The 25th
International Toki Conference
Creating the Future:
ー Innovative Science of Plasma and Fusion ー
3rd - 6th November, 2015, Ceratopia Toki, Toki, Gifu, Japan

Book of Abstracts

  
  
  
  
  
  
  
  
  
  
  
  
  
  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

25th  International  Toki  Conference  (ITC-­25)    Creating  the  Future  –  Innovative  Science  of  Plasma  and  Fusion  –  Book  of  Abstracts  3rd  –  6th  November,  2015  Ceratopia  Toki,  Toki-­city,  Gifu,  JAPAN  

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Scope  The  International  Toki  Conference  is  a  topical  meeting  dedicated  to  a  specific  subject  related  to  plasma  and  fusion  science.  The  series  began  in  1989  with  the  founding  of  the  National  Institute  for  Fusion  Science  and  has  been  held  on  an  annual  basis  ever  since.  The  last  10  years  have  seen  great  advances  in  the  plasma  and  fusion  sciences  based  on  progress  in  basic  physics  and  technology.  These  developments  related  to  plasma  and  fusion  sciences  also  accelerate  innovative  applications  of  plasmas  in  industrial,  biomedical,  agricultural,  and  other  fields.  With  this  background,  we  are  planning  the  25th  International  Toki  Conference  to  explore  new  aspects  in  creating  the  future  through  the  innovative  sciences  of  plasma  and  fusion.  
    Topics  Helical  Plasma  Research  Tokamak  and  ITER  Related  Research  Innovative  Applications  of  Plasmas              -­  Industrial/Chemical/Medical  and  Agricultural  Applications  -­  Basic  Plasma  Science  -­  Laboratory  and  Space  Plasmas  -­  Fusion  Technology  Research  Theory  and  Simulation  Research  Laser  Plasma  Sciences  Plasma  Wall  Interaction  Research  Other  Fields  of  Sciences  
  
  
  

General  Contact  Address  National  Institute  for  Fusion  Science  322-­6  Oroshi-­cho,  Toki  509-­5292,  Japan  E-­mail:itc25@nifs.ac.jp  
  
  
  
  
  
  
  
  
  
  
  
  
The   25th   International   Toki   Conference   is   organized   by   National   Institute   for   Fusion  Science   (NIFS),   and   is   supported   by   SOKENDAI   (The   Graduate   University   for  Advanced   Studies),   National   Institutes   of   Natural   Sciences   (NINS)   and   YU-­KWAI  (Fusion  Science  Association).    
  -3-International  Advisory  Committee  Y.  Takeiri  (Japan,  Chair)     N.  Ohno  (Japan)     K.  Hanada  (Japan)     M.  Mori  (Japan)     T.  Morisaki  (Japan)     A.  Sagara(Japan)     T.  Mito  (Japan)     K.  Nishimura  (Japan)     T.  Klinger  (Germany)     V.  Antoni  (Italy)      H.  Azechi  (Japan)     Y.  Ogawa  (Japan)     A.  Fukuyama  (Japan)     M.  Wada  (Japan)     M.  Osakabe  (Japan)     S.  Imagawa  (Japan)     T.  Muroga  (Japan)     H.  Sugama  (Japan)     M.  Zarnstroff  (USA)       International  Program  Committee  M.  Ichimura     S.  Kubo     (Japan,  Chair)     (Japan,  Vice-­Chair)     H.  Laqua  (Germany)     V.  Antoni  (Italy)     J.  Noterdaeme  (Germany)      M.  Nagatsu  (Japan)     S.  Murakami  (Japan)     A.  Hatayama  (Japan)     M.  Inomoto  (Japan)     Y.  Nakamura  (Japan)     I.  Murakami  (Japan)     Y.  Suzuki  (Japan)     K.  Tsumori  (Japan)     S.  Ohdachi  (Japan)     K.  Nagaoka  (Japan)     T.  Ii  (Japan)      Local  Organizing  Committee  (LOC)  T.  Mutoh  (Chair)     K.  Nagaoka     (Scientific  Secretary)     M.  Osakabe     S.  Ohdachi     Y.  Yoshimura     K.  Ikeda     M.  Kisaki     A.  Ito     T.  Kobayashi     R.  Makino       T.  Shimozuma     K.  Tsumori     T.  Seki     H.  Kasahara     H.  Takahashi     S.  Kamio     T.  Ii     M.  Nakata     S.  Sugie  
  
  
  
  

  

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A.  Ando  (Japan)  Y.  Nakashima  (Japan)  T.  Mizuuchi  (Japan)  O.  Kaneko  (Japan)  R.  Horiuchi  (Japan)  T.  Mutoh  (Japan)  S.  Ishiguro  (Japan)  S.  Sakakibara  (Japan)  M.  Kwon  (Korea)    T.  Mutoh  (Japan)  M.  Ono  (USA)  Y.  Sakawa  (Japan)  H.  Idei  (Japan)  H.  Sugama  (Japan)  J.  Miyazawa  (Japan)  T.  Nagasaka  (Japan)    S.  Kubo  Y.  Nakamura  K.  Saito  H.  Igami  H.  Nakano  R.  Seki  S.  Takada  H.  Noto  T.  Kato    Contents      Program     

  
  
  
  

  
Tuesday  Wednesday  Thursday  Friday    Poster    Session  1    Session  2    3rd  November  4th  November  5th  November  6th  November      3rd  November  4th  November  Abstracts  of  Papers     Plenary  /  Invited  /  Oral  presentations    Tuesday  3rd  November    Wednesday  4th  November    Thursday  5th  November    Friday  6th  November      Poster  presentations       Session  1  3rd  November    Session  2  4th  November  

  
  
  

  
  
  
  
  
  
  
  
  

  
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Program    Tuesday,  3rd  November,  2015    Time  

  9:00  -­      9:20        

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Presenter  Title  Registration    9:20  -­      9:25                 Opening  and  Welcome  Addresses    9:50  -­  10:10                 Group  Photo                            10:10  -­  10:55                  Plenary  1  session  (M.  Ichimura)         10:10  

              PL  -­   1     N.J.  Fisch  Pushing  Particles  with  Waves:  Current  Drive  and  Alpha  Channeling  10:55  -­  11:15                 Coffee  Break  11:15  -­  12:15                  Invited  1  session  (M.  Ichimura)  

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11:15  

      I   -­   1     M.  Osakabe  Current  status  of  the  LHD  project  and  its  prospect  for  deuterium  experiment  24  11:45  

      I   -­   2     K.  Itoh  Non-­equilibrium  and  Extreme  State  of  Plasmas  25  12:15  -­  14:15                 Lunch                          14:15  -­  16:15                  Poster  Session  1            P1-­1  ~  P1-­118                        16:15  -­  16:35                   Coffee  Break  16:35  -­  18:15                  Invited  2  /  Oral  1  session  (M.  Osakabe)    Page  

  
  
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16:35  

      I   -­   3     M.  Shiratani  Plasma  assisted  enhancement  of  agricultural  yield  26  17:05  

      I   -­   4     K.  Kitano  Innovative  medical  technology  in  plasma  disinfection  of  human  body  with  low-­temperature  atmospheric-­pressure  plasmas  ~  the  reduced-­pH  method  and  the  plasma-­treated  water  (PTW)  ~  27  17:35  

      O   -­   1     H.  Saitoh  Recent  status  of  the  PAX  and  APEX  projects  toward  the  formation  of  electron-­positron  plasma  28  17:55  

      O   -­   2     M.  Sato  An  Application  of  Fusion  Oriented  Neutrons  for  Transmutations  of  Long  Life  Fission  Products  

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Wednesday,  4th  November,  2015    Time  

  

No  

  

Presenter  Title  

9:00  -­  10:15                  Plenary  2  /  Invited  3  session  (N.J.  Fisch)    9:00  

      PL  -­   2     M.  Ono  Spherical  Tokamaks  and  Fusion  Energy  Development  Path  31  9:45  

      I   -­   5     H.  Meyer  Research  towards  operation  with  tolerable  ELMs  on  European  Tokamaks  32  Coffee  Break    10:15  -­  10:35                 10:35  -­  12:05                  Invited  4  session  (A.  Sagara)  10:35  

      I   -­   6     H.  Tamura    Design  status  of  structural  components  of  helical  fusion  reactor  FFHR-­d1  33  11:05  

      I   -­   7     M.  Sakamoto  Hydrogen  recycling  in  the  divertor  simulation  plasma  on  GAMMA  10/PDX   34  

11:35  

      I   -­   8     Y.  Ueda  Interactions  of  tungsten  plasma  facing  material  with  edge  plasma  35  12:05  -­  13:35                 Lunch  

  13:35  -­  14:45                  Invited  5  /  Oral  2  session  (C.Z.  Cheng)  13:35  

      I   -­   9     B.  Unterberg  14:05  

      O   -­   3     Y.  Ueki  14:25  

      O   -­   4     A.  Kimura  Research  on  Plasma  Wall  Interaction  Facilities  for  Fusion  Reactors  

  
36  

Velocity  profile  measurement  of  lead-­lithium  duct  flow  by  ultrasonic  Doppler  velocimetry  37  Impacts  of  Material  Innovation  on  Advanced  Blanket  Design  38  Coffee  Break    14:45  -­  15:05                 15:05  -­  15:35                  Invited  6  /  Oral  3  session  (B.  Unterberg)    15:05  

      I   -­  10     A.  Kasugai  Progress  of  Linear  IFMIF  Prototype  Accelerator  in  Rokkasho  39  15:35  

      O   -­   5     H.  Shishido  Discussion  on  the  effect  of  NaF  on  physical  properties  of  Flinabe  in  comparison  to  Flibe  from  molecular  dynamics  simulations  40  Experimental  and  numerical  study  of  H-­  production  in  DC  H-­  source  of  medical  cyclotron  41  15:55  

      O   -­   6     H.  Etoh                       16:15  -­  18:15                  Poster  Session  2         P2-­1  ~  P2-­89    Page  

  

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Thursday,  5th  November,  2015    Time  

  No     

9:00  -­  10:15           Title  

      Plenary  3  /  Invited  7  session  (H.P.  Laqua)  Page  
  9:00  

      PL  -­   3      J-­M.  Noterdaeme  Innovations  that  make  the  Ion  Cyclotron  Range  of  Frequency  Power  suitable  for  Fusion  Reactors  43  9:45  

      I   -­  11     44  S.  Murakami  Predictions  of      Plasma  Performance  and  Fusion  Reactions  in  the  Deuterium  Experiment  Plasma  of  LHD  10:15  -­  10:35                10:35  -­  12:25                 Invited  8  /  Oral  4  session  (H.  Meyer)  Coffee  Break      10:35  

      I   -­  12     G.  Serianni  The  full-­size  source  and  injector  prototypes  for  ITER  neutral  beams  45  11:05  

      I   -­  13     K.  Ikeda  Recent  Studies  of  Hydrogen  Negative  Ion  Source  and  Beam  Production  for  NBI  in  Large  Helical  Device  46  11:35  

      I   -­  14      N.C.  Luhman,  Jr.   Advances  in  2-­D  Millimeter-­Wave  Plasma  Imaging  of  Magnetic  Fusion  Plasmas  47  12:05  

      O   -­   7     B.D.  Blackwell   Recent  results  from  the  H-­1  heliac  on  MHD  activity,  fluctuations  and  RF  propagation  48  Lunch    12:25  -­  13:25                               13:25  -­  17:30     

     

  

                      18:30  -­  20:30         Presenter      Social  Program  /  Excursion  

     

  

  

             Banquet  

  

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Friday,  6th  November,  2015    Time  

     

No  

  

Presenter  Title  

9:00  -­      9:15                  Plenary  4  /  Invited  9  session  (B.D.  Blackwell)    9:00  

      PL  -­   4     H.P.  Laqua  Commissioning  and  First  Plasma  Operation  at  the  Wendelstein7-­X  Stellarator  49  9:45  

      I   -­  15     M.  Nunami  Progress  of  Gyrokinetic  Turbulence  Simulation  in  Helical  Plasmas  50  10:15  -­  10:35                10:35  -­  12:25                 Invited  10  /  Oral  5  session  (N.C.  Luhmann,  Jr.)  coffee  break      10:35  

      I   -­  16     C.Z.  Cheng  Kinetic  physics  of  magnetic  reconnection  in  collisionless  plasmas  51  11:05  

      I   -­  17     Y.  Miyoshi  Geospace  Exploration  Mission:  ERG  52  11:35  

      I   -­  18     T.  Minea  Magnetized  plasma  modelling:  common  approach  interesting  for  fusion      and  processing  plasmas  53  12:05  

      O   -­   8     N.  Sato  Self-­Organization  and  Heating  by  Inward  Diffusion  in  Magnetospheric  Plasmas  54  12:25  -­  13:55                13:55  -­  14:55                 Invited  11  session  (J.-­M.  Noterdaeme)  Lunch      13:55  

      I   -­  19     K.  Nishiyama  Development  and  Initial  Cruise  Operation  of  the  Hayabusa2  Ion  Engine  System  55  14:25  

      I   -­  20     M.  Tendler  The  origin  and  the  impact  of  stochasticity  resulting  from  edge  localized  perturbations  56                

  

  

  

  Summary                        T.  Mutoh  Summary  of  ITC-­25      14:55  -­  15:25       15:25  -­  15:40       Page  

                       Closing  

  

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Poster  Session  1  
  

No  

Presenter  Title  

P1  -­   1  T.  Kanzaki  Change   in   Reynolds   stress   induced   by   end-­plate  biasing  in  a  cylindrical  laboratory  plasma  57  P1  -­   2  Y.  Miwa  How   to   Identify   Ion   Temperature   Gradient   Mode   in  Cylindrical  Plasma  58  P1  -­   3  T.  Tarumi  Observation   of   radial   scale   length   of   fluctuations   in  PANTA  59  P1  -­   4  F.  Kin  Reconstruction   of   Cross-­Sectional   Structure   of  Reynolds   Stress   in   End-­Plate   Biasing   Experiment   in  PANTA  

