic field become parallel, and the effect of suppressing thedissipation of electrons can be expected. On the other hand, in the case of N-N,the magnetic field is almost zero at the central region. At the end of the electrode,the magnetic field is directed outward from the electrode. It is expected that thismagnetic field structure inhibits the reciprocating motion of electrons and cannotelicit the hollow cathode effect.The hollow cathode electrode is 30 x 70 mm wide by 10 mm thick and madeof stainless steel. The anode is made of copper and a cylindrical neodymiummagnet with a diameter of 20 mm and a thickness of 20 mm is attached. Theelectrodes are placed in a Pyrex vacuum vessel. In N-S, the magnetic flux densityat the center of the discharge space is 16.3 mT and increases quadratically in theelectrode direction. Argon is used as the filling gas, and the gas pressure is 0.05to 0.1 Torr.In our poster, we will report the analysis results of electron density and electrontemperature of hollow cathode discharge by probe measurement. The results arecompared for the three magnetic field structures of no magnetic field, N-S, andN-N, and we will clarify the magnetic field structure that can obtain the highhollow cathode effect.

P2-50

Particle simulation for high-density hollow

cathode discharge

Ryo MATSUMOTO1, Kojiro SEKIGUCHI1,

Takahiro URANO1, Toshiki TAKAHASHI1,Makoto GOTO2, Tomio OKADA3

1. Graduate School of Science and Technology, GUNMA UNIVERSITY2. Polytechnic Center Gunma3. NPO Wireless Brain Network

There is a cancer treatment called boron neutron capture therapy. This treat-ment requires a high current density ion beam. To achieve this, a high-densityplasma source is required. High-density plasma can be obtained by the hollowcathode effect using hollow cathode discharge.The purpose of this research is to reproduce the high-density plasma generated bythe hollow cathode discharge by simulation and evaluate the consistency with thehollow cathode discharge experiment conducted. The calculation model repro-duced the size of the electrode used in the experiment. To reduce the calculationload, it is necessary to limit the number of particles and change the weight ofthe particles to conservation momentum. We employed the method called“ thelimited weight probability method”to limit the number of particles[1, 2]. In thismethod, the number of particles is controlled so that charge conservation in theslab can be achieved using the space structure of the fine-subslab model. In thefuture, the challenge will be to incorporate this method into the calculation.

[1] M. Goto, Y. Kondoh, and A. Matsuoka: Jpn. J. Appl. Phys. 36 (1997) 4815.[2] M. Goto and Y. Kondoh: Jpn. J. Appl. Phys. 37 (1998) 308.

P2-51

Development of Cs-free negative ion source by

sheet plasma

Keito Hana1, Toshikio Takimoto1, Hiroki Kaminaga1,Akira Tonegawa1, Kohnosuke Sato2,

Kazutaka Kawamura1

1. Tokai University, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa, Japan2. Tokyo University of Science, 1-3 Kagurazaka, Shinjuku-ku, Tokyo 162-

8601, Japan

Negative ion source plays an essential role for NBI system of steady state mag-netic nuclear fusion. These negative ion sources produce mainly negative ions bysurface conversion of atoms impinging on a caesium covered surface to achieve ahigh ion density. However, caesium degrades the time stability and uniformity ofthe beam. Therefore, the development of a negative ion source without caesiumseeding is strongly desired, and we have development of negative ion source ina caesium-free discharge by the magnetized sheet plasma device, TPDsheet-U[1]. Negative ions are formed by volume-production, that is, the dissociative at-tachment of low energy electrons (Te = 1-2 eV) to highly vibrationally excitedmolecules. These molecules are attributed to the electron-impact excitation ofmolecules by high energy electrons (Te > 10 eV) in the plasma column. Negativehydrogen/deuterium ion beams are extracted through the small single-aperturegrid (diameter; 4 mm) at a neutral gas pressure Pe of 0.3 Pa. At an extractionvoltage of 7 kV, the maximum negative current densities Jc is about 11.7 mA/cm2and 6.03 mA/cm2 with hydrogen and deuterium discharge, respectively [2]. Forincrease the negative ion current, a preliminary experiments of the negative ionbeam extraction at TPDsheet-U have been performed using the extraction elec-trodes of multi-aperture grid (124 mm length, 24 mm width, 3 × 13 holes with4mm diameter). We have succeeded in the extraction of the negative ion beamfrom the Cs-free negative ion source by sheet plasma.

This study have been supported by the JSPS KAKENHI Grant NumberJP19K03795 and Joint Project of the National Institute for Fusion Science(NIFS19KBAR025). .

[1] A.Tonegawa et al., Japanese Jornal of Applied Physics, 45, 8212, (2006).[2] Keito Hanai, et al., Fusion Engineering and Design,https://doi.org/10.1016/j.fusengdes.2019.04.096

P2-52

Numerical analysis of quantum mechanical

E × B drift in non-uniform electric fields

Hiroshi SHIMIZU1, Shun-ichi OIKAWA2

1. Graduate School of Engineering, Hokkaido University, Sapporo 060-8628,Japan

2. Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Japan

In fusion plasma, there are many charged particles under the presence of an elec-tromagnetic field. Therefore, considering about a single charged particle move-ment will be a clue for understanding physics of entire fusion plasma. The be-havior of particles in fusion plasma is different from that the classical theorypredicts(e.g. diffusion of plasma). Hence, there needs to consider about somequantum effects on such phenomena.The guiding center drift of a charged particle has attracted the interests of manyresearchers and been studied in classical and quantum physics for a long time.Especially, Spitzer found the expression for the drift velocity and the diffusioncoefficient of a charged particle under an non-uniform electromagnetic field. How-ever, quantum mechanical approaches are not enough for this theme.This research aimed to obtain quantum effects on a charged particle in a weak elec-tric and an uniform magnetic field. In this study, time-dependent two-dimensionalSchrodinger equation is solved with numerical methods to calculate the movementof a charged particle. The equation is difficult to solve analytically in this situ-ation. With higher order non-uniform electric field, low energy particles woulddrift differently from the prediction of the classical drift theory. When the electricfield order is higher than linear, time evolution of the variance of position andmomenta increases.

References

[1] P. K. Chan, S. Oikawa, and W.Kosaka,“ Quantum mechanical grad-B driftvelocity operator in a weakly non-uniform magnetic field,”Phys. Plasmas 23,022104 (2016).

[2] P. K. Chan, S. Oikawa, and W. Kosaka,“Quantum mechanical expansion ofvariance of a particle in a weakly non-uniform electric and magnetic field,”Phys. Plasmas 23, 082114 (2016).

[3] P. K. Chan, S. Oikawa, and W. Kosaka,“ Quantum mechanical E × B driftvelocity in a weakly inhomogeneous electromagnetic field,”Physics of Plasmas24, 072117 (2017)

P2-53

GLOBAL CALCULATION OF IMPURITY

TRANSPORT INCLUDING THE VARIATION

OF ELECTROSTATIC POTENTIAL ON THE

FLUX SURFACE IN HELICAL PLASMAS

Keiji FUIJTA1, Shinsuke SATAKE1,2,

Masanori NUNAMI1,2, Jose Manuel GARCıA-REGANA3

1. Sokendai2. National Institute for Fusion Science3. Laboratorio Nacional de Fusion CIEMAT

In order to explain the mechanism behind the impurity hole phenomenon ob-served in LHD[1], several efforts have been made. Inclusion of the variation ofelectrostatic potential over the flux surface, Φ1, in neoclassical transport theoryis one of the attempts[2-5].

Impurity hole plasmas are characterized by low density, low collisionality andthe negative ambipolar radial electric field with relatively small amplitude. Ithas been shown that the effect of the magnetic drift, vm, is not negligible indetermining the background ion profile and the discrepancy in the Φ1 profilebetween models with and without Φ1 becomes significant in such plasmas[5,6].In particular, vm plays a similar role as E×B drift due to positive radial electricfield in orbit equations, and the profile of Φ1 becomes close to those obtained bylocal calculation with positive radial electric field.

However, all of the previous studies on the effect of Φ1 on impurity transporthave been conducted with local simulation codes and no global calculation hasbeen performed yet. Thus, we update to enable a global neoclassical simulationcode FORTEC-3D to perform multi-species calculation including the Φ1-effect.

In the presentation, the result of the first calculation of global neoclassicalimpurity fluxes in LHD including the effect of Φ1 will be presented. The resultwill be compared with previous local results and the condition under which theconsideration of Φ1 is necessary will be discussed.

References

[1] K. Ida et al. Phys. Plasmas 16 056111 (2009).

[2] J. M. Garca-Regaa et al. Plasma Phys. Control. Fusion 55 074008 (2013).

[3] J. M. Garca-Regaa et al. Nucl. Fusion 57 056004 (2017).

P2-54

[4] A. Molln et al. Plasma Phys. Control. Fusion 60 084001 (2018).

[5] J. L. Velasco et al. Plasma Phys. Control. Fusion 60 07400 (2018).

[6] K. Fujita et al. Plasma and Fusion Research 14 3403102 (2019).

P2-54

Strange Shapes of Ion Velocity Distribution

during Magnetic Reconnection in the Presence of

a Guide Field

Shunsuke USAMI1,2, Ritoku HORIUCHI1,3,Hiroaki Ohtani1,3

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. The University of Tokyo3. Sokendai (The Graduate University of Advanced Studies)

By means of two-dimensional particle simulations, we investigate ion dynamics inthe downstream region of magnetic reconnection in the presence of a guide field.The ion dynamics are deeply related to ion effective heating.

Plasma merging experiments in spherical tokamaks (STs) have shown thations are heated mainly in the downstream of reconnection, but the mechanismremained unsolved. In Ref. [1], our particle simulations have demonstrated thatring-shaped proton velocity distributions are formed, and protons are effectivelyheated during magnetic reconnection with a guide field. The proton behaviorsare the Pick-Up-Like applied to magnetic reconnection [2]. We construct thefollowing basic theory of the ion motion. When ions enter the downstream acrossthe separatrix, they behave as nonadiabatic and are energized in the downstream.

Recently, our simulations further show strange shapes of velocity distribution,such as not only a simple ring-shape, but also a horn-shape [3], a circular-arc-shape, and further a multi-structure-shape combing them. For explaining theformation process of them, we have improved theory. In the basic theory [1],we have assumed that the electromagnetic field is uniform, and have ignored ionacceleration in the separatrix. In the extended theory [3], we have taken accountof the ion acceleration in the separatrix. In this work, we construct “enhanced-extended theory,” dealing with non-uniformity of the electromagnetic field in thedownstream. We succeed in accounting for the anomalous structures seen invelocity space.

[1] S. Usami et al., Phys. Plasmas 24, 092101 (2017).[2] J. F. Drake, et al., Astrophys. J. 700, L16 (2009).[3] S. Usami et al., Plasma Fusion Res. in press.

P2-55

Observation of Heat Flux Time Evolution

Carried by Reentering Fast Ions

Hiroto MATSUURA1, HIdeki Muraoka1,Bui Xuan Nhat Son1, Kenichi Nagaoka2

1. Osaka Prefecture University2. National Insititute for Fusion Science

For the design of fusion reactors, confinement property of fast ions is one ofthe most important issues, since neutral beam (NB) heating, and alpha particleheating, are dominant heating process in deuterium-tritium burning plasmas.Such a ion orbit shift from the magnetic surface and depends not only on magneticfield configuration but also on the geometrical structure of plasma facing surface.In a helical system such as large helical device (LHD), there is some fast ionswhose orbits go out across the last closed flux surface (LCFS) and then re-enterinside the LCFS, and they are called as re-entering particle [1]. such a particlemight have effect on magnetohydrodynamics equilibrium and induced instabilityenhanced fast ion loss.

In LHD, there are three tangential NB injectors (co:BL2 and ctr:BL1 and 3)and fast ion orbits circulating in the co- and ctr-direction slightly shift outwardand inward, respectively. So even if total hetaing power is kept constant, fast iondistribution just outside of LCFS might drastically change. In order to measuresuch fast ions and so-called divertor leg plasma, a hybrid directional langmuirprobe (HDLP) was installed in LHD [1]. On the contrary to fast ion pressure,however, energy carried by fast ions need time dependent heat balance analysis.

We proposed a new heat conduction model in which time-depend heat flux istreated as sum of many step-like pulses and each step size and sign are determinedso as to measured probe tip temperature [2]. By setting HDLP on the LHDdivertor leg, this model has been used to monitor heat flux evolution duringdetach plasma formation. In this work, this model is applied to HDLP data,which is just outside of LCFS, where only co-direction facing probe tips showstemperature increment during a shot. By careful comparison of NB injectiontiming, heat flux received by HDLP is found to start decreasing after co-injection(BL2) terminated, although plasma is still sustained by other beam line. Sincetemperature response to heat flux change is not so fast, decay time of estimatedfast ion heat flux does not directly relates with confinement time of fast ion. Evenso it is possible to compare the fast ion behavior in D plasma and H plasma. Thiswork is partially performed with the support and under the auspices of the NIFSCollaborative Research Program. ( NIFS18KLPR044/ NIFS18KUGM134 )

[1] K.Nagaoka et al., Rev. Sci. Instrum. 79, (2008)10E523.[2] H.Matsuura et al., Contrib. Plasma Phys. 54, (2014) 285 290.

P2-56

Particle Simulation of Divertor Plasma with

Electrical Biasing

Trang LE1, Yasuhiro SUZUKI1,2, Hiroki HASEGAWA1,2,

Seiji ISHIGURO1,2

1. SOKENDAI (The Graduate University for Advanced Studies), Toki, Gifu509-5292, Japan

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

One of the critical problems in the fusion plasma experiments is the high heatflux to the divertor plates, which damages the material of the wall. In order toavoid this high heat loads, power and particle control must be studies carefullywith Plasma-wall interaction. There are various approaches have been proposedfor solving this issue. One of them is adding electrical biasing to the divertortarget plates. Electrical biasing to the divertor has been implemented on severaltokamaks such as DIII-D. It has been demonstrated that divertor biasing hasthe promising applications to tokamak reactor. It helps to increases the plasmatransport and decreases the heat loads to the divertor by inducing an E×B

convection flow. Divertor biasing also has strong effect on increasing Scrape-off-layer (SOL) width. However, there is a few number of theoretical and numericalstudies on this technique. Two main models used in particle simulation arefluid model and kinetic model. The fluid model cannot clearly explain the heatfluxes due to the effect of high energetic ions and electrons which move alongthe magnetic field to reach the divertor. Kinetic model is suitable to deal withhigh energetic particles. Particle-In-Cell (PIC) simulation is method to modelthe magnetic field and electrical potential structure self-consistently using fullykinetic description. PIC has been considered as one of the powerful method forplasma study especially for the study of plasma edge region. In this work, wedevelop 2 spatial dimensions, 3 velocities coordinates (2D3V) PIC code to studyof the behavior of potential and energy transport in the SOL region at the divertorwhen the electrical biasing to the divertor plates is applied.

P2-57

Comparison of integral-form and differential-form

of dielectric tensor in kinetic full-wave analysis of

cyclotron waves in tokamak plasmas

Atsushi FUKUYAMA1

1. Kyoto University

In order to describe the wave structure and power deposition profile in ion andelectron cyclotron wave heating in tokamak plasmas, a considerable progress hasbeen made in developing full wave codes including kinetic effects. Most of pre-vious kinetic analyses of wave propagation and absorption in an inhomogeneousplasma are based the wave number, since the dielectric tensor in a uniform hotplasma are usually expressed as a function of wave number. Using this type of di-electric tensor, various approximate methods have been proposed and applied tothe analysis of cyclotron heating. They require, however, some constraints suchas weak inhomogeneity, long wave length, and large computational resources. Inorder to describe the response of inhomogeneous plasma without wave number,an approach using an integral form of the dielectric tensor has been proposed.Maxwell’s equation with the integral form of dielectric tensor can be numericallysolved as a boundary-value problem by the finite element method (FEM). Nu-merical analysis with FEM may have higher performance with parallel processingowing to sparse coefficient matrix. Though the integration is usually localized inan element in FEM for differential equations, coupling between different elementsoccurs in FEM for integro-differential equations. One-dimensional analyses havebeen successfully applied to the energetic particle effects in ion cyclotron heat-ing and the mode-conversion to the electron Bernstein wave in electron cyclotronheating. In this paper, two-dimensional mode structure of the cyclotron waves onthe poloidal plane of tokamak is shown for the integral-form and the differential-form of the dielectric tensor. Comparison of the wave structure and the numericalperformance will be presented.

P2-58

Estimating ripple transport of fast tritons by

D-D fusion in JT-60SA tokamak

Anggi Budi KURNIAWAN1, Hiroaki TSUTSUI1,

Keiji TANI1, Kouji SHINOHARA2, Shunji Tsuji-IIO1

1. Tokyo Institute of Technology2. National Institutes for Quantum and Radiological Science and Technology

The transport of fast triton is analyzed under the presence of magnetic ripplein the JT-60SA operation scenario #3 which is expected to employ DD plasma.Numerical study was done by using guiding-center (GC) and full-orbit (FO) equa-tions. The tritons with wide range of practical high energies were launched at theouter mid-plane on a poloidal cross section. Given the same initial position, ki-netic energy, and pitch angle, the FO scheme obtained an extension of precessionmotion compared to that from the GC scheme, resulting in a difference of ba-nana tip positions between them. Consequently, the transport coefficients givenby GC and FO equations were different. Different ripple-resonance energies fromGC and FO equations were found due to this phenomena. The particle energyof 0.3 MeV and 1.17 MeV were considered as resonance energies by GC scheme,while the FO scheme gives 0.25 MeV and 0.99 MeV. Consequently, the differenceof precession island in Poincare mapping was observed. The tritons with ripple-resonance characteristic were found at the epicenter of precession island, whereparticles stagnate, in a phase space.

P2-59

Numerical Investigation of Energetic Particle

Driven Interchange Mode in LHD

Malik IDOUAKASS1, Yasushi TODO1, Hao WANG1,Jialei WANG1,2

1. National Institute for Fusion Science2. Dalian University of Technology

The Energetic particle driven Interchange mode (EIC) is a mode that was ob-served recently [1] in the Large Helical Device (LHD) while a strong perpendicularNeutral Beam Injection (NBI) was active. The EIC is a magnetohydrodynamic(MHD) instability with m/n=1/1 (where m and n are the poloidal and toroidalmode numbers respectively) that occurs in bursts and causes losses of energeticions from the core plasma. It was found that the frequency was consistent withthe helically trapped energetic ions helical precession frequency. The fast ionlosses are observed through the drop in neutron emission in deuterium experi-ments and are deleterious for the energy confinement and plasma heating. Animportant feature of this mode is that it experiences a frequency down-chirpingduring each burst.

For the study of this mode, the code MEGA [2] is used. It is a fully nonlinearhybrid MHD-kinetic code where the bulk plasma is treated using the fluid MHDdescription and the energetic particles are treated using a drift kinetic descrip-tion. This code uses a realistic geometry and equilibrium, and has been usedsuccessfully on both tokamak and stellarator configurations for energetic particledriven modes.

The mode is investigated using an equilibrium reconstructed from a deuteriumshot where EIC was observed. The role of the energetic ion β, profile and pitchangle is investigated. Preliminary results on destabilization, nonlinear saturation,chirping, and pressure profile redistribution, are presented.

References

[1] X. D. Du et al., Phys. Rev. Lett. 114, 155003 (2015)

[2] Y. Todo, Phys. Plasmas, 13, 082503 (2006)

P2-61

Refinement of Interatomic Potential for Medium

Energy Atomic Collision

Atsushi M. ITO1

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

In the simulation research on the plasma-wall interaction, the injection of theplasma ion onto target material is traditionally solved by the binary collision ap-proximation (BCA). In this simulation, the injected ion particle, which is calleda projectile, repeats the collision with atoms in the target material. At each colli-sion, following the classical mechanics, an asymptotic trajectories of the projectileand the target atom is solved by numerical integration along an interatomic po-tential curve. Therefore, the reliability of the interatomic potential is importantfor the BCA simulation.

The interatomic potential model for BCA was modeled as the Thomas-Fermipotential which is the Coulomb potential with the screening function. Althoughthere are several kinds of the screening function, theoretically the most reliablemodel is the universal model developed by the Ziegler, Biersack, and Littmarkwhich is called the ZBL potential. In their approach, the interatomic potentialenergies were calculated by the Thomas-Fermi-Dirac density functional theory(DFT). As an assumption, the electron density around two atoms is simply sumof the fixed electron density of independent atom. After that, they found afunction being able to well fit the calculated interatomic potential energies for216 pairs randomly chosen from all atomic elements.

Recently, the ZBL potential is often used as the two-body repulsive term ofthe potential model for molecular dynamics (MD). However, the function of theZBL potential is a just fitting function, and the its target used in BCA was highenergy collision when the energy range if more than 1 keV. By this reasons, theZBL potential is not good for the medium energy range from 1 eV to 1 keV.

In this work, the interatomic potential is reconstructed to refine the accuracyfor the medium energy range. Although the same assumption that the interatomicpotential energy is defined by the DFT on the sum of the fixed electron densityis employed, the interatomic potential function is given by analytical solution.

[1] J.F. Ziegler, J.P. Biersack, U. Littmark, ”The Stopping and Range of Ions inMatter”, Vol. 1, edited by J.F. Ziegler, pp 24-49, PERGAMON PRESS, INC.1985.

P2-62

Nonlinear energy transfer between parity reversal

invariant subspaces in incompressible Hall

magnetohydrodynamic turbulence

Keisuke ARAKI1, Hideaki MIURA2

1. Okayama University of Science2. National Institute for Fusion Science

The purpose of this study is to apply the analytical mechanical approach forincompressible Hall magnetohydrodynamics to fully developed turbulences. Thefunction expansion using the generalized Elsasser variables [1], which have theiranalytical basis on the helicity-based particle-relabeling operator [2], naturallydecomposes the function space into the slow-modes space (ion cyclotron waves)and the fast-modes space (whistler waves). In the present study, the nonlinearenergy transfer between these two subspaces is investigated as well as the nonlin-ear energy redistribution in each subspace. It is found that shell-to-shell analysisshows remarkable asymmetry of the energy transfer between these two subspaces.