60  P1  -­   5  R.  Minato  Evaluation   of   flow   velocity   with   a   frequency   comb  Doppler  Backscattering  in  PANTA  61  P1  -­   6  H.  Takahashi  Helium   volumetric   recombining   plasma   formation   for  energetic   ion   injection   in   a   radio-­frequency   plasma  device  DT-­ALPHA  62  P1  -­   7  S. Tsubota  Development   of   linear   plasma   device   for   proof   of  principle  of  energetic  ion  injection  63  P1  -­   8  T.  Onda  Observation  of  Plasma  Dynamics  in  TPD-­sheet  IV  64  P1  -­   9  Y.  Kaminou  Experimental  and  Numerical  Studies  of  Hall  Effect  on  Merging   Formation   Process   of   a   Field-­Reversed  Configuration  65  P1  -­   10  T.  Ushiki  Generation   of   Energetic   Electrons   during   Spherical  Tokamak  Merging  in  UTST  66  P1  -­   11  M.C.C.  Lacdan   Characterization   of   an   atmospheric   pressure   plasma  by  a  Gerdien  condenser  67  P1  -­   12  S.  Nakanobo  Supersonic  Gas  Injection  for  High  Density  Operation  and   Associated   Development   of   140   GHz  Interferometer  in  RELAX  RFP  68  P1  -­   13  M.  Nishiura  Evidence  of  increased  ion  temperature  by  slow  wave  heating  in  RT-­1  69  

P1  -­   14  Y.  Kawazura  Anisotropy  in  tail  component  of  Hα  in  RT-­1  70  

P1  -­   15  Y.  Nakatsuka  Studies   on   power   dependence   of   plasma   production  and  ion  heating  by  a  single  helical  antenna  for  a  simple  thruster  71  P1  -­   16  R.  Yanai  Characteristics   of   Magnetic   Fluctuations   during  Magnetic  Reconnection  in  Counter-­helicity  Spheromak  Merging  Experiment  72  P1  -­   17  N.  Pathak  Role  of  Kinetic  Alfvén  Wave  in  Energy  Transportation  in  Magnetopause  73  - 10 -

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P1  -­   18  H.  Kawazome   Numerical  study  of  spectral  line  shapes  in  high-­density  He  plasmas  74  P1  -­   19  J.  Kohagura  Doppler   reflectometer   system   for   measuring   plasma  rotation  velocity  in  GAMMA  10  75  P1  -­   20   C.L.S.  Mahinay   Co-­extraction  of  Low  Energy  Ar+  and  C2H2-­  Ions  for  Space  Charge  Effect  Mitigation  and  Increased  Beam  Intensity  76  P1  -­   21   C.J.M.  Dela  Cruz   Synthesis  of  hydrogenated  amorphous  carbon  nitride  using  DC  graphite  hollow  cathode  discharge  77  P1  -­   22  T.  Fukuyama  Study   on   operating   principle   of   Cockcroft-­Walton  circuit   to   produce   plasmas   using   high-­voltage  discharge  78  P1  -­   23  M.-­J.  Lee  Nonthermal   and   geometric   effects   on   the   dual-­mode  surface  waves  in  a  Lorentzian  dusty  plasma  slab  79  P1  -­   24  S.  Kogoshi  High   density   plasma   sustainment   in   microwave  plasma  source  with  slot  antennas  80  P1  -­   25  J.P.N.  Gonda   Formation   of   Diamond-­like   Carbon   (DLC)   on   Silicon  using  Low  Energy  Ion  Beam  Deposition  81  P1  -­   26  M.  Toida  Parallel   electric   fields   in   nonlinear   magnetosonic  waves  in  a  multi-­ion-­species  plasma  82  P1  -­   27    

   (  Withdrawal  )  83  P1  -­   28  M.  Kondo  Design  study  on  double  layer  type  interface  structure  for  heat  exchanger  of  Fusion  DEMO  blanket  84  P1  -­   29  D.  Kato  Thermal  quench  of  ion-­beam  induced  luminescence  of  Er2O3  85  P1  -­   30  J.  Yagi  Hydrogen  solubility  in  molten  salt  FLiNaK  mixed  with  nano-­Ti  powder  86  P1  -­   31  T.  Tanaka  Fabrication   and   characterization   of   MOD   ceramic  coatings  on  coolant  duct  materials  87  P1  -­   32  S.  Takayama  Investigation   of   the   selective   heating   of   the   metal  powder   with   hydrogen   in   the   molten   salts   using   by  microwave  88  P1  -­   33  D.  Isshiki  Performance   evaluation   of   a   three-­surface-­multi-­
layered  channle  with  a  reinforced  metal  layer  89  P1  -­   34  Y.  Hishinuma  Adhesion   strength   of   Er2O3/buffer   stacked   MOCVD  coating  for  advanced  breeding  blanket  90  P1  -­   35  T.  Goto  Proposal   of   the   maintenance   method   of   breeder  blankets  of  LHD-­type  helical  fusion  reactor  FFHR  91  P1  -­   36  M.  Tanaka  Observations  of  gas  stream  from  fusion  test  device  for  the  design  of  exhaust  detritiation  system  92  - 11 -

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P1  -­   37  Y.  Aoki  Development   of   Soft-­X   Ray   Computed   Tomography  Diagnostic   System      for   MHD   Dynamics   Study   in  RELAX  RFP  93  P1  -­   38  M.  Hamabe  Studies  on  the  effect  of  radio  frequency  field  in  a  cusp-­
type   charge   separation   device   for   direct   energy  conversion  94  P1  -­   39  T.  Katsura  Studies   on   the   axial   position   of   the   decelerator   in  traveling  wave  direct  energy  converter  95  P1  -­   40  T.  Watanabe  Improvement   of   charge   separation   by   increase   of  magnets  in  a  PM-­CuspDEC  simulator  96  P1  -­   41  H.  Noto  Development   of   high   strength   W/V/Au/ODS-­Cu   joint  using  HIP  process  97  P1  -­   42  T.  Yamada  Development   of   Oxide   Dispersion   Strengthened-­
Copper  Using  MA-­HIP  Method  98  P1  -­   43  M.M.  Alam  Real-­Time   Display   System   of   the   Equilibrium  Reconstruction  for  Long  Time  Discharge  in  QUEST  99  P1  -­   44  H.  Braune  W7-­X  ECRH  plant  -­  Control  System  Design  Features  100  P1  -­   45  Min  Thu  San  Experimental   Study   on   W-­Band   Oversized   Surface   101  Wave  Oscillator  Driven  by  Weakly  Relativistic  Electron  Beam  

P1  -­   46  Y.  Annaka  Experimental   Study   on   Slow   Cyclotron   Interaction   102  under  Anomalous  Doppler  Effect  in  W-­band  Backward  Wave  Oscillator  P1  -­   47  S.  Geng  Depth  of  Influence  on  the  Plasma  by  Beam  Extraction   103  in  a  Negative  Hydrogen  Ion  Source  for  NBI  

P1  -­   48  R.  Makino  Power  and  polarization  monitor  developed  for  ECRH   104  and  ECCD  on  LHD  P1  -­   49    P1  -­   50  (  Withdrawal  )  105  E.  Wakai  Absorption  Behavior  of  Nitrogen  in  Nitrogen  Hot  Trap  106  P1  -­   51  E.  Wakai  Materials   Test   Plan   and   Design   Summary   in   PIE   107  Facility  of  IFMIF  P1  -­   52  R.  Ichimiya  Installation   and   commissioning   of   components   for   108  IFMIF/EVEDA  Prototype  Accelerator  P1  -­   53  R.  Nishikiori  Evaluation   of   dark   current   profile   for   prediction   of   109  voltage  holding  capability  on  multi-­aperture  multi-­grid  accelerator  for  ITER  P1  -­   54  T.  Wada  Variation  in  secondary  electron  collection  and  energy   110  recovery  concerning  electrode  angle  in  a  high  energy  ion  penetration  scheme  P1  -­   55  K.  Tsumori  Numerical  Design  on  New  Steering  Grid  of  Negative   111  Ion  Source  for  NBI  - 12 -

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P1  -­   56  T.  Tanaka  Fabrication  and  property  investigation  of  metal  hydride   112  blocks  for  neutron  shielding  in  a  fusion  reactor  P1  -­   57  H.  Muta  Improvement   of   weight   density   and   thermal   113  conductivity   of   hydride   neutron   shielding   material  using  metal  coated  powder  P1  -­   58  H.  Watanabe  Environmental   Effects   on   V-­4Cr-­4Ti   Alloy   during   114  Irradiation  P1  -­   59  Y.  Hamaji  ACT2:  a  high  heat  flux  test  facility  using  electron  beam   115  for  fusion  application  P1  -­   60  J.  Miyazawa  Conceptual   design   of   the   new   liquid   metal   divertor   116  VOLD-­X  P1  -­   61  Y.  Tanaka  Experimental   simulations   for   detached   plasma   with   117  Super-­X  divertor  using  a  liner  divertor  simulator  TPD-­
Sheet  IV  P1  -­   62  K.  Ogawa  Effects   of   gamma-­ray   irradiation   on   electronic   and   118  non-­electronic  equipment  of  the  Large  Helical  Device  P1  -­   63  M.  Kondo  Desorption  of  non-­metal  impurities  from  Pb  and  Pb-­Li   119  alloy  

P1  -­   64  M.  Kondo  Conceptual   design   of   temporally   storage   area   in   hot   120  cell  for  fusion  DEMO  reactor  P1  -­   65  H.  Fu  Effect   of   HIP   temperature   and   cooling   rate   on   the   121  microstructure  and  hardness  of  dissimilar-­metal  joints  between  ODS-­RAFM  steels  and  JLF-­1  steel  P1  -­   66  X.  Ji  Investigation   on   the   Magnetic   Field   Distortion   by   122  Ferromagnetism  of  the  Blanket  for  the  Helical  Fusion  Reactor  P1  -­   67  S.  Imagawa  Feasibility  study  of  react-­and-­wind  method  for  helical   123  coils  wound  from  cable-­in-­conduit  conductors  P1  -­   68  L.  Aparicio  Evaluation   of   Shear   Strength   in   Soldered   and   124  Mechanical   Lap   joints   of   High-­Temperature  Superconducting   tapes   Intended   for   a   Remountable  Magnet  P1  -­   69  N.  Yanagi  NITA   Coil   ―Innovation   for   Enlarging   the   Blanket   125  Space  in  the  Helical  Fusion  Reactor  P1  -­   70  T.  Obana  Summary   of   conductor   and   joint   test   results   of   JT-­
60SA  CS  and  EF  coils  using  NIFS  test  facility  P1  -­   71  M.M.  Islam  Effect   of   Laval   Nozzle   in   the   GAMMA   10   SMBI   127  Experiments  P1  -­   72  A.  Kamitani  Speedup   of   shielding   current   analysis   in   high-­ 128  temperature   superconducting   film:   implementation   of  H-­matrix  method  - 13 -

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P1  -­   73  S.  Koike  Fuelling   effect   by   axial   injection   of   torus   plasma   into   129  ST  plasma  P1  -­   74  R.  Ikeda  Progress   in   development   of   TE31,11   mode   gyrotron   130  for  ITER  EC  system  P1  -­   75  Y.  Sasaki  Characteristics   of   a   Large   diameter   RF   negative   131  hydrogen  ion  source  P1  -­   76  N.  Ikemoto  Temperature   dependence   of   tritium   enrichment   by   132  electrolysis  with  solid  polymer  electrolyte  P1  -­   77  H.  Date  P1  -­   78  K.  Furuichi  Evaluation   of   tritium   sorption   rate   in   soil   packed   bed   134  by  numerical  analysis  P1  -­   79  T.  Sugiyama  Effects  of  gas-­liquid  ratios  on  separative  performance   135  of   dual-­temperature   water   hydrogen   chemical  exchange  column  P1  -­   80  Y.  Morita  Preparation   of   hydrophobic   platinum   catalyst   with   136  porous   silica   supports   for   water   hydrogen   chemical  exchange  P1  -­   81  N.  Akata  A   new   pretreatment   technique   for   environmental   137  tritium  analysis  with  microwave  heating  method  P1  -­   82  N.  Ashikawa  Tritium  decontamination  from  carbon  deposited  layer   138  by  baking  P1  -­   83  T.  Muroga  Microstructural  and  Microchemical  Evolution  of  V-­4Cr-­
4Ti  under  Heavy  Ion  Irradiation  P1  -­   84  H.  Funaba  Transport   properties   of   high-­beta   plasmas   in   the   140  inward  shifted  configurations  on  LHD  P1  -­   85  K.  Nagaoka  Heat   Transport   Characteristics   of   High   Temperature   141  Discharges   in   LHD      -­   Dependence   on  Hydrogen/Helium  Ratio  -­  P1  -­   86  P.  Vincenzi  Analysis  of  LHD  NBI-­plasma  interaction  with  upgraded   142  FIT3D  module  of  TASK3D-­a  analysis  transport  suite  P1  -­   87  H.  Nakano  Design   study   for   high   spatial   resolution   of   beam   143  emission  spectroscopy  viewing  from  6-­O  port  on  large  helical  device  P1  -­   88  B.J.  Peterson   Preliminary  design  of  an  imaging  bolometer  for  W7X  Study  of  tritium  release  behavior  from  soil  particle  Page  