[1] Galtier, J. Plasma Phys., vol. 72, p. 721-769 (2006)[2] Araki, Phys. Rev. E, vol. 92, 063106 (2015)

P2-63

Parameter dependence of equilibrium with flow

in reduced MHD models

Atsushi ITO1, Noriyoshi NAKAJIMA2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences, 322-6 Oroshi-cho, Toki, Gifu 509-5292 Japan

2. Rokkasho Research Center, National Institute for Fusion Science, NationalInstitutes of Natural Sciences, 2-166 Omotedate, Obuchi, Rokkasho, Kamikita,Aomori 039-3212 Japan

Equilibrium with flow in high-beta toroidal plasmas is investigated based on re-duced MHD models. In the reduced MHD models in Refs. [1,2], as well as toroidaland poloidal flows, two-fluid and ion finite Larmor radius (FLR) effects are takeninto account. These effects are important in the steep profiles of improved con-finement modes of magnetically confined plasmas. The two-fluid and FLR effectsare small scale effects that introduce diamagnetic effects. In the reduced MHD,the poloidal flow is composed of E×B and diamagnetic drifts when these effectsare included. The modifications of magnetic flux surfaces, isosurfaces of pressureand ion stream functions due to the diamagnetic effects can be observed. Thefour models, MHD, two-fluid MHD, MHD with FLR and two-fluid MHD withFLR, are compared with each other. The dependence of equilibria in these mod-els on parameters such as flow velocity and beta value is examined analyticallyand numerically.

Analytic solutions are obtained for simple profiles. The Shafranov shift ofthe magnetic axis, equilibrium beta limit and curvature of magnetic flux sur-faces are obtained from the analytic solutions and their parameter dependencesare examined in detail. Numerical solutions are obtained by using the finite el-ement method. The analytic solutions are also used for benchmark with thenumerical solutions. The parameter dependence for complicated profiles such asnon-constant density and temperature is examined.

[1] A. Ito, J. J. Ramos and N. Nakajima, Plasma Fusion Res. 3, 034 (2008).[2] A. Ito and N. Nakajima, AIP Conf. Proc. 1069, 121 (2008).

P2-64

Evaluation of Pressure Anisotropy derived NBI

by using the Monte Carlo code

Ryosuke SEKI, Kiyomasa WATANABE,

Yasuhiro SUZUKI

1. National Institute for Fusion Science

Three tangential neutral beam (NB) injectors and a perpendicular-NB injectorhave been installed on the Large Helical device (LHD). A volume averaged betareached 5

The pressure anisotropy is evaluated from the experimental evaluation ofplasma pressure by using a magnetic measurement and/or the numerical eval-uation of the beam pressure by using the Monte-Carlo code. A 3-dimensionalbeam-pressure profile evaluated by the Monte-Carlo code is important for theforce balance including the beam-pressure.

In the numerical simulation, it is pointed out that re-entering fast ions havethe large role in the evaluation of beam pressure in the LHD high beta plasma.In analyses including the re-entering fast ions, a fast ion loss due to the chargeexchange reaction with neutral particles is important. Using the MORH code[3]based on orbit following in the real coordinates where the re-entering fast ions[4]are taken account, we evaluate the beam pressure with including the effect ofthe fast ion loss due to the charge exchange reaction. The dependence of thepressure anisotropy on a neutral particle density is investigated. In addition, theforce balance including the beam pressure is discussed. In the conference, thedependence of the beam pressure on a neutral particle density will be shown.

Reference[1] H. Yamada, et al., Nucl. Fusion 51 (2011) 09421.[2] Y. Asahi, et al., Proc. of 38th EPS Conference on Plasma Physics. (2011)

P1.076[3] R. Seki , et al., Plasma and Fusion Res. 5 (2010) 027.[4] R. Seki, et al., Plasma Fusion Res. 3 (2008) 016.

P2-65

Convergence-property Improvement of k-skip CG

and k-skip MrR

Akira MATSUMOTO1, Takayasu MORISHITA1,Kuniyoshi ABE2, Soichiro IKUNO1,

Hiroaki NAKAMURA3,4

1. School of Computer Science, Tokyo University of Technology2. Faculty of Economics and Information, Gifu Shotoku Gakuen University3. Department of Energy Engineering and Science, Nagoya University4. National Institute for Fusion Science

A large and sparse simultaneous linear system often appears in the field of nu-merical simulation of plasma physics. Krylov subspace method mainly consistsof matrix-vector product and inner product; thus, it is easy to parallelize. Onthe other hand, inner products calculated several times in an iteration cause col-lective communications at short intervals. Communication-avoiding technique isinvestigated in order to mitigate degradation of parallelization efficiency causedby collective communications.

In recent years, a k-skip technique has been researched as one of the communication-avoiding technique. The basic concept of the k-skip technique is to avoid innerproduct in k times iterations by pre-calculating Krylov bases. Two communication-avoiding Krylov subspace methods based on k-skip technique have been proposed.The first method is k-skip CG; initially proposed by Motoya et al. k-skip CG hasissues involved in convergence-property deterioration, and some improvement hadbeen made. However, the numerical property still deviates from CG. The secondmethod is k-skip MrR, which has better convergence-property than k-skip CG.MrR is an alternative minimum residual method for symmetric matrices usingthe coupled two-term recurrences formulated by Rutishauser; initially proposedby Abe et al. The numerical property of k-skip MrR is similar to MrR. However,convergence-property deteriorate as k increases.

The purpose of this study is to improve convergence-properties of k-skip CGand k-skip MrR. Furthermore, parallelization efficiency is numerically evaluated.

P2-66

Model of hydrogen recycling on divertor by

molecular dynamics simulation for neutral

transport analysis in LHD

Seiki SAITO1, Hiroaki NAKAMURA2,3,Takumi SAWADA3, Keiji SAWADA4,

Masahiro KOBAYASHI2, Gakushi KAWAMURA2,Masahiro HASUO5

1. Yamagata University, Japan2. National Institute for Fusion Science, Japan3. Nagoya University, Japan4. Shinshu University, Japan5. Kyoto University, Japan

When the divertor plates are irradiated by hydrogen plasma, some hydrogenatoms and molecules reflect and back to the plasma while the other hydrogenatoms retain in the divertor. This recycled neutral hydrogen atoms and moleculesaffect the plasma parameter of core plasma by ionization and charge-exchangereaction caused during traveling to core plasma via edge plasma. Moreover, in thesituation of rising expectations of detached divertor, recent research revealed thatthe recombination processes caused by hydrogen molecule, so called molecularassisted recombination (MAR), play an important role for the neutral transport atthe edge plasma. In particular, the effects of charge exchange recombination anddissociative recombination described below cannot be ignored for understandingof the process of the neutral transport at the edge plasma.

1. H2 +H+→ H+

2+H, H+

2+ e → H+ H∗, H∗

→ H

2. H2 + e → H− +H, H− +H+→ H+ H∗, H∗

→ H

The transport of the neutral particles can be numerically simulated by neutral-transport code. However, conventional neutral transport codes cannot treat theprocess of MAR. To avoid this problem, Shinshu University group is developinga new neutral-transport code which can treat the process of MAR. In order toprovide the information of recycled hydrogen atoms and molecules to neutral-transport code, we have developed a molecular dynamics simulation model forhydrogen recycling process on carbon divertor. By the hydrogen recycling model,the distributions of emission angle, translational energy, vibrational states, androtational states of emitted hydrogen atoms and molecules are obtained.

P2-67

Integrated Transport Simulation of LHD Plasma

Applying the Ensemble Kalman Smoother

Yuya MORISHITA1, Sadayoshi MURAKAMI1,

Masayuki YOKOYAMA2, Genta UENO3

1. Kyoto University2. NIFS3. The Institute of Statistical Mathematics

Data assimilation is one of the forms of machine learning developed mainly inmeteorology studies. This method finds an optimum combination of a numericalmodel and observations to estimate the state of a system more accurately. In thisstudy, we apply a data assimilation technique to the integrated transport simu-lation (TASK3D) of NBI plasma in LHD (shot:114053). We apply the EnsembleKalman Smoother (EnKS) as a sequential advanced data assimilation method forthe estimation of state variables, that are defined by the electron and ion temper-ature, density, constant factors of turbulence models, and NBI heat depositionevaluated by GNET-TD code. We assimilate the time series of observed data(temperature and density) to the TASK3D simulation. The simulation resultsreproduce accurately both the electron and ion temperature, and the constantfactors of turbulence models are optimized with the spacial and temporal distri-butions. These results show the effectiveness of the data assimilation approach forthe more accurate turbulence modeling and prediction of the behavior of fusionplasmas.

P2-68

Discovery of Alfven-Slow Eigenmodes in

Tokamaks

C. Z. CHENG1,2, G. J. KRAMER2, M. PODESTA2,R. NAZIKIAN2

1. University of Tokyo2. Princeton Plasma Physics Laboratory

Discrete global Alfven eigenmodes such as TAE, RSAE were discovered in toka-mak plasmas from 1980s [1, 2, 3]. Much theoretical and experimental studiedhave confirmed that the discrete Alfven eigenmodes can interact with fast par-ticles and cause significant fast particle loss from the tokamak plasmas. Mostdiscrete Alfven eigenmodes have frequencies above the lowest TAE continuumfrequency. However, recent experiments have found modes with frequencies be-low the TAE frequency continuum gap. In this paper we present the discoveryof new discrete Alfven-Slow Eigenmodes (ASEs) with frequencies below the TAEcontinuum gap based on the full MHD model [4]. In finite pressure plasmasslow magnetosonic modes couple with Alfven waves and the Alfven continuousspectrum below the TAE gap is broken up to form frequency gaps. Using thefull MHD NOVA code several new ASEs are discovered to exist with frequenciesin the Alfven-Slow (AS) gaps without suffering continuum damping. The exis-tence of ASEs is robust for normal and reverse q-profiles and for broad range ofplasma beta values. The slow-mode approximation of the full MHD model doesnot produce these modes because the coupling between Alfven waves and slowmagnetosonic modes are neglected.

[1] C. Z. Cheng and M. S. Chance, Phys. Fluids 29, 3695 (1986) [2] C. Z.Cheng. L. Chen, M. S. Chance, Ann. Physics 161, 21 (1985) [3] Breizman etal., Phys. Plasmas 10, 3649 (2003); G. J. Kramer et al., Plasma Phys. Control.Fusion 46, L23 (2004) [4] C. Z. Cheng, G. J. Kramer, M. Podesta, R. Nazikian,Phys. Plasmas 26, 082508 (2019)

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Investigation into quench detection for a

mutli-stacked pancake coil wound with Nb3Sn

CIC conductors

Tetsuhiro Obana1, Kazuya Takahata1,Haruyuki Murakami2

1. NIFS2. QST

To operate superconducting magnets stably and safety, quench protection for themagnets has to be conducted properly; otherwise superconducting magnets willsuffer serious damage due to quench. In quench protection, accurate detection ofquench occurrence in superconducting magnets is demanded. There are two typesof the quench detection method which is utilized in the magnets for the fusiondevices. One is quench detection based on the balanced voltage between eithertwo coil windings or two segments into which one coil winding is divided. Theother method is quench detection by using a pickup coil whose configuration iseither a co-wound coil or a disk-shaped coil. In this method, the balanced voltagebetween a superconducting coil and a pickup coil is utilized to detect a quench. Inthis study, these two methods of quench detection are investigated experimentallyby using a multi-stacked (MS) pancake coil wound with Nb3Sn cable-in-conduit(CIC) conductors, which is the JT-60SA CS1 module. The experiment with theMS pancake coil was conducted in the coil test facility at the National Institute forFusion Science. This paper describes the measurements of the balanced voltagefor the MS pancake coil by the two quench detection methods. And the quenchdetection methods are discussed based on the measurement results.

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Comprehensive Investigation of 25 Events of

Propagation of Normal-zones in the LHD Helical

Coils

Shinsaku IMAGAWA1,2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences

2. SOKENDAI (The Graduate University for Advanced Studies)

Propagation of normal zones were observed 25 times in a pair of helical coilsof the Large Helical Device in 20 year operation since 1998. The only fourthpropagation resulted in quick discharge due to the imbalance voltage higher thanthe setting value of 0.2 V, whereas the propagation in the other cases stoppedwithin few seconds. Each the coil is divided into three blocks, which are namedH-I, H-M, and H-O from the inside. Since a conductor of the helical coils consistsof NbTi/Cu strands, a pure aluminum stabilizer clad with CuNi layer, and acopper sheath, the current center shifts from the superconducting wires to a purealuminum stabilizer at the normal zone. Therefore, imbalance voltages betweenH1 and H2 are induced in all the blocks during propagation of a normal-zone. Thecross-sectional position of the conductor in which the normal zone propagates canbe estimated from the difference of the imbalance voltages among the blocks. In2002, pickup coils were installed along the helical coils by the pitch of 30 degreeof the poloidal angle in order to detect the position of a propagating normal zone.The pickup coils detect the change of magnetic field by a shift of current center atthe normal zone. The position and the velocity of propagating normal-zones weredetected successfully 15 times after the 10th propagation. They were induced 12times in the H1 coil and three times in the H2 coil. Most of the normal zoneswere induced at the bottom of the coils, and all of them propagated to one side,which is the downstream of the transport current, with recovery at the oppositeside. The propagation velocity can be estimated from the time delay of the peakvoltage of the pickup coils.

Using these measured data, cross-sectional positions were estimated for allthe normal-zones. As the results, 17 and 3 normal-zones were induced in theconductor in the last turn in the first layer of H1-I and H2-I, respectively. 2and 1 normal-zones were induced in the last turn in the second layer of H1-Iand H2-I, respectively. 1 and 1 normal-zone were induced in the first turn inthe second layer of H1-I and H2-I, respectively. From the comparison betweenthe simulation of resistive voltage and the measured data, it is considered thatthe fourth normal-zone firstly propagated to both sides for 0.3 seconds, secondlypropagated to one side with recovery at the opposite side in almost all turns inthe first layer for 20 seconds, and lastly propagated to the neighbor layers.

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AC dissipation current of dielectric material at

high electric field in cryogenic

Tomohiro KAWASHIMA1, Kazuki YAMADA1,Yoshinobu MURAKAMI1, Masayuki NAGAO1,

Naohiro HOZUMI1

1. Toyohashi University of Technology

Dielectric materials have been widely used in a wide temperature range from hightemperatures of several hundred degrees Celsius to cryogenic temperatures. Ina cryogenic temperature, the dielectric properties of materials such as dielectricloss become better than that at room temperature due to suppression of carrierinjections and freezing of dipole motion. Since Joule loss of conductors is alsosignificantly reduced, total electrical loss at cryogenic temperature is much lessthan that at room temperature. In particular, the dielectric loss at room tem-perature does not become large problems because it is smaller than the Jouleloss. However, the small dielectric loss stands out at cryogenic temperatures andit can be one of the factors determining the dielectric breakdown of materials.In general, the dielectric properties of materials are evaluated under low electricfield. They are often used directly for the discussions of the properties under highelectric field. At high field, space charges are easily injected into a material andthe internal field in the material is distorted. As a result, the dielectric propertiesof a material show non-linearity. Since the dielectric loss increases significantlyin high field, there is a risk of misunderstanding when the dielectric propertiesunder high field is discussed by using that under low electric field The currentresponse of dielectric materials under AC field shows dielectric and conductingproperties related to the frequency of AC field. Although tan-delta is commonlyused to evaluate the AC dielectric loss, the information obtained from tan-deltais limited because tan-delta is averaged value indicating the ratio of AC dielectricloss. However, AC dissipation current waveforms contain many useful informa-tion for AC loss mechanisms related to the non-linear behavior of space charges inmaterials. In this paper, the measurement method for evaluating AC loss mecha-nisms of dielectric materials in cryogenic temperature was investigated using ACdissipation current waveforms. The current response of dielectric materials showsthe both a capacitive (IC) component and a conductive (IR: AC dissipation cur-rent) component. The IC component is the 90-degree phase advance comparedto the IR component. Therefore, in order to obtain only the IR component,the signal with 90-degree phase delay was supplied and the IC component wascanceled from the current response. If the amplitude of the IR component isused, it can be evaluated with tan-delta as well as previous method. By usingthis method, it is found that the AC dissipation current shows non-linearity at

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room temperature. In addition, it is confirmed the unique phenomenon that thephase of the waveform shifts with increase of applied electric field. It could beexplained by considering the space charge behavior in the dielectric material. Onthe other hand, we confirmed the significantly increase of tan-delta in cryogenictemperature under high field. The value of tan-delta in cryogenic temperature hasenough values to be detected by AC dissipation current. From the above result,even in cryogenic temperature, it is suggested that the dielectric loss mechanismis possible to evaluate under high field by using AC dissipation current waveform.

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Assessment of electrical insulation performance of

cryogenic fluids using partial discharge waveform

Kazuki YAMADA1, Tomohiro KAWASHIMA1,Tetsuhiro OBANA2, Yoshinobu MURAKAMI1,

Masayuki NAGAO1, Naohiro HOZUMI1

1. Toyohashi University of Technology2. National Institute for Fusion Science

Superconducting technology can be contributed to CO2 reduction and low en-vironmental load as well as realization of energy saving and resource saving byhigh efficiency of energy supply. Development of superconducting equipment havebeen promoted all over the world and the on-site tests were performed in actualpower systems. One of the common basic technologies to guarantee the long-termreliability of superconducting equipment is cryogenic electrical insulation technol-ogy. In general, there are two cooling methods for superconducting equipment:indirect cooling and immersion cooling. In the immersion cooling method, theelectrical insulation systems are mainly composite systems composing a solid in-sulating material and coolant. The electrical insulation performance of coolantsshould be clarified in detail because coolants itself play a role not only for coolingbut also for electrical insulation. It is important to understand partial discharge(PD) characteristics. If PD continually occur, coolants are easily vaporized, andthe electrical insulation performance become lowered as a result. Liquid nitrogenis commonly adopted as a coolant of superconducting equipment. In practice,subcooled liquid nitrogen (66 K) is used, but by considering the boiling point ofliquid nitrogen (77 K), the temperature range available for cooling in the liquidstate is narrow as 10 K. Its heat capacity is also small (20 kJ/kg). As a result,a large amount of liquid nitrogen is required for cooling, and the overall size ofthe superconducting systems become also large. It has been reported that slushnitrogen, which is a mixed two-layer fluid of liquid nitrogen and solid nitrogenparticles, have the excellent cooling performance due to the solid nitrogen parti-cles having high latent heat of melting. There are many reports on the coolingand transport properties of slush nitrogen, but only a few reports on the electri-cal insulation and dielectric performance. In this paper the electrical breakdowncharacteristics of slush nitrogen and liquid nitrogen were investigated. In addi-tion, the assessment of the electrical insulation performance was attempted usingthe difference between the PD waveforms themselves. The PD waveforms in liq-uid nitrogen has a shorter fall time than that in air at room temperature. Thisis because the mobility of ions decreases at cryogenic temperature. It was alsofound that the number of PDs in slash nitrogen generated is 20 times higher thanthat in liquid nitrogen. In addition, its breakdown voltage is reduced to 80 per-

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cent of that in liquid nitrogen due to the surface of solid nitrogen particles actedas insulation weakness. PD waveforms should reflect PD phenomena such as pro-cesses like electron avalanche, development of streamer, and sweep out of ions.The shorter fall time of PD waveform in liquid nitrogen may reflect ion sweepout process. In addition, PD waveforms could be change due to degradation orimprovement of insulation performance. In previous study, it is found that thePD waveforms changes depending on electrical charges on a surface of insulatingmaterials at room temperature. By analyzing the PD waveforms in statistically,the effect of solid nitrogen particles on the electrical insulation performance ofcoolants will be clearer.

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Comparison of the mechanical and

superconducting properties for various

superconducting wires

Hidetoshi OGURO1, Yoshimitsu HISHINUMA2,Akihiro KIKUCHI3, Satoshi AWAJI4

1. Tokai University2. National Institute for Fusion Science3. National Institute for Materials Science4. Tohoku University

A large electromagnetic force is applied to a superconducting wires or cables ina fusion magnet because of the large radius, large current and large magneticfield. This means that the high strength superconducting wires are useful for afusion magnet. We developed some Nb3Sn wires with high strength bronze, suchas a Cu-Sn-Zn. The tensile stresses of the maximum critical currents for highstrength Nb3Sn wires with Cu-Sn-Zn bronze were over 150 MPa, although themaximum point of the critical current for a Nb3Sn wire without reinforcementwas obtained at 100 MPa of the tensile stress. In addition, it is well knownthat the REBCO coated conductors have high superconducting and mechanicalproperties. The typical REBCO coated conductor with Hastelloy substrate keptlarge critical current at 800 MPa tensile stress. These high mechanical propertiesare suitable for using large fusion magnets. In this report, the mechanical andsuperconducting properties of Nb3Sn and REBCO wires were compared.