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P1  -­   89  K.  Mukai  Development   of   in-­situ   calibration   system   for   foil   145  detector  of  infrared  imaging  video  bolometer  in  Large  Helical  Device  P1  -­   90  R.  Sano  Improvement  of  3D  radiation  imaging  using  Euclidean   146  distance  from  model  phantom  in  LHD  P1  -­   91  A.  Tani  Design  and  evaluation  of  broad-­band  high  pass  filter   147  using  metal  perforated  plates  with  circular  holes  - 14 -

No  Presenter  P1  -­   92  K.  Nagasaki  Development   of   Electron   Bernstein  Diagnostic  for  Heliotron  J  and  LHD  P1  -­   93  H.  Tsuchiya  Validation  of  the  Digital  Correlation  ECE  measurement   149  technique      using  low  frequency  fluctuation  in  LHD  P1  -­   94  Y.  Goto  Development   of   OXB   mode   conversion   window   150  searching  method  using  inverse  emission  process  in  the  LHD  P1  -­   95  Y.  Shimooka  Magnetic  flux  surface  measurements  on  TOKASTAR-­
2  P1  -­   96  A.  Shimizu  The  development  of  2D  potential  profile  measurement   152  with   heavy   ion   beam   probe   on   the   Large   Helical  Device  P1  -­   97  M.  Irie  X-­Y-­t   Edge   Plasma   Reconstruction   with   Digital   153  Imaging  Technique  P1  -­   98  H.  Zhang  Estimation  of  total  carbon  radiation  at  each  ionization   154  stage  of  C2+  to  C5+  ions  in  Large  Helical  Device  P1  -­   99  T.  Sakito  Measurement   of   tokamak   plasma   with   the   external   155  helical   field   using   a   high-­speed   camera   in  TOKASTAR-­2  P1  -­  100  P1  -­  101  P1  -­  102  Title  

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Emission   148  151  Y.  Nagayama   Observation  of  Electron  Density  Fluctuations  by  Using   156  the   O-­mode   Microwave   Imaging   Reflectometry   (O-­
MIR)  in  LHD  H.  Igami  Examination   of   efficient   electron   Bernstein   wave   157  heating   and   measurement   scenario   with   ray-­tracing  and  full-­wave  calculations  in  LHD  Y.  Yoshimura   Improvement  in  flexibility  of  ECCD  by  use  of  upgraded   158  ECH  antenna  system  on  LHD  P1  -­  103   T.  Ii  Tsujimura   Optimization  of  ECH  injection  polarization  in  real-­time   159  polarization  scan  experiments  P1  -­  104  H.  Takahashi  Ion   heating   by   ECRH   through   the   equipartition   160  process  in  the  LHD  P1  -­  105  S.  Kamio  Experimental   Evaluation   of   Fast-­ion   Velocity   161  Distribution  Produced  by  Neutral  Beam  with  Si-­FNA  in  LHD  P1  -­  106  T.  Tokuzawa  Developments   of   millimeter-­wave   interferometer   for   162  electron  density  measurements  with  beam  extraction  in  the  negative  ion  source  P1  -­  107   H.  Yamaguchi   Simulation  study  of  NBI  beam  ion  distributions  in  the   163  presence  of  magnetic  islands  in  helical  plasmas  P1  -­  108  S.  Okamura  Optimization  of  helical  coil  structure  for  the  improved   164  magnetic  configuration  of  LHD  - 15 -

  
  

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P1  -­  109  T.  Ido  P1  -­  110  Y.  Takemura  Slowing   down   mechanism   of   instability   leading   to   166  minor  collapse  in  LHD  P1  -­  111  X.D.  Du  Stabilization   of   Helically   Trapped   Energetic   Ions   167  driven   Resistive   Interchange   Mode   by   High   Power  Injection  of  Electron  Cyclotron  Resonance  Heating  P1  -­  112  N.  Nakada  Development   of   fast-­neutron   directional   detector   for   168  fusion  neutron  profile  monitor  at  LHD  P1  -­  113  M.  Shoji  Observation   of   termination   process   of   long   pulse   169  plasma   discharges   using   stereoscopic   fast   framing  cameras  in  the  Large  Helical  Device  P1  -­  114  A.  Sanpei  Identification  of  helical  structure  of  visible  light  in  Low-­
A  RFP  with  concave  lens  array  P1  -­  115  T.  Nishitani  Radiation   field   estimation   for   the   diagnostic   and   171  control   components   by   Monte   Carlo   neutronics  calculations  with  LHD  3-­dimensional  modeling  P1  -­  116  S.  Maeta  Integrated   heat   transport   simulation   of   the   helium-­hydrogen  plasma  of  LHD  by  TASK3D  P1  -­  117  J.  Tachibana  Full-­orbit   simulation   of   fast   α-­particle   in   heliac   173  configurations  P1  -­  118  M.  Isobe  Recent   progress   of   fast-­ion   loss   detector   project   in   174  magnetic  confinement  fusion  devices  in  Japan,  Korea  and  China  Spatial   structures   of   energetic-­particle  geodesic  acoustic  mode  in  the  LHD    - 16 -

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P2   -­   1   U.  Jenica  Rozette  Silicon  Nitride  Film  Deposition  on  Glass  Substrates  via   175  Plasma  Sputter-­Type  Negative  Ion  Source  P2   -­   2  S.J.  Park  Anti-­fog   on   PMMA   by   Nitrogen-­Argon   Atmospheric   176  Plasma  Jet  Treatment  P2   -­   3  M.  Wada  Ion   species   separated   atmospheric   pressure   plasma   177  source  P2   -­   4  G.  Tang  Degradation   of   acetic   acid   in   water   using   gas-­liquid   178  plasma      with  SPG  membrane  P2   -­   5  S.M.  Palardonio   Fourier   Transform   Infrared   Spectroscopy   of   Plasma   179  Polymerized  Polyaniline  P2   -­   6  Y.  Okuno  Effect  of  electrode  shapes  and  the  working  gases  on   180  production   amount   of   hydroxyl   radicals   in  atmospheric-­pressure  plasma  jets  P2   -­   7  H.  Nabuchi  Effects   of   a   magnetic   Laval   nozzle   to   the   thrust   181  performance  on  an  MPD  thruster  P2   -­   8  D.  Adachi  Design   of   electron   supplier   in   a   new   concept   of   182  neutron  generator  for  BNCT  P2   -­   9  T.  Matsui  Beam   ion   confinement   in   a   new   concept   of   neutron   183  generator  for  BNCT  P2   -­   10  A.M.  Gouda  Formation   Mechanism   of   Periodic   Nano-­Grating   184  Structure      by  Surface  Plasma  Wave  P2   -­   11  S.  Ikuno  Krylov   Subspace   Method   with   Communication   185  Avoiding  Technique  for  Linear  System  Obtained  from  Electromagnetic  Analysis  P2   -­   12  K.  Terashima  Magnetic   property   analysis   for   spherical   shell   186  magnetic  substance  using  Monte  Carlo  simulation  P2   -­   13  S.  Fujiwara  Dissipative   particle   dynamics   simulation   of   self-­
assembly  in  bolaamphiphilic  solution  P2   -­   14  H.  Matsuura  Improvement   of   divertor   probe   array   for   heat   flux   188  measurement  of      Heliotron  J  divertor  leg  plasma  P2   -­   15  M.S.  Islam  Study   of   Heat   and   Particle   flux   in   the   case   of   Gas   189  Injection  in  the  D-­module  of  GAMMA  10/PDX  P2   -­   16  K.  Ichimura  Study  of  the  axial  plasma  confinement  for  high  particle   190  flux  operations  in  GAMMA  10/PDX  P2   -­   17  H.  Bi  A  study  on  hydrogen  transport  in  liquid  lithium  under   191  steady  state  plasma  bombardment  P2   -­   18  Y.  Hayashi  Characterization   of   detached   recombining   plasma   in   192  Pilot–PSI  - 17 -

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P2   -­   19  S.  Saito  Erosion  of  bubble  formed  tungsten  materials  P2   -­   20  Y.  Xu  Hydrogen  gas-­driven  permeation  through  F82H  steel      194  coated  with  vacuum  plasma  sprayed  tungsten  P2   -­   21  Y.  Torikai  Tritium  accumulation  on  carbon  deposition  layer  and   195  its  decontamination  by  glow  discharge  P2   -­   22  K.  Katayama  Measurement  of  tritium  sorbed  in  tungsten  deposition   196  layer  by  imaging  plate  technique  P2   -­   23  H.  Nakamura  Tungsten-­surfaces-­structure  Sputtering  Yield  for  Noble  Gas  P2   -­   24  G.  Motojima  Color   measurements   of   the   first   wall   toward   the   198  systematic   evaluation   of   the   thickness   distribution   of  deposition  layer  in  LHD  P2   -­   25  T.  Futakuchi  Linear   stability   analysis   of   Alfven   eigenmodes   with   199  arbitrary  velocity  distribution  in  tokamak  plasmas  P2   -­   26  T.  Nicolas  Bifurcation   of   the   interchange   mode   growth   rate   and   200  rotation   frequency   due   to   the   perpendicular   heat  conductivity  in  stellarator  plasmas  P2   -­   27  B.  Huang  Comparison  of  bootstrap  current  calculation  in  helical   201  plasmas   among   different   types   of   approximations   in  drift-­kinetic  equation  P2   -­   28  Y.  Ohno  Chaos  of  nonlinear  ion  acoustic  waves  with  vorticity  P2   -­   29  G.  Purohit  Electron  impact  ionization  cross  sections  of  the  fusion   203  plasma  hydrocarbon  gas:  methane  P2   -­   30  T.  Pianpanit  Kinetic  dynamics  simulation  of  the  detached  plasma  204  P2   -­   31  H.  Sugama  Local  approximation  of  the  drift  kinetic  equation  205  P2   -­   32  Y.  Kawamoto  Modification  of  neutron  emission  spectra  and  its  effect   206  on   determination   of   fuel   ion   ratio   nd/nt   in   beam  injected  DT  plasma  P2   -­   33  H.  Matsuura  A  verification  scenario  of  knock-­on  tail  formation  due   207  to   nuclear   elastic   scattering   using   γ-­ray-­generating  6Li+D  reaction  in  deuterium  plasmas  P2   -­   34  Y.  Fujita  Transmission  Mode  Analysis  in  Polarizer  Miter  Bend  P2   -­   35  S.A.  Khan  Kinetic   full   wave   analysis   of   electron   cyclotron   wave   209  mode  conversion  in  tokamak  plasmas  P2   -­   36  S.  Maeyama  Benchmark   of   electromagnetic   gyrokinetic   codes   in   210  high  performance  fusion  plasma  Dependence  193  of   197  

202  

208  

P2   -­   37   T.-­H.  Watanabe   Moment   extracted   formulation   for   describing   kinetic   211  Alfven  waves  P2   -­   38  D.  Sharma  Equilibrium  reconstruction  of  plasma  discharges  in  the   212  Aditya  tokamak  - 18 -

No  Presenter  Title  

P2   -­   39  M.  Yanagi  Two   dimensional   hybrid   simulation   of   a   translated   213  field-­reversed  configuration  P2   -­   40  K.  Matsuzaki  Influences   of   Hall   effect   on   MHD   behavior   of   two   214  axially  colliding  FRCs  P2   -­   41  R.  Sekiguchi  Flux   supply   by   Neutral   Beam   Injection   into   Field-­
Reversed  Configration  plasma  with  resistive  decay  P2   -­   42  M.  Homma  Estimations  of  Beam-­Beam  Fusion  Reaction  Rates  in   216  the  Deuterium  Experiment  Plasma  on  LHD  P2   -­   43  H.  Wang  Simulations   of   Energetic   Particle   Driven   Geodesic   217  Acoustic  Mode  in  3-­dimensional  LHD  equilibrium  P2   -­   44  S.  Toda  Advances   in   applying   the   gyrokinetic   heat   diffusivity   218  model  to  transport  simulations  for  helical  plasmas  P2   -­   45  R.  Seki  Development   of   TASK3D/WM   module   for   evaluation   219  the  ICRF  wave  propagation  and  absorption  in  LHD  P2   -­   46  S.  Mochizuki  Analysis  of  High  Energy  Electron  Tail  and  Its  Effect  on   220  the  H-­  Production  in  the  Linac4  H-­  Source  P2   -­   47  I.  Goto  Effect   of   Coulomb   Collision   on   the   Negative   Ion   221  Extraction  Mechanism  in  Negative  Ion  Sources  P2   -­   48  Y.  Idomura  Saturation   mechanism   of   ion   temperature   gradient   222  driven  trapped  electron  mode  turbulence  P2   -­   49  T.  Takahashi  Kinetic  equilibrium  of  a  non-­adiabatic  trap  P2   -­   50  M.  Hata  Kinetic   behavior   of   electron   magnetohydrodynamic   224  structures  P2   -­   51  A.  Takayama  Simulation  study  on  helium  in  the  vicinity  of  tungsten   225  grain  boundary  P2   -­   52  K.  Ichiguchi  Three-­dimensional   numerical   analysis   of   shear   flow   226  effects  on  MHD  stability  in  LHD  plasmas  P2   -­   53  R.  Ueda  Pressure   anisotropy   analyses   based   on   MHD   227  equilibrium   calculations   and   magnetic   diagnostics   in  LHD  P2   -­   54  S.  Satake  Benchmark   of   multi-­ion-­species   collision   operator   for   228  delta  f  Monte-­Carlo  neoclassical  simulation  P2   -­   55  S.  Usami  Improvement   of   a   Multi-­Hierarchy   Simulation   Model   229  for  Magnetic  Reconnection  Studies  P2   -­   56  S.  Sugiyama  Neutron   incident   angle   and   energy   distribution   to   230  vacuum   vessel   in   beam-­injected   LHD   deuterium  plasmas  P2   -­   57  T.  Haruki  Particle-­in-­cell   simulations   of   a   kink   instability   in   a   231  pinched  current  filament  P2   -­   58  R.  Tatsumi  Basic  Comparison  of  the  Numerical  Method  to  Solve   232  Fluid  Equations      for  SOL/Divetror  Plasmas  - 19 -

Page  

215  

223  

No  

Presenter  Title  

Page  

P2   -­   59  T.  Ogawa  Multi-­scale  simulation  of  magnetic  reconnection  using   233  particle-­in-­cell   with   magnetohydrodynamics   on  adaptively  refined  mesh  P2   -­   60  N.  Tsujine  Three  dimensional  magnetic  reconnection  of  stressed   234  X-­point  collapse  P2   -­   61  Y.  Nagamine  Simulation   study   on   resistive   instabilities   in   a   small   235  aspect  ratio  reversed  field  pinch  P2   -­   62  T.  Takayama  Large-­scale  simulation  of  contactless  crack  detection   236  in   HTS   films:   application   of   H-­matrix   method   to   fast  matrix-­vector  multiplication  P2   -­   63  A.  Saitoh  Performance   Improvement   of   Extended   Boundary   237  Node   Method   for   Solving   Elliptic   Boundary   Value  Problems  P2   -­   64  Y.  Ohi  Stability   analysis   on   node   arrangement   in   Meshless   238  Time-­Domain  Method  P2   -­   65  H.  Miura  Structure   transition   in   forced   magnetohydrodynamic   239  turbulence  with  uniform  background  magnetic  field  P2   -­   66  M.  Sato  Extension   of   integrated   transport   analysis   suite,   240  TASK3D-­a,   to   be   capable   for   analysis   of   fast   ion  particles  and  plasma  flows  in  LHD  P2   -­   67  Y.  Shimizu  Validation   of   Tungsten   Transport   Model   in   JT-­60U   241  Plasmas  P2   -­   68  H.  Hasegawa   Study   of   Plasma   Coherent   Structure   Transport   242  Dynamics  with  Particle-­in-­Cell  Simulation  P2   -­   69  Y.  Tamura  Intuitive  interface  for  visualizing  numerical  data  using   243  gesture  recognition  P2   -­   70  H.  Ohtani  Applications  of  the  virtual-­reality  technology  to  Plasma   244  Physics  and  Fusion  Plasmas  P2   -­   71  M.  Yoshikawa   Direct   observation   of   electron   heating   by   electron   245  Landau   damping   of   Alfvén   ion   cyclotron   waves   with  Thosmon   scattering   system   in   the   tandem   mirror  GAMMA  10/PDX  P2   -­   72  T.  Yamamoto   New  Proposal  on  Development  of  Machine  Protection   246  Function   in   Conventional   Control   System   Using  Information  Technologies  P2   -­   73  H.  Kiyonaga  Studies   of   AM   He   Atomic   Magnetometer   for   247  Application  in  Fusion  Plasmas  P2   -­   74  R.  Rajpal  Digital   Signal   Processing   on   FPGA   for   Plasma   248  Diagnostics  P2   -­   75  J.-­H.  Lee  Progress   of   KSTAR   Thomson   Scattering   Diagnostic   249  System  in  2015  - 20 -

No  Presenter  P2   -­   76    P2   -­   77  Title  

Page  

(  Withdrawal  )  250  S.  Purohit  Behavior   of   non-­thermal   electrons   during   ECR   pre-­ionization  at  Aditya  tokamak  251  P2   -­   78  J.-­Y.  Kim  Fast  ion  loss  associated  with  RMP  in  KSTAR  252  P2   -­   79  D.  Seo  The  temperature  calibration  for  IR  camera  of  the  IRVB   253  system  including  periscope  system  for  KSTAR  P2   -­   80  M.  Tanaka  The   effect   of   thermal   history   on   microstructure   of   254  Er2O3  coating  layer  prepared  by  MOCVD  process  P2   -­   81  M.  Ishikawa  Development   of   the   In-­vessel   Components   of   ITER   255  Microfission   Chamber   System   for   Fusion   Power  Measurements  in  ITER  P2   -­   82  M.  Ichikawa  Clarification   of   arcing   mechanism   on   filament-­driven   256  high-­current  arc  discharge  in  Cs-­seeded  negative  ion  source  P2   -­   83  H.  Nuga  Fokker-­Planck   Simulation   for   Runaway   Electron   257  Generation  in  ITER  Disruptions  P2   -­   84  M.  Uchida  Effect   of   UHR   location   on   non-­inductive   formation   of   258  overdense   spherical   tokamak   by   electron   Bernstein  wave  heating  and  current  drive  on  LATE  P2   -­   85  H.  Kudoh  Inter-­strand   conductance   measurement   of   ITER   TF   259  coil  under  epoxy  impregnation  P2   -­   86  T.  Maruyama  Preliminary  thermal  analysis  for  ITER  Lower  Port  #02   260  Integration  Engineering  in  JADA  P2   -­   87  O.  Watanabe  Reduction   Rate   of   MHD   potential   energy   by   toroidal   261  plasma   current   drive   at   tokamak   plasma   formation  with  stationary  direct  current  of  central  solenoidal  coil  P2   -­   88  T.  Wakatsuki  Investigation   of   pressure   profile   controllability   during   262  plasma   current   ramp-­up   with   reduced   magnetic   flux  consumption  in  JT-­60SA  P2   -­   89  F.  Wang  The  effects  of  current  density  profiles  on  core  toroidal   263  rotation  in  EAST  plasmas  - 21 -

    - 22 -

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

Tuesday,  3rd  November,  2015  
  
  
  

  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

  

PL-1
Pushing Particles with Waves: Current Drive and α Channeling
N. J. Fisch
Princeton Plasma Physics Laboratory
Princeton University, Princeton, NJ 08543, USA

It can be advantageous to push particles with waves in tokamaks, relying on wave-particle
resonances to accomplish specific goals.
Waves that damp on electrons or ions in toroidal fusion devises can drive currents if the
waves are launched with toroidal asymmetry. Theses currents are called non-inductive, since
they operate in the absence of an electric field with curl. Driving non-inductive toroidal
current enables steady state operation of tokamaks. The lower hybrid wave and the electron
cyclotron wave have been demonstrated to drive significant currents. Non-inductive current
also stabilizes deleterious tearing modes. To the extent that waves carry momentum, they
may also cause toroidal rotation.
Waves can also be used to broker the energy transfer between energetic α particles andthe background plasma. The energetic byproducts of DT fusion reactions, the 3.5 MeV α
particles, tend to slow down on electrons, but that could take up to hundreds of milliseconds
for typical tokamak reactor densties. Before that happens, the energy in these α particles
can destabilize on collisionless timescales toroidal Alfven modes and other waves, in a way
deleterious to energy confinement. However, it has been speculated that this energy might
be instead be channeled instead into useful energy, that heats fuel ions or drives current.
This effect would be catalyzed by waves that diffuse α particles, but in such a way that the
diffusion in energy of the α particles is fundamentally coupled to diffusion in space. If these
diffusion paths in energy-position space point from high energy in the center to low energy
on the periphery, then α particles will be cooled while forced to the periphery, while theamplified wave can heat ions or drive current. This process or paradigm for extracting alphaparticle energy collisionlessly has been called α channeling. Realized to its full potential, the
α-channeling effect would divert most of the α particle power to rf power.
It is likely that it will take more than one wave to realize the full potential of this effect.
Yet how waves might work in concert to realize this potential remains to a large extent an
open question for research. While the α-channeling effect is speculative, the upside potential
for economical fusion is immense. We are now entering an age of advanced fusion devices,where α particles may be produced in abundance. Yet concerns regarding the economic
uncertainties of fusion energy remain. It thus becomes particularly urgent to explore now
the advantageous possibilities in tokamak fusion when rf waves play a dominant role.

- 23 -

I-1
Current status of the LHD project and its prospect for deuterium
experimenta,bM. Osakabe, a,bM.Yokoyama, cK.Nagasaki, a,bR. Sakamoto, a,dK.Y.Watanabe, a,bK.Nagaoka,
a,b
H.Takahashi, eS.Murakami, dT.Fujita, fY.Ohya, a,bS.Sakakibara, a,bM.Isobe, aT.Akiyama,a,gT. Morisaki, and LHD experiment groupab

National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (Grad. Univeristy for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292, Japanc
Kyoto University, Gokasho, Uji 611-0011, Japand
Nagoya University, Furo-cho, Nagoya 464-8603, Japane
Kyoto University, Nishikyou-ku, Kyoto 615-8540, Japanf
Shizuoka Univerisity, Suruga-ku, Shizuoka 422-8529, Japang
Nagoya University, Furo-cho, Nagoya 464-8602, Japan

Achievement of reactor relevant plasma condition in Helical type magnetic configurations
and exploration in related plasma physics and fusion technologies are the aim of Large Helical
Device (LHD) project. Following achievements are major goals of the LHD project;achievements of (1) high temperature plasmas exceeding 10[keV] at electron density of
2x1019[m-3], (2) high density plasmas exceeding 4x1020[m-3] with electron temperature of
1.3[keV], (3) high beta plasmas exceeding 5[%] at 1[T], and (4) long pulse plasma operation
exceeding 1 [hour] with 3[MW] input heating power. Among them, the achievement of hightemperature, especially ion temperature, is the primary important. In the recent experiments on
LHD, we have achieved ion-temperature of 8.1[keV] at 1x1019[m-3] by the optimization of wallconditioning using long pulse discharge by Ion Cyclotron Heating (ICH). The electron
temperature of 10keV at 1.6x1019[m-3] was also achieved by the optimization of Electron
Cyclotron Heating (ECH).
For further improvement in plamsa performance, a project of deuterium experiments areplanned on the Large Helical Device (LHD). The objectives of the experiment are (1) to realize
high-performance plasmas by confinement improvement in helical plasmas, (2) to study theisotope effects on plasma confinement in toroidal plasma devices and (3) to demonstrate theconfinement capability of energetic ions in helical systems. In the presentation, current staus of
LHD and its prospects for Deuteriume experiments will be discussed.

- 24 -

I-2

Non-equilibrium and Extreme State of Plasmas
Kimitaka Itoh1,212National Institute for Fusion Science, Toki, 509-5202, Japan

Research Center for Plasma Turbulence, Kyushu University, Kasuga, 816-8580, Japan

Research of plasmas is basic to the modern science, civilization and industry, and will be
the origin of future progress of mankind.

In particular, the study of theextremely-nonequilibrium plasmas has the impact in advancing the central issue ofmodern physics and has possibility to accelerate the scientific innovation.
Non-equilibrium properties of plasmas are characterized by, e.g., the spatial
inhomogeneity, deviation from equi-partition, selective excitation of fluctuation,
cross-scale interaction of fluctuations, interferences of dynamics with different timescales (violation of scale-separation for time and space), and others.

In order toilluminate the elementary processes, we explain three prototypical examples. One is the
problem of the physics of H-mode, in which the cross-scale interaction plays the essential
role [1]. The second is the cross-ferroic turbulence, in which multiple structure formation
mechanisms in scalar and vector fields take place [2]. The third is the physics of abrupt
onset of large-scale deformation, which will solve the mystery of trigger problems [3,4].
Creating the Future: This is the time to develop the research to unify thephysics progresses that have been established in the past for various types of plasmas.
This work is partly supported by a grant-in-aid for scientific research of JSPS, Japan(15H02155, 23244113) and by the collaboration programs of NIFS and of the RIAM of
Kyushu University.
[1] S.-I. Itoh and K. Itoh: Nucl. Fusion 54 (2014) 114017[2] S.-I. Itoh, et al.: “Cross-Ferroic Plasma Turbulence“, Meeting of Phys. Soc. Jpn.
(September 2015) 17aCN-10[3] S.-I. Itoh, et al.: Plasma Phys. Control. Fusion 40 (1998) 879[4] A. Bhattacharjee: Lecture Notes in Physics 614: Turbulence and Magnetic Fields in
Astrophysics (ed. E. Falgarone and T. Passot, Springer 2003) 351- 25 -

I-3
Plasma assisted enhancement of agricultural yield
Masaharu Shiratani and Kazunori Koga
Graduate School of Information Science and Electrical Engineering, Kyushu University
744 Motooka, Nishi-ku, Fukuoka, 819-0395, JapanThe sharp growth in food demand is mainly the result of world population growth to over 9
billion by 2050 [1]. There are three main categories of improvement methods of the
agricultural productivity and output: irrigation, fertilization, and crop protection. Atmosphericpressure nonthermal plasmas can contribute such improvements in the three categories by
various ways such as sterilization, fertilization, water treatment and purification, soil
treatment, seed treatment, storage improvement, insecticide, pre-harvest treatments,
post-harvest treatments, because they provide radicals, ions, electrons, light, as well as electricfield without appreciable thermal damage to plants, crops, and fruits. Improvement ofagricultural productivity has been tried by gas fumigation, seed treatments using radiations,
UV light, laser, electric field, and magnetic field. It is important to clarify advantages anddisadvantages of plasma treatments over other methods. One important advantage of plasma
treatments is the fact that highly reactive and short lifetime species of high density can beprovided by plasmas. Highly reactive species are more effective than stable chemicals andhave much less residual harmful effects due to their short lifetime.
As one of such approaches using plasmas, we have shown that atmospheric pressure plasma
irradiation to seeds can induce cell proliferation and continuous growth enhancement ofseveral plants for weeks after their germination [2-6]. The main results are summarized asfollows. 1) Air, O2, and NO (10%)+N2 plasma irradiation enhance their growth, whereas N2,
He and Ar plasma irradiation do not. 2) 3 min irradiation of air plasma with 40-90% humidity
leads to more than 2.3 times faster growth for several weeks. 3) Even for a long storage
duration of 30 days, seeds with air plasma irradiation still keep growth enhancement ability.
4) OH and O are key species to growth enhancement based on growth enhancement
experiments together with optical emission spectra of plasmas and radical detection using
ESR. 5) The growth enhancement is attributed not to an increase in each cell volume but to anincrease in the number of cells. 6) The growth enhancement is attributed not to DNA mutation
but to modification of DNA expressions induced by plasma irradiation.
Plasma physics contributes to agricultural applications of plasmas in various ways. We will
give some issues of agricultural applications of plasmas related with plasma physics.
Work supported by JSPS.
[1] The environmental food crisis, UNEP, ISBN: 978-82-7701-054-0.
[2] S. Kitazaki, et al., Proc. IEEE TENCON (2010) p.1960.
[3] N. Hayashi, et al., Jpn. J. Appl. Phys. 50 (2011) 08JF04.
[4] S. Kitazaki, et al., Jpn. J. Appl. Phys. 51 (2012) 01AE01.
[5] S. Kitazaki, et al., Curr. App. Phys. 14 (2014) S149.
[6] T. Sarinont, et al., J. Phys.: Conf. Series 1 (2014) 015078.