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Tape Shaped Nb3Al Conductor toward Large

Helical Coils for Future Fusion System

Kyohei YAMADA1,2, Akihiro KIKUCHI1,

Yasuo IIJIMA1, Shigeki NIMORI1,Kiyosumi TSUCHIYA3, Wang XUDONG3,Norihito OHUCHI3, Akio TERASHIMA3,

Yoshimitsu HISHINUMA4, Kazuya TAKAHATA4,Toshiyuki MITO4, Shinsaku IMAGAWA4,Shinji HAMAGUCHI4, Hidetoshi OGURO5,

Tomoaki TAKAO2

1. National Institute for Materials Science, Tsukuba, Ibaraki, Japan2. Sophia University, Tokyo, Japan3. High Energy Accelerator Research Organization, Tsukuba, Japan4. National Institute for Fusion Science, Toki, Gifu, Japan5. Tokai University, Hiratsuka, Kanagawa, Japan

The future Force Free Helical Reactor (FFHR) system requires a pair of helicalcoils with a main radius of 15.6 m. NbTi cables cannot be applied because themaximum magnetic field of the helical coil reaches 11.9 T. Furthermore, largehelical coils exhibit three-dimensional complex architecture, so that is preferableto apply the React & Wind method for coil winding. Since Nb3Al conductorshows an excellent strain tolerance, and the application of the React & Windmethod has been expected. So far, a Fusion research group of National Institutesfor Quantum and Radiological Science and Technology (QST) has demonstratedtrial production of a large solenoid magnet by a React & Wind method using thediffusion processed Nb3Al wires.

We have fabricated tape shaped Nb3Al conductors through the Rapid-Heating/Quenching and Transforming (RHQT) process. Thin tape shaped conductorswould be reduced a bending moment especially with a flat wise direction. It be-comes an advantage for a React & Wind method. In this paper, superconductingproperties as well as microstructures of tape shaped Nb3Al conductors were in-vestigated. In addition, we will also report the bending effect on the non Cu Jcat 4.2 K using brass fixtures having different radius down to 15 mm. Maximumbending strain in this study corresponds to 0.66%.

Acknowledgements

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This study was financially supported by 7 years research project (No. 1; De-velopments in materials processing technology for actualizing technological seedsfor advanced materials and structures) at NIMS and LHD Project CollaborationResearch (No. NIFS18KOBF040) at NIFS.

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Initial studies on large-current high-temperature

superconductor WISE and its application to

helical fusion devices

Yoshiro NARUSHIMA1,2, Shinnosuke MATSUNAGA2,Junichi MIYAZAWA1,2, Nagato YANAGI1,2

1. NIFS2. SOKENDAI

High-temperature super conductor (HTS) is thought to be one of the key materi-als to improve the total performance of the magnetically confined fusion reactor.The critical temperature and the critical magnetic field of the HTS are higherthan those of the Low-Temperature Superconductor (LTS) such as Nb-Ti andNb

3Sn. Because the plasma energy confinement time τ

Eis proportional to the

B0.8, where B is the magnetic field, the plasma performance will be improvedif high magnetic field can be stably produced by using HTS. To realize HTSfusion reactors, there are many issues to be solved. Here, studied are the criti-cal current density of the HTS coil j

cand its application to the helical winding

which is necessary to construct the helical fusion reactor. We have initiated ademonstrative study to develop the conductor manufactured by a concept namedWISE (Wound and Impregnated Stacked Elastic tapes). The flexibility of thisconductor before impregnation gives the advantage for winding helical coils. Fora trial of helical winding, a small helical device named FFHR-0001 is designedand its coil-case is ready and the winding method is under trial. The majorradius of FFHR-0001 is R = 0.0875 m, which corresponds to 1/446 the size ofLHD. The HTS tapes would be installed in this micro helical device. In parallelto the trial of helical winding, an experiment was conducted. Twenty HTS tapeswere stacked and put into a flexible metal tube, and this conductor was insertedinto the aluminum case. The low-melting metal was then poured into the casefor impregnation. This method is expected to stabilize the current path and tokeep the temperature cool, which leads to mitigate a quench phenomenon. Theinitial demonstration was performed in liquid nitrogen at a temperature of 77Kand the degradation was observed. The mechanism for causing this degradationis being investigated and fixed, so that this WISE conductor is applied to thehelical coil winding. Details of the basic R&D and FFHR-0001 will be given inthe presentation.

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Feasibility study of HTS coil cooling assist

technology by magnetic refrigeration

Naoki HIRANO1, Yuta Onodera1, Toshiyuki MITO1,Setsura NAGAI2, YUNZHI XIE2, Tetsuji OKAMURA2

1. NIFS2. Tokyo Institute of Technology

Recently, it has become difficult to supply helium, which has affected cryogenicexperiments. For this reason, we started the development of a cryogenic systemthat does not rely on helium. The feasibility study of HTS coil cooling assist tech-nology by magnetic refrigeration has been studied. Using magnetic refrigerationtechnology, we are focusing on numerical analysis to see if the cooling efficiencyaround 20 K of high-temperature superconducting coils can be enhanced. Mag-netic refrigeration technology needs to apply a change of magnetic field to acertain type of magnetic material, and the configuration of applying a change ofmagnetic field to the material in the cryostat is the point of the proposed coolingassist method. It was confirmed that the magneto caloric effect can be obtainedby changing the magnetic field with the magnetic shield. In addition, we confirma system that uses a magnetic refrigerator in combination with a refrigerant cir-culation cooling system has higher refrigeration performance than the refrigerantcirculation cooling system alone. In this presentation, we report the feasibilityof a highly efficient cooling system near 20K by realizing the exhaust heat ofmagnetic refrigeration by circulating cooling.

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Non-destructive inspection of local defect in HTS

conductor by using magnetization method

Yuta ONODERA1, Toshiyuki MITO1, Naoki HIRANO1

1. NIFS

As a candidate for the next project of the Large Helical Device, continuous wind-ing helical system using high temperature superconductor (HTS) is considered.In manufacturing high current capacity HTS conductor in which stacked REBCOtapes are bundled and are twisted in order to secure mechanical homogeneity ofthe conductor which is necessary for a three-dimensional helical winding. It isimportant to detect local degradation which may occur inside the conductor,because REBCO tape resists uniform compressive stress and tensile stress, butdegradation occurs by the stress of shear or spallation direction. Therefore, itis required the measurement method to detect the local defect in the producedHTS conductor, however, it has not yet established. In this study, we have inves-tigated to inspect the degradation position in the conductor for non-destructivelyby adopting a new magnetization measurement method. Typically, magnetiza-tion method applies magnetic field perpendicular to the REBCO tape surface,and measures the magnetic field generated by the induced current flowing in thetape. However, since the REBCO has a tape shape, the magnetic field cannotbe applied enough when the direction of the external magnetic field and the tapesurface become closed to parallel. Namely, it becomes difficult to magnetize thetwisted conductor depending on the position of the inner REBCO tapes. There-fore, instead of the conventional magnetization by sweeping the external magneticfield, we used a new magnetization method of rotating in a static magnetic field.Using this method, it is confirmed that the magnetization signal strength and thewaveform changed at the degradation position. From this result, it is possible todetermine the defect area for non-destructively by the magnetization signal.

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The Role Of Ionization Versus Transport InSetting Plasma Density Profiles In LAPD

Conor PERKS, Saskia MORDIJCK, Troy CARTER

1. North Carolina State University2. College of William and Mary3. University of California, Los Angeles

In this paper we study the effects of neutral density variations and turbulenceupon the electron density profile on LAPD. Both magnetically confined plasmasfor fusion applications as well as astro-physical plasmas have regions that are amixture of plasma and neutral interactions. LAPD allows us to do experimentsto address how the plasma dynamics vary with various neutral pressures andpower. In short linear devices, 1D theory gives a good approximation for theplasma density and temperature. Conforming to this model, we do see the gen-eral trends predicted such as temperature being set by ionization and thereforegenerally decreasing with increasing neutral pressure and that density is set byincreasing discharge power. In this paper, we will compare the results againstthe 1D theory and investigate the changes in turbulence and particle flux usingprobe measurements. Finally, we will expand the 1D theory to include the mea-sured radial transport effects and investigate whether they improve the matchesto experimental profile measurements.

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The Use of Divertor End Plates as Diagnostics inthe Princeton Field Reversed Configuration II

Justin Ty COHEN1, Charles SWANSON2,

Samuel COHEN1

1. Princeton Plasma Physics Laboratory2. Princeton Satellite Systems

The Princeton Field Reversed Configuration utilizes odd-parity rotating magneticfields to form and heat an FRC. The reactor includes three main compartments,the Center Cell (CC), Source End Cell (SEC), and Far End Cell (FEC). RMFpower is deposited into the Center Cell where it is coupled to a low-density seedplasma that acts as a target. The goal of this experiment was to utilize thedivertor plates within the end cells to characterize plasma parameters withinthose regions of the PFRC-2. It was then possible to use this data to understandparticle and energy transport from the center cell. The divertor plate’s floatingpotential, as well as current to ground during RMF discharges were measured toquantify electron energy and density. At an RMF power of 70 kW, a floatingpotential of -700V was measured on the FEC divertor, which implies electronswith a temperature of 210 eV. An electron density of 9e9 electrons/cm3 at 49kWof RMF, was measured in the SEC using the grounded divertor. When thefloating potential and current to ground were compared to data from the CCinterferometer, there was strong agreement on oscillations seen during electrondensity decay.

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Day3 - November 7, 2019 (Thu)

09:30-10:20 PL3: Richard J. KURTZEffects of Helium and Irradiation Damage on Microstructure and Mechanical Properties of Iron-BaseAlloys for Fusion Structural Applications

10:20-10:50 I3-01: Shaoning JIANGEffect of post-weld annealing on microstructure and hardness of joints between low-activation vanadiumalloy and Hastelloy X alloy

10:50-11:10 Break

11:10-11:40 I3-02: Hiroyasu UTOHDesign studies for key engineering issues related to the JA DEMO concept

11:40-12:00 O3-01: Taku KITASAKANeutronic and thermophysical characteristics evaluation of molten salts specialized for LLFPstransmutation in a fusion reactor

12:00-12:30 I3-03: Anne HOUBENHydrogen permeation and retention in fusion materials and the development of tritium permeation barriers

12:30-13:00 I3-04: Guangnan LUORecent Progress of Tungsten PFCs at ASIPP and Plasma-Tungsten Interactions on EAST

13:00-18:30 Excursion

18:30~ Banquet

Effects of Helium and Irradiation Damage on

Microstructure and Mechanical Properties of

Iron-Base Alloys for Fusion Structural

Applications

Richard J KURTZ

1. Pacific Northwest National Laboratory

The products of a fusion nuclear reaction are a 14.1 MeV neutron and a 3.5 MeValpha particle. The neutrons dissipate their energy in the materials, componentsand structures surrounding the plasma. About ten percent of the incident neu-tron energy is deposited in the first-wall, while the balance is transferred to theblanket behind the first-wall. The neutrons are slowed by nuclear collisions andreactions as they penetrate the reactor structure. Neutrons interact with theatoms of structural materials in two ways. One involves transmutation or conver-sion of one chemical element into another. In iron-base alloys the transmutationreactions of greatest concern introduce large quantities of helium and hydrogenby end-of-life. The other mechanism involves elastic and inelastic scattering col-lisions that displace atoms from their equilibrium positions. The vast majorityof the displaced atoms recombine with a vacant site, but a small fraction do not.These surviving defects can lead to significant changes in mechanical and physicalproperties.