- 26 -

I-4
Innovative medical technology in plasma disinfection of human bodywith low-temperature atmospheric-pressure plasmas~ the reduced-pH method and the plasma-treated water (PTW) ~K. Kitano1,4, S. Ikawa2, Y. Nakashima2, A. Tani3, T. Yokoyama1, T. Ohshima4,1
1

Graduate School of Engineering, Osaka University, Suita, 565-0871, Japan,
Technology Research Institute of Osaka Prefecture, Izumi, 594-1157, Japan
3
Graduate School of Science, Osaka University, Toyonaka, 560-0043, Japan
4
School of Dental Medicine, Tsurumi University, Yokohama, 230-8501, Japan2

1/half-life [1/min.]half-life [min.]Considering wet conditions of human tissues, plasma process in liquid is an essential
concept. Various kinds of chemical species in liquid induced by gas plasma can be used forsome desirable medical treatment. For disinfecting human body in dental and surgicalapplications by low-temperature atmospheric-pressure plasmas, the inactivation of bacteria inbody fluid is essential. The reduced-pH method was developed that strong bactericidal activity
can be achieved if the solution is sufficiently acidic [1]. Induced HOO• into solution was
confirmed to key active species, where O2–• in liquid is known to be changed into HOO• , which
has much stronger bactericidal activity, with acid dissociation equilibrium (pKa 4.8). Infectedmodels of human extracted tooth were effectively sterilized only at low pH condition [2].
In addition to direct and remote plasma exposures, the plasma-treated water (PTW) was
confirmed to be effective with the reduced-pH method. Attenuation rate of PTW bactericidal
activity were in accordance with Arrhenius equation both in liquid and solid states (Fig. 1).
From the experimental results of ESR (Electron Spin Resonance) measurement of O2–• in liquid[3] against PTW with spin trapping method, half-lives of PTW were also in accordance with the
Arrhenius equation (Fig.1). Half-lives of PTW activities at deep freezer temperature (-80 °C)
and body temperature (+37 °C) are respectively calculated to 7 centuries and 3.9 seconds fromthe Arrhenius equation. This indicates that PTW can be cryopreserved in freezer and toxicity tohuman body seems to be low. Key species can be preserved in liquid under cool condition tobring high concentration PTW as integrated value. Based on cryopreservation concept,
sequential production apparatus of hightemperature [degree C]
25 15
0
-18
-30
concentration PTW was developed with 1 meter0
1
10glass tube and cooling system. The bactericidal
sterilization
ESR
activity of the PTW is extremely high that undiluted
-1
10
10
-22
PTW corresponded 22 log reduction (i.e. 10 ) of
-2
100
10
spore cell (B. subtilis). This corresponds to 65%
H2O2. PTW is ideal disinfectant solution for safety
-3
1,000
10
and strong disinfection. Recently, many types of-4
10,00010
PTW, containing plasma activated medium, arereported. But the active controls of pH and
3.4
3.6
3.8
4.0
4.2
temperature have been still omitted. We believe that
1000/T [1/K]
more physicochemical study must be required for Fig. 1: Arrhenius plots for bactericidal
further scientific discussions of plasma medicine.
activity and ESR measurement of PTW.
[1] S. Ikawa, K. Kitano, S. Hamaguchi, Plasma Process. Polym., 7, 33 (2010).
[2] E. Usui, T. Oshima, H. Yamazaki, S. Ikawa, K. Kitano, N. Maeda, Y. Momoi, The Japanese
Journal of Conservative Dentistry Vol.58, No.2, 101-108（2015).
[3] A. Tani, Y. Ono, S. Fukui, S. Ikawa, K. Kitano, Appl. Phys. Lett., 100, 254103 (2012).

- 27 -

O-1
Recent status of the PAX and APEX projects toward theformation of electron-positron plasma
U. Hergenhahna, H. Niemannab, N. Paschkowskia, T. Sunn Pedersenab, H. Saitohaf, J. Stanjaa,
E. V. Stensona, M. R. Stonekingac, C. Hugenschmidtd, C. Piochaczd, S.Vohburgerd,
L. Schweikhardb, J. R. Danielsone, and C. M. SurkoeaMax-Planck-Institute for Plasma Physics, Greifswald and Garching, GermanybErnst-Moritz-Arndt-Universität Greifswald, Greifswald, GermanycLawrence University, Appleton, Wisconsin, USAdTechnische Universität München, Garching, GermanyeUniversity of California, San Diego, USAf
The University of Tokyo, Kashiwa, Japan

haruhiko.saitoh@ipp.mpg.de

Electron-positron plasmas are predicted to exhibit unique properties as a result of the equalmasses of the two particle species [1]. Motivated by recent successful confinement of pure
electron non-neutral plasmas in toroidal geometries [2,3], we aim to create the firstmagnetically-confined electron-positron plasma in stellarator and dipole traps [4] operated at
the NEPOMUC facility [5], the world’s strongest positron source (~109 moderated positrons
per second). In order to realize confinement of positrons and electrons as plasma, we havestarted two experimental projects in parallel. One is the PAX (Positron Accumulation
eXperiment), which aims to accumulate on the order of 1011 positrons using a buffer-gas trap
and a multicell type linear trap [6]. On the way of constructing such a trap system, electron
confinement experiments have been conducted with a Penning-Malmberg trap in a high-field (5T) magnet. More than 1 hour of confinement and observation of collective mode for ~109
electrons have been realized. Based on these results, we will design a new accumulator withlarge electrostatic potential wells and rotating wall fields.
The second project is APEX (A Positron Electron eXperiment), focusing on efficient
injection and confinement of positrons and electrons in toroidal geometries. In APEX, westarted positron injection experiments using a prototype dipole field trap generated by apermanent neodymium magnet. In the first experimental campaign at NEPOMUC in 2015, we
achieved efficient (40%) injection of the remoderated 5 eV positron beam into the dipole field.
Injection was realized using static electric fields applied by ExB plates located outside theconfinement region. We also measured the parallel and perpendicular positron beam energies,
which are important parameters for further optimization of the injection efficiency. In the near
future, we will construct a levitated dipole trap which can simultaneously confine positrons andelectrons in closed field lines. Design studies on the trap configurations and the levitation
system are ongoing.
[1] P. Helander, Phys. Rev. Lett. 113 (2014) 135003.
[2] P. W. Brenner and T. S. Pedersen, Phys. Plasmas 19 (2012), 050701.
[3] Z. Yoshida et al., Phys. Rev. Lett. 104 (2010) 235004.
[4] T. Sunn Pedersen et al., New J. Physics 14 (2012) 035010.
[5] C. Hugenschmidt et al., New J. Physics 14 (2012) 055027.
[6] J. R. Danielson, T. R. Weber, C.M. Surko, Phys. Plasmas 13 (2006), 123502.

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O-2

An Application of Fusion Oriented Neutrons for Transmutations ofLong Life Fission Products
Motoyasu Sato, Norimasa Yamamoto, Atsuo IIyoshi
Yoshihiko Hirooka, Takashi Muto,
An application of D-T fusion neutrons is proposed for the transmutation
processing of Long Life Fission Products (LLFP).
Our simulations have revealed that the LLFP with a half lifetimes of million
years can be transmuted to the isotopes or other atoms with shorter life timeof 5~20 years under one year of continuous irradiations of 14.1MeV neurons
at the rate of 1019/s.
The mirror confinement devices are the candidates of the neutron production
system. The ion of high energy tail will serve for more effective productions
than the ordinary fusion reactors using thermally equilibrium ions to
produce the large amount of neutrons of 1019/s.
Tsukuba University reported an estimation for which a neutron of 1x1014/s isprovided at magnetic field strength 2T powered by the 20keVx30A NBI on
Gamma-10. Neutrons of 5x1018/s may be possible by upgrading to 10T and
NBI input to 8 times. Slow wave of ICRH is the other candidate.
Our basic study performs underlying research and development
about the implementation of neutron source characterized by the
monochromatic energy spectrum and uniform distributed radiation attern.
As an alternative to ADS device, that uses a high energy accelerator, ourstudy suggests to new academic views and development to an application offusion technology to solve a nuclear processing problem on a large scale.
This study is proceeded by the support of ImPACT research program by	
 Japanese government, # 2015-PM08-05-01.

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- 30 -

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

Wednesday,  4th  November,  2015  
  
  
  
  

  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

  

PL-2
Spherical Tokamaks and Fusion Energy Development Path*
M. OnoPlasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543, USA

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its
aspect ratio (A = R0/a) reduced to A near 1.5, well below the normal tokamak operating rangeof A equal to 2.5 or greater. As the aspect ratio is reduced, the ideal tokamak beta (radio ofplasma to magnetic pressure) stability limit increases rapidly, approximately as 1/A achieving~ 40% toroidal beta (βT). The plasma current Ip it can sustain for a given edge safety factor
q-95 also increases rapidly with Ip approaching ITF. Because of the above, as well as thenatural plasma elongation which makes its plasma shape appear spherical, the ST
configuration can yield exceptionally high tokamak performance in a compact geometry. Dueto its compactness and high performance, in addition to the longer term goal of an attractive
fusion energy power source, the ST configuration has various near term applications,
including a compact Fusion Nuclear Science Facility (FNSF) with relatively low fusion
neutron power. A compact ST-based FNSF has the promise of achieving the high neutron
fluence needed for reactor component testing with relatively modest tritium consumption. At
the same time, the unique operating regimes of STs can contribute to several important issues
in the physics of burning tokamak plasmas to optimize the performance of ITER. Since thestart of the two mega-ampere class ST facilities in 2000, the National Spherical Torus
Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in the UK,
active ST research has been conducted worldwide. More than sixteen ST research facilities
operating during this period have achieved remarkable advances in all areas of fusion
research, including fundamental fusion energy science as well as technological innovation [1].
These results suggest exciting future prospects for ST research in both the near and longer
term. Recently, a series of upgrades are being implemented in the ST facilities including
NSTX,
MAST,
and
GLOBUS-M (RF) where thetoroidal magnetic fields are
increased from ~ 0.5T to ~ 1T.
The first plasma of NSTX-U isplanned in August 2015. The
talk will summarize the key
physics results from worldwide
ST experiments, and describe
ST community plans to provide
the database for FNSF design
while improving predictive
capabilities for ITER andbeyond.
The initial resultsfrom the newly commissioned
NSTX-U will be also presented.
*This work supported by DoE Contract No. DE-AC02-09CH11466[1] M. Ono and R. Kaita "Recent progress on spherical torus research," Physics ofPlasmas 22, 040501 (2015).

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I-5
Research towards operation with tolerable ELMs on European Tokamaks
H Meyer1
1 CCFE, Culham Science Centre, Abingdon, OX14 3DB, UK

The high confinement scenarios currently foreseen for ITER/DEMO are unlikely to be compat1/2
ible with the heat flux factor, ηELM = EELM /(Aw τELM ), for the divertor target (EELM : energy of
the ELM to the divertor, Aw : wetted area and τELM : ELM duration) or the ELM energy losses
due to the naturally occurring type-I ELMs. Two routes are open for mitigating the effect of
the ELMs: active ELM control, or the development of natural small or no ELM scenarios. The
coordinated research on the European tokamaks AUG, JET, MAST and TCV aims at advancing
the understanding in both areas.
With resonant magnetic perturbations a reduction of ηELM by a factor of 2-4 by has been demonstrated at low pedestal top collisionality ν∗ ≤ 0.4 on AUG and MAST. This reduction is observed
for a magnetic perturbation which is most strongly amplified by the H-mode barrier edge pressure gradient and that is significantly off field alignment. ELM mitigation is accompanied witha strong density pump-out leading to a 15%-30% reduction in stored energy. On MAST thelost density could be replenished by pellets whilst still maintaining the ELM mitigation. At
high collisionality type-I ELMs have been suppressed on AUG and JET with negligible loss inconfinement and are replaced by smaller, more frequent ELMs symptomatic of type II-ELMs.
Increasing the heating to lower the collisionality in these regimes lead to a reappearance ofthe type-I ELMs on both devices as also observed in plasmas with naturally occurring type-II
ELMs.
ELM pacing with vertical kicks has been demonstrated on AUG, TCV and most recently on JET.
The JET results show that a minimum kick amplitude depending on the pedestal temperatureis needed. The data are consistent with an increased edge current caused by the movement.
The trigger potential is only limited by the achievable edge current and the distance from thestability boundary, which allows triggering of an ELM quite close to a preceding ELM.
The pellet pacing efficiency, seen to be greatly diminished by the lack of C in all-metal devices,can be restored by seeding of N to levels that also increase the confinement. The potential of
triggering an ELM after the previous ELM on AUG is scenario dependent and independent ofpellet mass and velocity. This is in contrast to JET results where 20Hz-30Hz pellet pacing isachieved reliably with sufficiently large mini-fuelling pellets in clean plasmas and ELMs can betriggered immediately after the previous ELM as well as shortly after the L-H transition.
N seeding can also lead to smaller type-I ELMs. Various small or no ELM regimes (type-II,
type-III, type-IV, grassy ELMs, QH-mode) have been demonstrated in medium sized tokamaks.
However, the operation with an all-metal wall has largely closed the access window for regimes
relying on low recycling such as the QH-mode. Type-II like ELMs have been accessed close to
the ITER baseline scenario on AUG and by magnetic perturbations at high recycling. In bothcases recycling at the high field side seems to be important. Type-III ELMs are encounteredat high density and high power with 5%-10% confinement loss. This research shows that apartfrom the pedestal region also the conditions in the SOL are important for the ELM characteristics. Our focus is on physics understanding, since an ITER/DEMO like high recycling SOL
together with low pedestal ν∗ can’t be achieved in today’s devices.