Neutron irradiation at low-temperature results in accumulation of point de-fects to high levels, which causes the yield and tensile strength to increase, andductility and fracture resistance to decrease. The effects of displacement damagesaturates at relatively low neutron dose, but the presence of helium causes in-creased levels of hardening and embrittlment beyond displacement damage alone.At intermediate irradiation temperatures iron-base alloys are susceptible to di-mensional instabilities such as volumetric swelling and irradiation creep. Again,helium can exacerbate these degradation mechanisms. Toward the upper oper-ating temperature regime the effects of displacement damage are less important,but helium might negatively impact service life by promoting grain boundarycreep cavitation. Consequently, development and qualification of structural ma-terials for fusion nuclear service requires knowledge of the effects of displacementdamage and helium over the entire operating temperature range.

Reduced activation ferritic/martensitic (RAF/M) steels are attractive materi-als for first-wall/blanket structural applications in future fusion power plants be-cause they are the most technologically mature option compared to other choices.Nanostructured ferritic alloys (NFA) do not enjoy the same level of technologicalmaturity as RAF/M steels, but this newer material class offers significant advan-

PL3

tages such as the potential for higher temperature operation and greater resistanceto radiation damage. The precise operating temperature and neutron dose lim-its for these materials remains to be fully established because neutron-induceddisplacement damage coupled with helium can lead to significant degradation ofmechanical properties and dimensional instabilities over the entire range of oper-ating conditions. In this paper we highlight recent experiments and modeling tocharacterize the effects of helium and irradiation damage on iron-base alloys forfusion applications.

PL3

Effect of post-weld annealing on microstructure

and hardness of joints between low-activation

vanadium alloy and Hastelloy X alloy

S.N. Jiang1, J.J. Shen2, T. Nagasaka2, K. Hokamoto3,

R. Kasada4, K. Yabuuchi5

1. Qilu University of Technology2. National Institute for Fusion Science3. Institute of Pulsed Power Science, Kumamoto University4. Institute for Materials Research, Tohoku University5. Institute of Advanced Energy, Kyoto University

Vanadium alloys are promising low-activation structural materials for Flibe moltensalt blanket of fusion demonstration power plant and commercial fusion reactors.Hastelloy X alloys are structural materials for the out-vessel components suchas heat exchanger and tritium fuel extractor. Connection is necessary betweenin-vessel cooling pipes of blanket (vanadium alloys) and out-vessel components(Hastelloy X alloy). In the present study, explosive welding (EXW) was used tojoin low-activation vanadium alloy to Hastelloy X alloy. The parent materialsused are plates of V-4Cr-4Ti alloy NIFS-HEAT-2 (NH2) and Hastelloy X (HX)with the material standard of JIS H4551-2000-HW6002. The size of NH2 and HXare 50 × 50× 2.7 mm and 50× 20× 0.25 mm, respectively. In EXW, base plateis HX while flyer plate is NH2 with a stand-off distance of 0.15mm. Cover plateof SUS304 with a thickness of 0.2 mm was located between the specimens stackand explosive. Post-weld annealing at 500℃, 700℃, and 900℃ for 1 h werecarried out respectively. Scanning electron microscope (SEM), and transmissionelectron microscope (TEM) with the sectioning technique with focused ion beam(FIB) machining were used for microstructural analyses. Nanoindentation testwas conducted with the maximum load of 50 mN and loading rate of 5 mN/s.Microstructural evolution and hardness variation at the interface during the postweld annealing were investigated. The results suggested that discontinuous eddyregions with a mixture composition of NH2 and HX appeared at the interface inwhich obvious hardening up to 7.5 9.4 GPa was observed, while the hardness ofNH2 and HX ranged 2.6 3.2 and 3.5 4.6 GPa, respectively, in a range of 100 mfrom the interface. Analysis showed that the eddy region is solid solution withwork hardening due to the deformation at EXW. The hardness decreased afterpost weld annealing at 500℃, while increased after annealing at 700℃. However,the hardness after annealing at 900℃ decreased again. SEM observation showedthat eddy region disappeared after annealing at 900℃. Further TEM observationresults after annealing and their effects on hardness will be discussed.

I3-01

Design studies for key engineering issues related

to the JA DEMO concept

Hiroyasu UTOH1

1. National Institutes for Quantum and Radiological Science and Technology

The Japan pre-conceptual DEMO design is investigated by the Joint SpecialDesign Team for fusion DEMO to establish the Japan’s DEMO concept [1], named“JA DEMO”. Major goals of the JA DEMO are to demonstrate (1) steady andstable electric power generation in a power plant scale, (2) reasonable availabilityusing a remote maintenance scheme anticipated in a commercial plant, and (3)overall tritium breeding to fulfill self-sufficiency of fuel. Since the original conceptof the JA DEMO was defined in JA Model 2014 [2], the pre-conceptual designhas been continued on the fusion DEMO reactor. Main design parameters of theJA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, thenet electricity of 0.2-0.3 GW, and magnetic field on the plasma axis of 6 T.

To demonstrate steady and stable electric power generation in a power plantscale, a magnetic field higher than that of ITER and toroidal field (TF) coil borelarger than that of ITER are necessary. The TF coil design studies indicated thatthe allowable design stress has a large impact on magnetic field. Therefore, theJA DEMO adopts an allowable design stress of 800 MPa, which is larger than thatof ITER TF coil material, JJ1. The current R&D plan for the cryogenic steels isto optimize nitrogen N content in existing steels, for example, JN1 and XM-19.Additionally, in order to mitigate technology gap on tolerances in large coil fab-rication, evaluation of error field based on accuracy of manufacture/assembly ofsuperconducting coils on DEMO were investigated. Assessment of the error fieldhas been performed, indicating that the tolerance can be mitigated by ∼2.5 timesas large as ITER’s with correction coil current of several 100 kAT per coil. Oneof the requirements for the JA DEMO is to demonstrate reasonable availabilityusing a remote maintenance scheme anticipated in a commercial plant. The JADEMO selected the vertical maintenance scheme as the primary maintenance op-tion. In order to improve plant availability, remote maintenance operation andprocess were identified. Based on evaluation and analysis of the maintenancetime, maintenance operation was optimized and updated. As the results, plantavailability of about 70% is foreseeable by four sector work in parallel. The over-all tritium breeding to fulfill self-sufficiency of fuel is one of the key design issuesto be solved. As an alternative option for higher tritium breeding ratio (TBR),the conceptual design of the blanket segment integrated with divertor baffle wasdeveloped. The divertor baffle is installed on the lower blanket segment, and hasthe tritium breeding. This concept leads to increase coverage of tritium breed-ing area. In contrast, the concept of blanket segment integrated with divertor

I3-02

baffle requires complex attitude control on the blanket replacement phase. There-fore, the conceptual design of the in-vessel transferring mechanism and remotehandling equipment for the blanket segment was investigated.

In this presentation, these recent development of engineering design and R&Dongoing for the JA DEMO concept is presented.

[1] K. Tobita et al., Fusion Sci. Technol., 72, 537 (2017).[2] Y. Sakamoto et al., 25th IAEA Int. Conf. on Fusion Energy, FIP/3-4Rb

(2014).

I3-02

Neutronic and thermophysical characteristics

evaluation of molten salts specialized for LLFPs

transmutation in a fusion reactor

Taku KITASAKA1, Hiroki SHISHIDO1,Hidetoshi HASHIZUME1

1. Graduate School of Engineering, Tohoku University

High level radioactive waste disposal the most important issue when using fission-based nuclear energy. Long-lived fission products (LLFPs) are one factor of theradioactivity and decay heat of high level radioactive waste. Transmutation ofLLFPs into non-radioactive or short-lived nuclides is an efficient way to reducethe amount of waste by using fusion neutrons [1]. Earlier studies have proposedto mix LLFPs into Flibe (LiF–BeF2), which is a candidate for a self-cooled liquidblanket system. However, in order to maintain a tritium breeding ratio (TBR),the amount of LLFPs added must be very little. As a result, the amount ofLLFPs transmuted is severely insufficient to contribute to nuclear fuel cycles.

As a solution, this study has proposed new molten salts specialized for trans-mutation of LLFPs. The molten salts consist of LLFP fluorides and BeF2, becauseBeF2 has a low melting point, low neutron absorption and high neutron multi-plication cross sections. By adding larger amounts of LLFPs and arranging themolten salt in very limited regions of the blanket system, it may be possible tomaintain the net TBR and increase the transmuted amount.

In this study a neutron transport and burn-up simulation was conducted toevaluate the feasibility of the new molten salt from a neutronic point of view. Thetarget LLFPs were 135Cs and 107Pd, the most promising nuclides to transmute.Results have indicated so far that the transmuted amount of each nuclides arequite sufficient by placing CsF–BeF2 or PdF2–BeF2 in 5% of the blanket region.More specifically, the transmuted amount was greater than the produced amountfrom 36 units of light water reactors in Japan.

Also in this study, since the structural material has a limit temperature andnuclear heating occurs, an experimental measurement of the melting point andnumerical prediction of Prandtl number was conducted to evaluate the feasibilityfrom a thermal point of view. The experiment was done by raising the temper-ature and visually observing the melting behavior of binary salts CsF–BeF2 andPdF2–BeF2. The Prandtl number was calculated by using the viscosity, specificheat, and thermal conductivity, which were calculated by molecular dynamicssimulations. The molar ratio of LLFP fluoride and BeF2 was changed in order toevaluate the most applicable ratio for the molten salt.

[1] T. Parish, et al., Nucl. Vol. 47 (1980) 324-342

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Hydrogen permeation and retention in fusion

materials and the development of tritium

permeation barriers

Anne HOUBEN1, Jan ENGELS1, Liang GAO2,Jana SCHEUER1, Marcin RASINSKI1,

Arkadi KRETER1, Bernhard UNTERBERG1,Christian LINSMEIER1

1. Forschungszentrum Juelich GmbH, Institut fuer Energie- und KlimaforschungPlasmaphysik, Partner of the Trilateral Euregio Cluster (TEC), 52425 Juelich,Germany

2. Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748, Ger-many

Fuel retention and hydrogen permeation in the first wall of future fusion devicesare crucial factors. Due to safety issues and in order to guarantee an econom-ical reactor operation, tritium accumulation into reactor walls and permeationthrough walls have to be estimated and prevented. Therefore, studies of hydro-gen retention and permeation in the first wall materials are performed and tritiumpermeation barriers (TPB) are developed.

Steels are foreseen as structure materials in fusion devices. Two kinds ofsteels, the ferritic-martensitic Eurofer97 steel, foreseen for the use in DEMO,and the austenitic 316L(N)-IG steel, implemented in ITER, were investigated.Detailed permeation studies were performed on this steels. Furthermore, theinfluence of a technical sample surface on the hydrogen permeation was studied.Since the structure material will be exposed to high energy deuterium and tritiumparticles in ITER and future fusion devices, steel samples were exposed in thelinear plasma device PSI-2 to a deuterium plasma. The influence of the exposureon the deuterium retention and permeation was studied in order to estimate theperformance in a fusion device.

Ceramics, such as metal oxides, carbides and nitrides, were identified as hightemperature resistant materials with a low hydrogen permeation. The hydrogenpermeation through various materials was measured in the last decades and com-pared. Due to the strong influence of the microstructure of the coating on thehydrogen permeation, the obtained results can vary widely. Therefore, in order tounderstand the hydrogen permeation process through the TPB, a detailed phaseand microstructure analysis is important.

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Y2O3 and WN layers prepared by magnetron sputter deposition on steel sub-strates were developed for TPB and studied. The stoichiometry, the crystalstructure and the microstructure of the deposited and annealed layers were in-vestigated by X-ray diffraction and scanning electron microscopy. The deuteriumpermeation was studied and compared to a reference measurement of an uncoatedsubstrate in order to determine the permeation reduction factor (PRF). In Y2O3,the influence of the microstructure on the hydrogen permeation was investigatedin detail. The PRFs of two Y2O3 samples with different microstructures differby two orders of magnitude. Thus, by improving the microstructure of the TPB,the permeation reduction behavior is strongly enhanced.

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Recent Progress of Tungsten PFCs at ASIPP and

Plasma-Tungsten Interactions on EAST

G.-N. Luo1, F. Ding1, Q. Li1, W.J. Wang1,X.H. Chen1, H. Xie1, R. Ding1, C. F. Sang2,Z.H. Hu1, Z. Sun1, L. Wang1, Y.W. Sun1,

J.S. Hu1, D.M. Yao1, EAST Team1

1. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei,230031 China

2. Dalian University of Technology, Dalian 116024, China

The EAST W/Cu upper divertor was finished in 2014, aiming at achieving ac-tively cooled full W/Cu-PFCs for the upper divertor, with heat removal capa-bility up to 10 MW m−2. Commissioning of the divertor was successful in the2015 Spring Campaign. In collaboration with the IO and CEA teams, ASIPPhas demonstrated technological capability to remove heat load of 5000 cyclesat 10 MW m−2 and 1000 cycles at 20 MW m−2 for the small scale monoblockmockups, and surprisingly over 300 cycles at 20 MW m−2 for the flat-tile ones.ASIPP won the competitive bidding and is now manufacturing the 456 monoblockW/Cu-PFUs for the WEST tokamak at CEA. ASIPP has also started a projectto upgrade the EAST lower divertor into full W/Cu-PFCs in 2020. Since 2015,great efforts have been being made to investigate W erosion on EAST via ex-periment and simulation to understand and control of the sources of W impurityin plasma. The results indicate that carbon is the dominant impurity causingW sputtering in L-mode plasmas. The mixture layer on the surface of W PFCsformed through redeposition or the wall conditioning can effectively suppress Werosion. Increasing the plasma density and radiation can reduce incident ion en-ergy, thus alleviating W sputtering. In H-mode plasmas, control of edge localizedmode (ELM) via resonant magnetic perturbation (RMP) proves to be capable ofsuppressing intra-ELM W erosion.

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Day4 - November 8, 2019 (Fri)

09:30-10:20 PL4: Masaki OSAKABERecent results of LHD deuterium experiment

10:20-10:50 I4-01: Naomichi EZUMIPlasma detachment in divertor simulation experimental module of GAMMA 10/PDX: role of moleculegases

10:50-11:10 Break

11:10-11:40 I4-02: Samuel Aaron LAZERSONFirst Neutral Beam Experiments on Wendelstein 7-X

11:40-12:10 I4-03: Masaomi TANAKAAstrophysics of gravitational wave sources and atomic data for heavy elements

12:10-12:40 I4-04: Shinji SAITOParticle acceleration and heating in magnetosonic/whistler mode turbulence

12:40-13:40 Lunch

13:40-14:10 I4-05: Hiroya YAMAGUCHINon-equilibrium plasma in astrophysical objects

14:10-14:30 O4-01: Hiroki HACHIKUBOMagnetic field dependence of transition to high ion saturation current phase in a linear plasma deviceNUMBER

14:30-14:50 O4-02: Naoto TSUJIIModeling of the lower-hybrid wave driven plasma equilibrium with a hybrid-MHD model on the TST-2spherical tokamak

14:50-15:10 Break

15:10-15:30 O4-03: Matthijs VAN BERKELSystem Identification and Real-time Control of the CIII Emission Front using MANTIS in TCV

15:30-15:50 O4-04: Hao WANGThe Systematic Study of Energetic Particle Driven Geodesic Acoustic Mode Channeling in LHD

15:50-16:10 Closing

Recent results of LHD deuterium experiment

Masaki OSAKABE1,2, Hiromi TAKAHASHI1,2,Sadayaoshi MURAKAMI3, Kenji TANAKA1,4,Shigeru INAGAKI4, Suguru MASUZAKI1,2,Mizuki SAKAMOTO5, Yasuhiro SUZUKI1,2,

Kazunobu NAGASAKI3, Tomohiro MORISAKI1,2

1. National Institute for Fusion Science2. The Graduate University for Advanced Studies, SOKENDAI3. Kyoto University4. Kyushu University5. University of Tsukuba

The deuterium (D) experiment was started since March, 2017on the Large HelicalDevice(LHD) and is the first full-fledged one performed in stellarator/heliotrondevices. The main objectives of the D experiment are as follows: (1) achievementof high performance plasma through confinement improvement due to isotopeeffect, (2) clarification of the isotope effect by experimental and theoretical ap-proaches, (3) demonstration of the confinement capability of energetic particles(EPs) in helical device, and (4) enhancement in research on the plasma wall in-teractions. In the first D experiment, LHD established one of the most importantmilestones towards the realization of the helical fusion reactor, i.e., ion tempera-ture Ti of 10 keV. This is the highest record among stellarator/heliotron devices.Clear reduction of the ion thermal diffusivity in both core and edge regions in Ddischarge from H was identified, indicating the confinement improvement by theisotope effect. This result was qualitatively consistent to the theoretical predic-tion by the gyrokinetic code, GKV. The neutron diagnostic in D plasma revealedthe good EP confinement property of inwardly shifted magnetic axis configu-ration of LHD. Precise measurement of the tritium exhaust demonstrated thetritium mass balance including the evacuation system. In the recent experiment,we could successfully expand the operational regime further. By increasing theECH power and optimization in its injection scheme, Te increased more than 6.5keV, while the Ti of 10 keV was maintained. In this conference, new findingsfrom the first D experiment aiming for the above objectives are reviewed, andvery recent results from the ongoing experiment will be presented.

PL4

Plasma detachment in divertor simulation

experimental module of GAMMA 10/PDX: role

of molecule gases

N. Ezumi1, R. Perillo2, T. Iijima1, K. Nojiri1,A. Terakado1, M. Sakamoto1, Y. Nakashima1,

M. Hirata1, J. Kohagura1, M. Yoshikawa1, T. Hara1,Y. Ando1, H. Gamo1, T. Sugiyama1, Y. Takami1,K. Honma1, A. Kondo1, T. Kuwabara3, H. Tanaka3,N. Ohno3, K. Sawada4, S. Kado5, A. Tonegawa6,

S. Masuzaki7

1. Plasma Research Center, University of Tsukuba2. Dutch Institute for Fundamental Energy Research3. Graduate school of Engineering, Nagoya University4. Faculty of Engineering, Shinshu University5. Institute of Advanced Energy, Kyoto University6. Graduate school of Science, Tokai University7. National Institute for Fusion Science

Formation and control of detached plasma are key issues for handling heat and/orparticle load to the plasma facing components and managing high-performancecore and edge-divertor plasmas of magnetic fusion devices, ITER and a demoreactor. In Plasma Research Center, University of Tsukuba, divertor simulationexperiments have been conducted at the end region of the tandem mirror deviceGAMMA 10/PDX utilizing the high temperature end loss plasmas which have afew hundred eV for ion temperature.

So far, characterizations of plasma detachment caused by various gas seedingin the divertor simulation experimental module (D-module) located at the end-loss region have been performed [1,2]. We have recently investigated synergisticeffect on plasma detachment by combined seedings of noble and molecular gassessuch as N2, Ne, Ar, Kr, Xe, and/or H2, in V-shape target region at the D-module.As one of notable results, we have found a combination of N2 and H2 puffs showedclear decrease of ion flux. The results indicate the possibility of achieving areliable divertor operation scheme and the importance of a deeper understandingof the catalytic molecular assisted recombination process enhanced by H2 andN2, that is supposed to play important role on particle/heat load mitigation inthe fusion-relevant magnetically confined devices [3,4]. In this paper, we showexperimental and analytical progresses of understanding the synergistic effect of

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nitrogen and hydrogen gas seeding on detached plasma. Influence of openingangle of the V-shaped target plate on the phenomena is also discussed.

This work is partly supported by JSPS KAKENHI Grant Number 19K03790,and NIFS Collaboration Research program (NIFS16KUGM110, NIFS19KUGM146).

[1] Y. Nakashima et al., Nuclear Fusion 57 (2017) 116033.[2] M. Sakamoto et al., Nucl. Mater. Energy, 12 (2017) 1004.[3] N. Ezumi et al., Nucl. Fusion 59 (2019) 066030.[4] R Perillo et al., Plasma Phys. Control. Fusion 60 (2018) 105004.

I4-01

First Neutral Beam Experiments on Wendelstein

7-X

Samuel Aaron LAZERSON1, Dirk HARTMANN1,Paul MCNEELY1, Norbert RUST1,

Sergey BOZHENKOV1, Simppa KSLOMPOLO1,Oliver FORD1, Lila VNO1, Kunihiro OGAWA2,

Mitsutaka ISOBE2, Dorothea GRADIC1,Torsten STANGE1, Marcin JAKUBOWSKI1,

Robert WOLF1

1. Max-Planck-Institut fr Plasmaphysik, Greifswald, Germany2. National Institute for Fusion Science, Toki-city, Japan

Stellarators must demonstrate adequate energetic particle confinement to moveforward as a re- actor design concept, having inherently demonstrated steady-state and transient free operation. To assess this topic, the Wendelstein 7-X(W7-X) experiment has recently commissioned the first of two 55 keV neutralbeam injectors (NBI) during the inertially cooled divertor campaign. Two ra-dio frequency driven positive ion hydrogen sources capable of injecting 1.5 MWper source were installed in the injector. The sources demonstrated up to 5 sof continuous, full power injection into the W7-X plasma. Experiments explor-ing the role of energetic particles in W7-X were conducted by firing the NBIsystem into electron cyclotron resonance heated (ECRH) plasmas . These ex-periments scanned both plasma density and magnetic configurations to explorethe underlying physics of energetic particle confinement in stellarators. Infraredcamera measurements of first wall structures confirmed the effectiveness of newlyinstalled shielding collars on camera immersion tubes. This confirms estimatesof wall loads made by the ASCOT code. Experiments where the plasma wasfully supported by NBI were also con- ducted. In these discharges, ECRH is usedto start the plasma and then supported by pure NBI injection. This allowedachievement of high plasma densities, opening the possibility of O-X-B ECRHoperation with the addition of two more sources in the next campaign. The roleof ECRH in density control was also evaluated using these plasmas.