- 32 -

I-6
Design status of structural components of helical fusion reactor FFHR-d1
H. Tamura, T. Goto, N. Yanagi, J. Miyazawa, T. Tanaka, S. Masuzaki, R. Sakamoto, A. Sagara,
S. Itoa, H. Hashizumea, and the FFHR Design GroupaNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aoba-ku, Sendai 980-8579, JapanThe FFHR-d1 is a conceptual design of the LHD-type fusion reactor being developed at NIFS.
Several design optimizations for FFHR-d1 have been conducted under a multipath strategy [1].
The base model has major and minor radii of 15.6 m and ~3.9 m, respectively. The
superconducting magnet system consists of one pair of helical coils and two sets of vertical fieldcoils. The magnetic field of the confinement center is 4.7 T. Since the FFHR-d1 is a fusion
power plant, a blanket system and a divertor system is composed in the device. In the LHD-type
configuration, the space between the plasma boundary and the helical coil is limited. Therefore,
the structural design started from a radial build design of components, especially at the inboardof the torus. The support structure has to be rigid enough so that a constant geometric position of
the components. On the other hand, large apertures are required for a maintenance work of
in-vessel components. A shape of the support structure was carefully chosen and the analytic
results showed that the stress level was in the permissible limit.
The divertor heat flux in the FFHR-d1 is anticipated to have a high peak of >10 MW/m2. The
neutron load on the divertor can be reduced by setting it behind a blanket to prevent direct
neutron irradiation of the divertor region [2]. Developing this concept, a novel divertor structure
was introduced in which the coil-support structure was modified to create a region receiving
relatively low amounts of radiation. Using this proposed design would increase the lifetime ofthe divertor components. In addition, the divertor could be accessed from either the upper orlower sides of the device, simplifying maintenance [3]. Another idea to solve the divertor highheat flux together with easy maintenance and low radioactive wastes has been proposed using
liquid metal flow [4]. This new divertor system can remove more than 20 MW/m2 heat flux andmore than 90% of the magnetic field lines directed toward the divertor region can be corrected.
The novel divertor structure also matches this liquid divertor system for ease of maintenance.
The feasibility study of a construction and maintenance for a blanket component with a helicalshape has been conducted [5]. In this study, the blanket is simply sliced by the plane with aconstant toroidal angle. Using large aperture of the support structure, these toroidally-sliced
modules can be extracted by a simple movement with radial and vertical directions.
An innovative winding method was developed by connecting high-temperature
superconducting conductors [6]. Winding feasibility investigations using an additive
manufacturing (3D printing) were conducted and it was found that the conductor which has one
helical pitch length could be inserted into the coil casing without deformation. Furthermore, a
new idea for enlarging the blanket space by installing additional helical coils is also being
considered [7] and to be integrated into the design.
[1] A. Sagara, et al., Fusion Engineering and Design 89 (2014) 2114.
[2] T. Tanaka, et al., Fusion Engineering and Design 89 (2014) 1939.
[3] H. Tamura, et al., Fusion Engineering and Design, in press.
[4] J. Miyazawa, et al., in this conference.
[5] T. Goto, et al., in this conference.
[6] N. Yanagi, et al., Nucl. Fusion 55 (2015) 053021.
[7] N. Yanagi, et al., in this conference.

- 33 -

I-7
Hydrogen recycling in the divertor simulation plasma on GAMMA 10/PDX
M. Sakamoto, A. Terakado, K. Nojiri, K. Okia, N. Ezumi, Y. Nakashima,
K. Ichimura, K. Shimizu, M. Fukumotob, D. Katoc, H. Sakauec, M. Yoshikawa,
J. Kohagura, T. Imai, M. IchimuraaPlasma Research Center, University of Tsukuba, 1-1-1 Tennodai, Tsukuba, Ibaraki 305-8577, Japan
Graduate School of Electrical and Electronic Engineering, Chiba University,1-33 Yayoicho, Inage-ku,
Chiba, 263-8522, Japanb
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japanc
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

In the tandem mirror GAMMA 10/PDX, a new project has been promoted to study theboundary plasma and plasma surface interaction (PSI). The divertor simulation experimental
module (D-module) has been installed in the end region of the mirror device to make best use ofa linear plasma device with plasma confinement. The features of GAMMA 10/PDX for theboundary plasma and PSI studies are the following: (1) high ion temperature of the plasma
exposed to the D-module (i.e. a few hundreds eV), (2) high magnetic field (0.15 ~ 1.5 T), (3)large plasma size (0.1 ~ 0.3 m), (4) low background pressure in the vacuum vessel and (5) highcontrollability of the plasma exposure since plasma heating systems of ECH, ICH and NBI are
equipped. Noted that ion energy of the plasma is distributed, meaning that condition of PSI isequivalent to that of the torus plasma from a viewpoint of hydrogen recycling. We have studiedhydrogen recycling with high temperature target and plasma detachment due to molecular
activated recombination (MAR) using the D-module. Moreover, an electrically excited state ofhydrogen atoms reflected on the tungsten target has been studied.
The D-module consists of a rectangular box (0.5 m square and 0.7 m in length) with an inlet
aperture (0.2 m in diameter) at the front panel and a V-shaped target system inside the box.
Tungsten target plates with the thickness of 0.2 mm are attached on a V-shaped base made of Cu.
The target size is 0.3 m in width and 0.35 m in length. The length between the front edge of thetarget and the inlet of the D-module is about 0.3 m. The open-angle of the V-shaped base canremotely be changed from 15 degrees to 80 degrees. The sheath electric heaters are attached on
the backside of the Cu base to control the target plate temperature (Ttarget) up to 573 K. Besides,
additional hydrogen gas can be supplied near the inlet of the D-module. Measurements with
Langmuir probe, spectroscopy, fast camera, ASDEX gauge and so on have been done for theplasma characterization. Moreover, another tungsten target plate with the diameter of 100 mm has
been used to study the hydrogen recycling instead of the D-module.

When the additional hydrogen gas was supplied into the D-module, the electron density on thetarget (0.175 m from the corner of the target) increased up to 2 x 1017 m-3 at ~0.4 Pa of the internalpressure of the D-module and then it decreased to ~0.6 x 1017 m-3, and the electron temperature
decreased from 34 eV to ~3 eV, suggesting the detachment occurred near the corner of the V-shaped
target. The dependence of the Hα and Hβ line intensities on the neutral pressure suggests that molecular
activated recombination (MAR) occurred near the corner of the V-shaped target. When the Ttarget wasincreased upto 573 K shot by shot in another experiment, it is found that the intensities of balmerlines inside the V-shaped target and the electron density increased with increase in Ttarget, indicating thatthe recycling was enhanced due to increase in the target temperature.

- 34 -

I-8
Interactions of tungsten plasma facing material with edge plasma
Y. Ueda, K. Ibano, H. T. Lee
Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871, Japan

Tungsten is a leading candidate for plasma facing materials for ITER and DEMO.
Becasuse it has various preferable natures such as high thermal conductivity, high melting
point, and low sputtering erosion by fuel ions. However, there still remain several issues suchas evaluation of lifetime and impacts on plasma performance under complicated heat andparticle exposure conditions, which include transient heat (ELMs and disruption) and mixed
plasma (DT, He, edge cooling gas). In addition, fuel particles behavior is still underinvestigation such as fuel particle recycling which affects edge plasma control, T retention
and permeation which affect system design, safety and T economy in fusion reactors. In
reactor conditions, heavy neutron irradiation makes these more complicated since it
deteriorate mechanical properties and increase retention.
In this presentation, recent research results and issues towards ITER and DEMO ontungsten plasma facing materials are reviewed especially on fuel retention and pulse heatloading effects.
On fuel retention, a large database as function of fluence [1] (≤1026 m-2), temperature(300-800 K), microstructure [3] and displacement damage [4] now exists from controlledlaboratory experiments. Therefore, within the above parameter range hydrogen retention
behavior is qualitatively well understood. However, quantitative understanding of the
microscopic nature of the trap sites (e.g. concentration, energies, etc) still requires further
work. Translating the laboratory findings to tritium inventory in fusion machines remains
challenging due to several unique and complicated conditions such as the effect of high flux
irradiations (≤1025 m-2s-1), transient temperature excursions by ELMs, and evolution ofradiation damage under complicated temporal and spatial temperature conditions. The
development of radiation damage resistant advanced W-materials often require processingthat increase retention and such conflicting requirements must be balanced.
On pulsed heat response, the most important issues remained are high cycle pulsed
heat loading effects on tungsten. It is known that unmitigated ELMs would cause frequentsurface melting, which is unacceptable in terms of enhanced erosion and plasma
contamination. But even under pulsed heat less than that of the melting threshold could cause
surface cracking and local melting. Acceptable ELM pulse energy for tungsten is still
unknown, but this information is very important for development of ELM control in DEMO.
Synergistic effects of pulsed heat and particle loadings (especially He) make conditions more
complicated. Comprehensive database and appropriate modeling are necessary. In addition,
melt layer behavior due to unmitigated disruption heat loading also should be known well.
This closely relates to lifetime evaluation and requirement of disruption supression. One of
the key physics is vapor shieling effects but they require more works.
[1]
Z. Tian et al, Journal of Nuclear Materials 399 (2010) 101–107[2]
V.K. Alimov et al., Journal of Nuclear Materials 438 (2013) S959–S962[3]
A. Manhard, Deuterium inventory in tungsten after plasma exposure: amicrostructural survey, Ph.D Thesis, University of Augsburg, 2011
[4]
M.H.J. ‘t Hoen et al., Nuclear Fusion 52 023008 (2012)- 35 -

I-9
Research on Plasma Wall Interaction Facilities for Fusion Reactors
B. Unterberg
Institut für Energie- und Klimaforschung – Plasmaphysik, Forschungszentrum Jülich, D52425 Jülich, Germany
Plasma-wall interactions (PWI) will decisively determine the availability and thus the
economy of a fusion reactor because of their impact on the lifetime of the first wall (erosionand deposition) and on safety (tritium retention and dust production).
In view of plasma-wall interactions in future fusion devices such as ITER and ademonstration reactor, new challenges have to be met: extended operational regimes withrespect to particle and heat flux densities onto plasma facing components, both steady-state
and transient, the use of toxic first wall materials (Be in ITER), the presence of Tritium, theimpact of neutron irradiation onto first wall materials and synergistic effects of the above
mentioned operational conditions in fusion reactors.
Important PWI issues are related to short time scales of physical processes and can beinvestigated in devices with short pulse length such as power exhaust, erosion behavior andthe subsequent transport of eroded wall material into the plasma edge. However, major PWI
issues become significant on long time scales because they are related to long fluencies: The
build-up of deposited layers and their stability, the evolution of surface morphology of plasma
facing components, the accumulation of dust, the build-up of fuel inventory in plasma facing
components, fatigue effects associated with a large number of transients and – last but notlease – the accumulation of neutron damage with its impact on PWI processes.
To characterize plasma surface interactions under these conditions, dedicated plasma surface
interaction facilities such as linear plasma devices can be used, which allow for detailed
investigations not possible in magnetic confinement devices. Such experiments arecomplementary to studies in toroidal confinement devices.
In this contribution, we will exemplarily show important contributions from linear plasma
devices for the understanding of PWI processes and introduce the JUelich Linear plasma
Experiment (JULE-PSI), the construction of which is currently starting and which will becapable to investigate PWI processes with neutron irradiated material samples inside a Hot
Cell.

- 36 -

O-3
Velocity profile measurement of lead-lithium duct flowby ultrasonic Doppler velocimetry
Y. Ueki, J. Yagia, Y. Noguchi, T. Tanakaa, T. Yokomineb, M. Hirabayashic,
K. Arac, T, Kunugib, A. Sagaraa
Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871, Japan
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
Kyoto University, Kyoto-Daigaku Katsura, Nishikyo-ku, Kyoto, 615-8540, Japanc
Japan Atomic Energy Agency, 4002 Narita, Oarai, Ibaraki, 311-1393, Japanaueki@mech.eng.osaka-u.ac.jp

The lead-lithium eutectic alloy is a promising liquid breeder. Lead-lithium flows need to bequantitatively evaluated, especially in the presence of a strong magnetic field, for better
understanding its thermofluid in fusion blankets and sophisticating blanket designs. However,any velocity profile measurement techniques were not proposed before our research anddevelopment demonstrated that ultrasonic Doppler velocimetry (UDV) is able to measure thelead-lithium flow under a certain condition that an ambient oxygen concentration is enough low[1-2]. Sagara et al. of NIFS have constructed a new lead-lithium loop named as the Oroshhi-2loop (See Fig. 1 [3]). In the present study, it is going to be reported that a velocity profile of alead-lithium flow in the Oroshhi-2 loop is measured by means of UDV. A circular pipe is
employed as our test-section (See Fig. 2), which is mounted on the Oroshhi-2 loop. The
velocity profile is measured at the point where the flow is fully developed, and thus themeasurement result is evaluated in comparison with a theoretical predication. As well as thevelocity profile, pressures of the lead-lithium are measured with high-temperature pressure
sensors. Based on them, UDV measurement characteristics and accuracy are quantitatively
evaluated.

Lead-Lithium
Flow

Fig. 1. Photograph of Oroshhi-2 loop [3].

Fig. 2. Schematic drawing of test-section.

[1] Y. Ueki, et al., Fusion Engineering and Design, 89 (2014) 77-81.
[2] Y. Ueki, et al., Fusion Sci. Technol. 60 (2011) 506-LMX510.
[3] A. Sagara, et al., presented at 21st TOFE, Nov. 10th, 2014 Anaheim, USA.

- 37 -

O-4
Impacts of Material Innovation on Advanced Blanket Design
A. Kimura, H.I. Je, Y.S. Ha, D. Chen, W. Han,
K. Yabuuchi, T. Okudaa, S. Ohnukib, S. Ukaib, S. Ohtsukac,
P. Doud, S.H. Nohe, A. Hasegawaf
Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto, Japana
KOBELCO Research Institute, Kobe, Japanb
Hokkaido University, Sapporo, Japanc
Japan Atomic Energy Agency, Oarai, Ibaraki, JapandChonghong University Chonghong, ChinaeKorea Atomic Energy Research Institute, Daejon, Korea
fDept. of Engineering, Tohoku University, Sendai, Japankimura@iae.kyoto-u.ac.jp
Materials development is essential for realization of fusion DEMO reactor and beyond. Highperformance materials R&D has been conducted for the last several decades, and there havebeen made some remarkable technology innovations of fusion blanket structural material
including first wall material. In this work, current status of blanket structural materials R&D
is summarized and the impacts of some material innovations on blanket design are introduced
for each blanket system.
Among the candidate blanket structural materials, oxide dispersion strengthened (ODS) steels
are considered to be promising for advanced blanket systems with high thermal efficiency,
because the ODS steels have high-strength at elevated temperatures and good resistance tocorrosion and irradiation degradation. The operation temperature can be elevated up to 700°C.
Corrosion resistance was considerably improved by a dispersion of nano-sized oxide particles,
which also resulted in suppression of irradiation-induced loss of tensile elongation, namely
irradiation embrittlement.
In comparison to 9Cr martensitic steels, both the advantages and critical issues exist for the
ODS steels, which will be discussed for each blanket system based on the surveillance test
results on corrosion, long-term aging and irradiation experiments of the steels.
Finally, the R&D issues of structural materials towards advanced DEMO blankets arepresented.

- 38 -

I-10
ITC-25 abstract
Progress of Linear IFMIF Prototype Accelerator in Rokkasho
A. Kasugai , J-M Ayala , B. Bolzon , P. Cara , N. Chauvin, S. Chel , E. Fagotti , D. Gex , R. Gobin ,F. Harrault , R. Heidinger , R. Ichimiya , J. Knaster , K. Kondo , A. Marqueta , J. Molla , Ohira ,Y. Okumura , M. Perez , G. Phillips , A. Pisent , G. Pruneri , K. Sakamoto , F. Scantamburlo , F. Senee ,K. Shinto , M. Sugimoto , H. Takahashi and M. Valette
A)

B)

D)

B)

D)

C)

C)

D)

A)

B)

C)

A)

B)

E)

E)

A)

B)

A)

B)

A)

A)

C)

D)

F)

B)

A)

D)

D)

A) Japan Atomic Energy Agency (JAEA), Rokkasho Fusion Institute, Rokkasho, Aomori, JapanB) IFMIF/EVEDA Project Team, Rokkasho, Aomori, JapanC) F4E, Fusion for Energy, BFD Department, Garching, GermanyD) Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA/Saclay), FranceE) Istituto Nazionale di Fisica Nucleare, Laboratori Nazionali di Legnaro (INFN-LNL), Italy
F) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Spain

Under the framework of Broader Approach (BA) agreement between Japan and EU in the field of fusion energy
research, the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International
Fusion Materials Irradiation Facility (IFMIF), in force since 2007, is validating the key technologies for IFMIF
construction on schedule and cost. Whereas the Engineering Design Activities were successfully accomplishedwithin the allocated time, as well as most of the validation activities related with the Target and Test facilities,
the Accelerator Facility is being validated with the Linear IFMIF Prototype Accelerator (LIPAc). LIPAc isdemonstrating the feasibility of the low energy section of an IFMIF deuteron accelerator up to 9 MeV with abeam current of 125 mA in CW at International Fusion Research Energy Center (IFERC) of Japan Atomic
Energy Agency (JAEA) in Rokkasho, Japan. The deuterium ion beam injector developed in CEA Saclay wasdelivered in Rokkasho in 2014, and is under commissioning. At this stage, a 100keV/120mA/CW hydrogen ion
beam and 100keV/90mA/CW deuteron ion beam have been produced with a low beam emittance of 0.21π
mm·mrad (rms, normalized) and mass fractions above 90%. Delivery of RFQ components, consisting of RFQ
accelerator developed by INFN in Italy and RF power supplies developed by CIEMAT in Spain is under way,followed by the installation of RF power supplies scheduled for this year as well. After the commissioning of theinjector, RFQ, MEBT, D-Plate, HEBT, beam dump and SRF linac will be installed and tested in step-by-step
approach.

- 39 -

O-5
Discussion on the effects of NaF on physical properties of Flinabe
in comparison to Flibe from molecular dynamics simulations
H. Shishido, N. Yusa, H. Hashizume, Y. Ishiia, N. Ohtorib
Department of Quantum Science and Energy Engineering, Graduate School of Engineering,Tohoku University, Sendai, Japana
Graduate School of Science and Technology, Niigata University, Niigata, Japanb
Department of Chemistry, Faculty of Science, Niigata University, Niigata, Japan

hshishi@karma.qse.tohoku.ac.jp

Molten salt Flibe, which is a binary mixture of LiF and BeF2, has been proposed as a candidate
for a self-cooled liquid fusion blanket system [1], because of its such excellent characteristics as
chemical stability, low vapor pressure, and much lower magnetohydrodynamic effects thanthose of liquid metals. Meanwhile, Flibe suffers from inefficiency in removing heat from thefirst wall due to the fact that it has high melting point and high Prandtl number. Another salt
Flinabe, a ternary mixture of LiF, BeF2 and NaF, has attracted attention recently [2] because it
has lower melting point than Flibe. Whereas physical properties are necessary to discuss theapplicability of Flinabe to a blanket system, experiment at high temperature involves technicaldifficulty because of high corrosiveness, and toxicity of beryllium. Furthermore, Flinabe is aternary mixture consisting of three compounds, which implies that finding a suitablecomposition of Flinabe experimentally would require many resources.
A polarizable ion model has been developed to simulate the physicochemical characteristics ofmolten salts by molecular dynamics (MD) simulations [3]. Several studies have parameterized
the interaction potentials for many molten salts based on first-principles simulations andindicated that molecular dynamics simulations with the polarizable ion model are quite usefulto predict the physical properties of molten salts. Encouraged by these studies, our previous
study [4] simulated density and viscosity of Flibe and Flinabe with the pair-potential
parameterized by ref [5]. Results revealed that Flinabe shows lower viscosity than that of Flibeeven when they contain almost the same BeF2. However, the maximum discrepancy between
the calculated viscosities and the experimental data was about 18 %. These results led us toconsider that it is necessary to improve the potential parameters in order to predict the properties
accurately.
This study aims to reconstruct the pair-potential parameters of Flibe and Flinabe, and evaluate
the effects of NaF on the physical properties. Firstly, the force and the dipole on each ion are
analyzed by first-principles simulations based on density functional theory. Then, theinteraction potentials are fitted to results from first-principles calculation. This study has
obtained the parameters fitted well, rather than the earlier study [5]. Then, the physicalproperties (density, viscosity, and thermal conductivity) are evaluated with the potential
parameters obtained above and compared with the experimental data. Finally, the influence of
NaF on the properties of the salts will be discussed.
[1] A. Sagara et al., Fusion Sci. Technol. 47 (2005) 524.
[2] A. Sagara et al., Fusion Eng. Des. 89 (2014) 2114.
[3] P. A. Madden and M. Wilson, Chem. Soc. Rev. 25 (1996) 339.
[4] H. Shishido et al., Fusion Sci. Technol., in press.
[5] R. J. Heaton et al., J. Phys. Chem. B 110 (2006) 11454.

- 40 -

O-6
Experimental and numerical study of H- production in DC H- source of
medical cyclotron
H. Etoha, M. Onaib, Y. Aokia, H. Mitsuboria, Y. Arakawaa, J. Sakurabaa, T. Katoa, T.
Mitsumotoa, S. Yajimaa, T. Shibatac, A. Hatayamab, and Y. Okumurada

Sumitomo Heavy Industries, Ltd., Tokyo, Japan
Graduate School of Science and Technology, Keio University, Kanagawa, Japanc
High Energy Accelerator Research Organization (KEK), Ibaraki, Japand
Fusion Research and Development Directorate, Japan Atomic Energy Agency, Aomori, Japanb

Negative hydrogen ion (H-) beam is widely used for medical proton cyclotrons whose
applications are cancer therapy and diagnostics, such as boron neutron capture therapy (BNCT)
and the radioisotope production for PET and SPECT. These applications require high intensity
beam extracted from cyclotron to shorten the irradiation time for BNCT and increase theproduction rate of medical radioisotope.
A high current DC H- source has been developed for improving the cyclotron’s beam intensity[1, 2]. The source is a multi-cusp DC arc-discharge source and H- ions are produced by thevolume production mechanism. The addition of Caesium (Cs) into this type of source is known
to increase the production rate of H- on the surface of the plasma electrode. Cs-free andCs-seeded operations have been tested on our source. 10 mA of H- beam current has beenobtained stably in the Cs-free operation, and H- current has reached 22 mA in the Cs-seeded
operation.
In addition to these encouraging results, the relationship between the H- production and thedesign/operating parameters (such as magnetic field configuration, arc-discharge power, and soon) is now being studied experimentally, in order to enhance the H- current. Moreover, asystematic numerical study of electron energy distribution function (EEDF) in the source
plasma is also now underway by KEIO-MARC simulation code [3]. The resultant Hproduction in the source is analyzed by the zero dimensional (0D) rate equation for H- ion.
Initial results show a reasonable agreement with the experimental results. For example, thesimulation by KEIO-MARC code reproduces the magnetic filter effects, i.e., the large
reduction of the electron temperature across the magnetic filter field. Figure 1 shows the resultsof EEDF at each distance from the plasma electrode. For further enhancement of the H- beam
current, more detailed comparison between the experiment and the modeling is now underway.

Figure 1. EEDF at each distance from the plasma electrode[1] H. Etoh et al., Rev. Sci. Instrum. 85, 02B107 (2014).
[2] H. Etoh et al., AIP Conference Proceedings, 1655, 030014 (2015).
[3] T. Shibata et al., J. Appl. Phys. 114, 143301 (2013).

- 41 -

- 42 -

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

Thursday,  5th  November,  2015  
  
  
  
  

  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

  

PL-3
Innovations	
  that	  make	  the	
  Use	  of	
  Ion	  Cyclotron	  Power	
  suitable	
  for	
  Fusion	
  Reactors	  	  J.-­‐M.	
  Noterdaeme1,2,	
  V.	
  V.	
  Bobkov1,	
  H.	
  Fuenfgelder1,	
  R.	
  D'Inca1,	
  R.	
  Ochoukov1,	
  J.	
  Jacquot1,	
  	
  
I.	
  Stepanov2,	
  W.	
  Zhang1,	
  H.	
  Faugel1,	
  K.	
  Crombé2,3,	
  R.	
  Ragona2,3,	
  A.	
  Messiaen3,	
  	
  
D.	
  Van	
  Eester3,	
  A.	
  A.	
  Tuccillo4,	
  O.	
  Tudisco4,	
  G.	
  Rocchi4,	
  Q.	
  Yang5,	
  D.	
  Milanesio6,	
  	
  
R.	
  Maggiora6,	
  L.	
  Colas7,	
  S.	
  Heuraux8,	
  E.	
  Faudot8,	
  J.	
  Moritz8,	
  A.	
  Silva9,	
  	
  
the	
  ASDEX	  Upgrade1	
  and	
  EUROfusion	
  MST1*	
  Teams	
  
1

	
  

Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748 Garching, Germany
2
Ghent University, Applied Physics Department, B-9000 Gent, Belgium
3
LPP-ERM/KMS, B-1000 Brussels, Belgium
4
ENEA, Frascati, Italy
5
ASIPP, Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China
6
Politecnico de Torino, Torino, Italy
7
CEA, IRFM, F-13108 Saint-Paul-Lez-Durance, France
8
IJL UMR 7198 CNRS-Université de Lorraine, BP70239, F-54506 Vandoeuvre, France
9
Instituto de Plasmas e Fusão Nuclear, IST, 1049-001 Lisboa, PortugalThe	
   coupling	
   of	   power	
   in	
   the	   Ion	
   Cyclotron	   Range	
   of	
   Frequency,	
   to	
   the	   plasma,	
   has	   been	
  
one	
  of	
  the	
  first	  heating	
  methods	
  used	  in	
  fusion	  research,	
  and	
  is,	  in	
  principle,	
  also	  one	
  of	
  
the	
   most	
   promising.	
   Over	   the	   years,	
   it	
   has	   encountered	
   challenges,	
   which	
   have	
   put	   its	  future	  use	
  into	  question.	
  These	
  can	  be	
  classified	
  into	  two	
  broad	  range	
  of	
  categories:	
  the	  sensitivity	
  to	
  the	
  plasma	
  edge	  density	
  profile	
   with	
   the	   corresponding	
   difficulty	
   to	
   couple	  the	
  power	
  to	
  the	
  plasma	
  and	
  enhanced	
  plasma-­‐antenna	
  interaction	
  with,	  among	
  others,	  the	
  resulting	
  impurity	
  production.	
  	