I4-02

Astrophysics of gravitational wave sources and

atomic data for heavy elements

Masaomi TANAKA1

1. Tohoku University

In 2017, the first gravitational wave observation from a neutron star merger wassuccessfully made. The detection triggered electromagnetic wave observationsover the entire electromagnetic wavelengths, which enabled the first identifica-tion of an electromagnetic counterpart of a gravitational wave source. In the ul-traviolet, optical, and infrared wavelengths, the counterpart shows characteristicproperties of ”kilonova”, electromagnetic emission powered by radioactive decaysof newly synthesized heavy elements. By this observations, production and ejec-tion of heavy elements by the neutron star merger is confirmed. However, exactabundances of heavy elements are still poorly understood. Development of accu-rate, complete atomic data is a key to understand the heavy element synthesisby neutron star merger.

In this talk, I introduce astrophysics involved in neutron star mergers andsummarize what we have learned about heavy nucleosynthesis by joint observa-tions of gravitational waves and electromagnetic waves. Then, I highlight theimportance of atomic data for heavy elements.

I4-03

Particle acceleration and heating in

magnetosonic/whistler mode turbulence

Shinji SAITO1, Yasuhiro NARIYUKI2,

Takayuki UMEDA3

1. National Institute of Information and Communications Technology2. Faculty of Human Development, University of Toyama3. Institute for Space-Earth Environmental Research, Nagoya University

Magnetic fluctuations in magnetohydrodynamics scales typically shows a mag-netic spectrum with the power-law index about -5/3 in the solar wind. Thefluctuation energy cascades into smaller scales, and expected to be dissipatedat kinetic scales. It implies that kinetic fluctuations have an important role toaccelerate and/or heat plasma particles in the solar wind. We have studied parti-cle heating and acceleration in decaying magnetosonic/whistler mode turbulenceby using results of two-dimensional fully kinetic particle-in-cell simulation. Thesimulation study suggests that ions can be accelerated in the turbulence intermit-tently, while electrons have anisotropic heating efficiently along the backgroundmagnetic field. The simulation results suggest a fundamental understanding ofparticle scattering in magnetosonic/whistler mode turbulence.

I4-04

Non-equilibrium plasma in astrophysical objects

Hiroya YAMAGUCHI

1. ISAS/JAXA

Hot plasmas are ubiquitous in the Universe, from solar coronae to galaxy clusters.Matters are highly ionized in those objects, such that the ionization and recom-bination rates are in equilibrium. However, in some ’young’ objects, such as solarflares and supernova remnants, plasmas are often in non-equilibrium ionization(NEI) condition. I will present observational characteristics of the astrophysicalNEI plasmas, focusing on some important atomic processes. For instance, in-nershell ionization and fluorescence frequently take place in a high-temperatureionizing plasma, whereas radiative and dielectronic recombinations become dom-inant in a recombining plasma. X-ray emission associated with these processesoffers powerful diagnostics on the NEI condition. I will review recent studies ofNEI plasmas, and discuss prospects for future high-resolution spectroscopy.

I4-05

Magnetic field dependence of transition to high

ion saturation current phase in a linear plasma

device NUMBER

Hiroki HACHIKUBO1, Atsushi OKAMOTO1,Takaaki FUJITA1, Hideki ARIMOTO1,

Ryosuke OCHIAI1, Minami SUGIMOTO1,Kento IIZUKA1

1. Graduate School of Engineering, Nagoya University

High density plasma is required for understanding and controlling the recombi-nation process in divertor plasma and for researching and developing high energyion generation methods applicable to alpha particle simulation in small annulardevices. For these purposes, a linear plasma device NUMBER (Nagoya UniversityMagnetoplasma Basic Experiment) was developed. NUMBER has a diameter of0.2 m and a total length of 1.8 m, and is axially divided into a production regionand a test region. Magnetic field of the production region is in steady operationby a DC power source, and magnetic field of test region is in pulse operation usinga capacitor bank. Plasma is generated by electron cyclotron resonance (ECR)using microwave (2.45 GHz). When a pulsed magnetic field is applied to the testregion, the test region magnetic field is increased, and the plasma generated inthe production region is transported to the test region along the magnetic fieldlines [1]. High density discharge mode was observed when the test region mag-netic field strength (Btest) was high, and a model considering the magnetic mirroreffect was proposed [2].In the experiment where the maximum of the test region magnetic field (Bmax

test)

during the pulse was scanned, a rapid increase or decrease of the ion saturationcurrent was observed with a Langmuir probe under the condition that Bmax

testwas

0.09 T or more. But, it was not observed under the condition that Bmax

testwas

0.08 T. Under the conditions of Bmax

testwas 0.23 T and 0.15 T, the ion saturation

current rapidly decreases when Btest became around 0.1 T during its decay phase.In order to observe this phenomenon clearly, we examined and found the ener-gization waveform where the time change of the magnetic field intensity is moregradual.A part of this research is supported by JSPS KAKENHI Grant Numbers JP19H01869,JP17H06231.

[1] D. Hamada, et al., Plasma Fusion Res. 13 3401044 (2018).[2] A. Okamoto, et al., Plasma Fusion Res. 14 2401005 (2019).

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Modeling of the lower-hybrid wave driven plasma

equilibrium with a hybrid-MHD model on the

TST-2 spherical tokamak

Naoto TSUJII1, Yusuke YOSHIDA1, Yuichi TAKASE1,Akira EJIRI1, Osamu WATANABE1,

Hibiki YAMAZAKI1, Yi PENG1, Kotaro IWASAKI1,Yongtae KO1, Kyohei MATSUZAKI1, James RICE1,

Yuki OSAWA1

1. The University of Tokyo

Removal of the central solenoid is considered essential for spherical tokamaks, butnon-inductive plasma start-up is a challenge. Start-up using lower-hybrid waveshas been studied on the TST-2 spherical tokamak at the University of Tokyo. Theequilibrium poloidal field is believed to be generated mostly by the wave drivenfast electrons, which have large orbit excursions from the flux surfaces due to lowplasma current. Such an equilibrium can be significantly different from the Grad-Shafranov equilibrium routinely used for internal magnetic field reconstruction ina tokamak. In this work, the effect of fast electrons on the equilibrium wasinvestigated by fitting the magnetic data on the hybrid-MHD model [Y. Todoand A. Bierwage, Plasma Fus. Res. 9, 3403068]. The fast electron distributionfunction was estimated from ray-tracing and Fokker-Planck calculations. Sincethe time evolution of the wave driven fast electrons is much slower than theMHD time scale, the estimated distribution function was introduced to the MHDmodel as a stationary distribution function. The bulk fluid force balance wasthen solved, and the solution was successfully fitted to the magnetic data. Theresulting bulk pressure was lower than that without any fast electrons, which ismore consistent with the Thomson scattering measurement.

O4-02

System Identification and Real-time Control of

the CIII Emission Front using MANTIS in TCV

Matthijs VAN BERKEL1,3, Timo RAVENSBERGEN1,3,

Artur PEREK1, Ricky VAN KAMPEN1,3,Joost LAMMERS3, Basil DUVAL2, Olivier FEVRIER2,

Cristian GALPERTI2, Christian THEILER2,Bryan LINEHAN4, Ivo CLASSEN1, Marco DE BAAR1,3,

TCV team*, EUROfusion MST1 team**

1. Dutch Institute for Fundamental Energy Research2. Swiss Plasma Centre, EPFL3. Eindhoven University of Technology4. Plasma Science and Fusion Center, MIT

One of the major challenges in realizing a commercially viable fusion reactoris the handling of the power and particle exhaust in the divertor. A proposedapproach is operation in the so-called detached regime. As actuator, we use (neu-tral) local gas puffing (carbon machine) which allows for controlling the powerloss mechanisms, and hence can significantly reduce the particle and heat fluxesat the divertor target. However, such a detached regime has its internal dynam-ics and is sensitive to external perturbations that need to be controlled throughactive monitoring and control of the gas puff actuator in real-time. A new di-agnostic called MANTIS has been developed with which we sense the radiationfront location in real-time in TCV with filtered imaging. In a detached TCVL-mode plasmas, we assume that the radiation front is well approximated bythe C-III emission front. The poloidal location of this front can be estimatedin real-time by a newly developed image processing software acting upon directcamera images. The present set-up features a temporal resolution of 5 ms and a-hardware limited- latency of 1 ms. A gas injection valve with Deuterium gas wasused as actuator. Using this set-up, we used state-of-the-art system identificationtechniques to identify the input-output dynamics of the detached plasmas usinga periodic multi-sine system identification approach in two L-mode discharges.This yielded the transfer function and covariance on the measurements from thegas valve actuation to the measured front location. This was used to synthesizea conservative controller off-line. Subsequent closed loop experiments were car-ried out in which good tracking of the radiation position reference was achievedwithout additional tuning of the feedback controller.

*See author list of“S. Coda et al 2019 Nucl. Fusion accepted (https://doi.org/10.1088/1741-4326/ab25cb)”**See the author list ”B. Labit et al 2019 Nucl. Fusion accepted(https://doi.org/10.1088/1741-4326/ab2211)”

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The Systematic Study of Energetic Particle

Driven Geodesic Acoustic Mode Channeling in

LHD

Hao WANG1, Yasushi TODO1, Masaki OSAKABE1,Takeshi IDO1, Yasuhiro SUZUKI1,2

1. National Institute for Fusion Science, National Institutes of Natural Sci-ences, Toki 509-5292, Japan

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

The energy channels established during the energetic particle driven geodesicacoustic modes (EGAMs) in the Large Helical Device (LHD) plasmas are studiedusing MEGA code. MEGA is a hybrid simulation code for energetic particlesinteracting with a magnetohydrodynamic (MHD) fluid. In the present work,both the energetic particles and bulk ions are described by the kinetic equations.A realistic 3-dimensional equilibrium generated by HINT code is used for thesimulation.

The properties of EGAM channeling is systematically studied for the firsttime. Five conclusions are found as follows. First, during the non-chirping EGAMactivities, EGAM channeling occurs in the linear growth stage but terminates inthe nonlinear saturated stage; while during the chirping EGAM activities, EGAMchanneling occurs continuously in both linear growth stage and nonlinear satu-rated stage. Second, the bulk ion heating power increases with the EGAM ampli-tude, because stronger mode activity transfers more energy to the bulk ions. Butthe energy transfer efficiency (Eion/EEP ) is not sensitive to the EGAM amplitude,because both the energy absorption of bulk ions and the energy loss of energeticparticles changes together. Third, lower frequency EGAMs make higher energytransfer efficiency, because the interactions between lower frequency mode andbulk ions are stronger. The EGAM mode frequency decreases with the increase ofenergetic particle pressure, while the EGAM energy transfer efficiency increaseswith energetic particle pressure. Also, the EGAM mode frequency increases withneutral beam velocity, while the EGAM energy transfer efficiency decreases withthe increase of neutral beam velocity. Fourth, in the case of deuterium plasmaand deuterium beam, the energy transfer efficiency is lower than that of the hy-drogen plasma and hydrogen beam. Last, the energy transfer efficiency increaseswith the decrease of dissipation coefficients. Less energy dissipates by decreasingthe dissipation coefficients, and thus, more energy can be transferred to the bulkions.

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