  Following	
   a	
   short	
   historical	
   overview,	
   we	
   concentrate	
   on	
   the	   systematic	
   approach	
   taken	
  and	
  recent	
  results	
  obtained	
  at	
  the	  Max-­‐Planck-­‐Institut	
  für	  Plasmaphysik	
  in	
  cooperation	
  with	
   other	
   institutes,	
   to	
   tackle	
   step	
   by	
   step	
   those	
   challenges,	
   leading	   to	
   solutions	
  applicable	
  to	
  fusion	
  reactors.	
  In	
   the	   category	
   sensitivity	
   to	
   the	   plasma	
   edge	
   density	
   profile,	
   the	   first	
   serious	
   problem	
  encountered	
  followed	
  the	
  appearance	
  of	
  the	
  H-­‐mode	
  and	  the	  resulting	
  strong	  ELMs.	
  The	
  
use	
  of	
  3-­‐dB	
  couplers	
  isolated	
  the	
  generators	
  from	  the	
  antenna.	
  More	  recently,	
  tailoring	  the	
   density	
   profiles	
   with	   local	
   gas	   puffing	
   was	   shown	
   experimentally	
   to	
   increase	
   the	  coupling.	
   A	
   systematic	
   study	
   confirmed	
   that	   the	   coupling	
   can	   be	
   properly	
   calculated	
   if	  the	
  local	
  density	
  profile	
  is	
  known.	
  	
  The	
  combination	
  of	
  a	
  theoretical	
  approach	
  to	
  model	  the	
   local	
   density	
   profile	   and	   local	
   measurements	
   at	
   the	   antenna,	
   which	
   will	
   soon	   be	  available,	
   is	
   expected	
   to	
   develop	
   into	
   a	
   predictive	
   capability	
   to	
   calculate	
   and	   optimize	  the	
   coupling.	
   We	
   will	
   be	
   able	   to	
   rely	
   on	
   those	
   tools	
   for	   the	   design	
   of	   a	   new	
   antenna	  concept	
  for	
  DEMO	  (360°	
  toroidally	
  distributed,	
  travelling	
  wave	  type	  antenna),	
  which	
  is	
  
by	
  itself	
  already	
  less	  sensitive	
  to	
  the	
  edge	  density	
  profile	
  and	
  has	  a	
  better	
  coupling.	
  	
  
In	
   the	   category	
   enhanced	
   plasma-­‐antenna	
   interaction,	
   a	
   persistent	
   enhanced	
   impurity	
  production	
  was	
  exacerbated	
  in	
  the	  all-­‐metal	
  ASDEX	  Upgrade	
  when	
  the	
  antenna	
  limiters	
  were	
   also	
   coated	
   with	
   W.	
   The	   hypothesis	
   that	   this	
   enhanced	
   interaction	
   is	   due	   to	
   RF	
  sheaths	
  –	
  possibly	
  a	
  consequence	
  of	
  undesirable	
  induced	
  current	
   -­‐	
  has,	
  up	
  to	
  now,	
  not	  been	
  invalidated.	
  New	
  3-­‐strap	
  antennas	
  in	
  ASDEX	  Upgrade,	
  designed	
  to	
  lower	  these	  RF	
  sheaths	
   by	
   cancelling	
   the	   undesirable	
   induced	
   currents,	
   have	
   shown	
   experimentally	
   to	  indeed	   reduce	
   the	   impurity	
   production.	
   Theoretical	
   approaches	
   to	
   model	   in	
   detail	
   the	  development	
   of	
   the	   sheaths	
   are	   being	
   developed	
   and	   will	
   be	
   checked	
   against	  measurements	
  on	
  ASDEX	  Upgrade	  and	  on	
  IShTAR,	
  a	
  dedicated	
  test	  stand.	
  Here	  too,	
  the	
  combination	
  of	
  theory	
  and	
  experiments	
  is	
  expected	
  to	
  lead	  to	
  a	
  better	  understanding	
  of	
  
the	
   sheath	
   production	
   mechanism,	
   and	   to	   a	   solidly	
   grounded	
   approach	
   to	
   design	
  antennas	
  with	  reduced	
  sheaths.	
  The	  new	
  antenna	
  concept	
  for	
  DEMO	  conceptually	
  avoids,	
  with	
   its	   360°	
   symmetry,	
   undesired	
   induced	
   currents	
   and	   possibly	
   prevents	
   the	  occurrence	
  of	
  sheaths.	
  The	
  optimization	
  of	
  the	
  DEMO	
  antenna	
  concept	
  will	  benefit	
  from	  a	
  better	  understanding	
  of	
  sheath	
  production	  and	  the	  availability	
  of	
  benchmarked	
  tools.

	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  	
  
*	
  http://www.euro-­‐fusionscipub.org/mst1	
  - 43 -

I-11
Predictions of Plasma Performance and Fusion Reactionsin the Deuterium Experiment Plasma of LHD
S. Murakami , H. Yamaguchi, M. Homma, S. Maeta, A. Sakai, A. Fukuyama, K. Nagaoka,
H. Takahashi, H. Nakano, M. Osakabe, M. Yokoyama, K. Tanaka, K. Ida, M. Yoshinuma,
M. Isobe, H. Tomita, K. Ogawa, LHD Experiment Group
Department of Nuclear Engineering, Kyoto Univ., Nishikyo, Kyoto, JAPANa
National Institute for Fusion Science, Toki, JAPANb
Department of Quantum Engineering, Nagoya Univ., Nagoya, JAPANThe deuterium experiment project from 2017 is planned in LHD, where the deuterium NBI
heating beams with the power more than 30MW is injected into the deuterium plasma. Main
objects of this project are to make clear the isotope eﬀect on the heat and particle transport in the helical plasma and to study energetic particle confinement in a helical magnetic
configuration measuring triton burn-up neutrons.
We have developed an integrated transport simulation code, TASK3D[1], and applied tothe high ion temperature plasma of LHD[2, 3]. The particle and heat transport equations
are solved and compared with LHD experimental results. Heat and particle transports are
assumed to be the sum of the neoclassical transport and the turbulent transport. We obtain
relatively good agreements between the simulated and experimental radial profiles of thedensity and temperature in the steady state plasma of LHD[2, 3]. Also, to study the kinetic
behavior of the energetic particle in a helical magnetic configuration, we have developed
GNET[4] code, in which the drift kinetic equation of energetic particles is solved in fivedimensional phase space. GNET code is improved to evaluate the D-D and D-T fusion
reaction between the energetic and thermal ions.
In this paper, we study the deuterium experiment plasma of LHD by TASK3D and GNET.
We solve the heat and particle transport of the deuterium plasma assuming the improvementof the turbulent transport from that of the hydrogen plasma due to the isotope eﬀect. We
consider a simple and H-He experiment based turbulent transport models. The radial profile
of densities and temperatures of the deuterium experiment plasma are evaluated. Next
we study the production and confinement of Tritons (1.01MeV) by GNET code. We show
characteristic distributions of Tritons in the real and velocity spaces. We also evaluate D-T
fusion reaction rates and Triton burn-up neutron signals of the diagnostic system using theobtained triton distributions.
References[1] M. Sato et al., Plasma Fusion Res. 3, S1063 (2008).
[2] S. Murakami et al., Plasma Phys. Control. Fusion 57, 054009 (2015).
[3] A. Sakai et al., Plasma Fusion Res. 10, 3403048 (2015).
[4] S. Murakami et al., NUCLEAR FUSION 46, S425 (2006).

- 44 -

I-12
The full-size source and injector prototypes for ITER neutral beams
G. Serianni, P. Agostinetti, V. Antoni, D. Aprile, C. Baltador, M. Cavenagoa, G. Chitarin, N.
Marconato, D. Marcuzzi, E. Sartori, P. Sonato, V. Toigo, P. Veltri, P. Zaccaria
Consorzio RFX (CNR, ENEA, INFN, UNIPD, Acciaierie Venete SpA), Corso Stati Uniti 4 – 35127
Padova, Italya
INFN-LNL, viale dell’Università n. 2, 35020 Legnaro, Italy

Two Neutral Beam Injectors (NBI) will provide a substantial fraction of the heating power
necessary to ignite thermonuclear fusion reactions in ITER. The development of the NBI systemat unprecedented parameters (40 A of negative ion current accelerated up to 1 MV) requires astrong demonstration activity, which was endorsed by ITER to optimise the crucial components
and systems. A test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), ispresently in the final phase of construction at Consorzio RFX (Padova, Italy) in the CNR
research area and will house two experiments, named SPIDER and MITICA. A full-size
negative ion source, SPIDER (Source for the Production of Ions of Deuterium Extracted fromRf plasma), will be operated in the facility to demonstrate the creation and extraction of a D-/Hcurrent up to 50/60A on a wide surface (more than 1m2) with uniformity within 10%. The
second experimental device is the prototype of the whole ITER injector, MITICA (Megavolt
ITer Injector and Concept Advancement), aiming to develop the knowledge and thetechnologies to guarantee the successful operation of the two injectors to be installed in ITER,
including the capability of 1MV voltage holding at low pressure. The beam source is the key
component of the system, whose design results from a tradeoff between requirements of the
optics and real grids with finite thickness and thermo-mechanical constraints due to the coolingneeds and the presence of permanent magnets. The experimental effort is supplemented bynumerical simulations devoted to the optimisation of the accelerator optics and to the estimationof heat loads and currents on the various surfaces.
In the paper the main requirements will be discussed and the design and the status of the main
components and systems will be described. Particularly a review of the accelerator physics anda comparison between the designs of the SPIDER and MITICA accelerators are presented.
This work was set up in collaboration and financial support of F4E.

- 45 -

I-13
Recent Studies of Hydrogen Negative Ion Source and Beam
Production for NBI in Large Helical Device
K. Ikedaa , P. Agostinettic , M. Brombinc , U. Fantzd , S. Gengb , O. Kanekoa , M. Kisakia ,K. Nagaokaa , H. Nakanoa , M. Osakabea , E. Sartoric ,G. Seriannic , Y. Takeiria ,K. Tsumoria , P. Veltric , C. Wimmerd , and D. Wünderlichda National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb The Graduate University for Advanced Studies, Toki Gifu, 509-5292, Japanc Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA) Corso Stati

Uniti 4 - 35127 Padova (Italy)d Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching, Germany

Neutral beam injector (NBI) is an indispensable device to produce high-density and hot
plasmas in fusion development machine. We have developed negative-ion-based NBI (NNBI) in LHD, that have been operated continuously since 1998. Three N-NBIs are arranged
tangentially, and was equipped with a negative hydrogen ion (H− ) source which enabled highenergy and high power beam injection. The nominal specification of beam energy and beam
power for each N-NBI was 180 keV and 5 MW, respectively. We had achieved 16 MW beam
injection by three N-NBIs in LHD experiment. Moreover, we installed two positive-ion-based
neutral beam injector (P-NBI) arranged perpendicularly to the LHD. High-power (12 MW)with lower energy (40 ∼ 60 keV) beam injection allowed eﬀective ion heating and profile
measurement of ion temperature by charge exchange recombination spectroscopy.
A hydrogen negative ion source is a key component in N-NBI system, that will also use
future NBI in ITER. Understanding of H− behavior from cesiated metal production surface
to extraction aperture is an important issue to maintain high power negative ion beam.
We have researched distribution of H− ions and its extraction behavior by cavity ring-down
spectroscopy and a spectrally-selective imaging system, respectively, in the 1/3 scaled ion
source (NIFS-RNIS). We found that H− ions widely distributed in the extraction region, andthese ions constituted H− beam.
Recently, we have a lot of domestic and international collaboration program for H− source.
We applied a spectrally-selective imaging system in RF hydrogen negative on source in ELISE
test facility in Max-Planck-Institut für Plasmaphysik (IPP) Garching (Germany). We foundclear signal response on Hα emission during beam extraction. This observation will contributeto obtain behavior of H− ions in RF source which is variable output for production highpower beam and optimization of beam shape in ITER source. We also observed an extracted
H− beam shape in NIFS-RNIS used the mini-STRIKE system which consists of a carbonfibre-carbon-composite (CFC) tiles and an infrared camera collaborated with Consorzio RFX(Italy). We found the clear dependence between the beam shape and operating parameters.
These findings have contributed to the high performance and safety NBI operation in the
LHD.

- 46 -

I-14
Advances in 2-D Millimeter-Wave Plasma Imaging of Magnetic Fusion
PlasmasN.C. Luhmann, Jr., Yu-Ting Chang, Ming Chen, Xi Chena, C.W. Domier, J. Gu, Fengqi Hu,
Xing Hu, G.J. Kramerb, Meijiao Li, X. Liu, M. Mamidanna, A.-V. Pham, T. Phan, Xiaoxin Ren,
L. Shib, A.G. Spear, B. Tobiasb, Yilun Zhu
University of California at Davis, Davis, CA 95616, USAa
General Atomics, San Diego, CA 92127, USAb
Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543

UC Davis and PPPL are developing advanced millimeter wave technologies for imaging
density and temperature fluctuations in magnetic fusion plasmas. A high resolution 12×8 MIR
system is under development for use on the EAST tokamak in Hefei, China. The diagnostic issimilar to the DIII-D Microwave Imaging Reflectometer (MIR) system operating on the DIII-D
tokamak [1, 2] and will co-exist with the Electron Cyclotron Emission (ECE)-Imaging systemdeveloped by UC Davis and currently operating on EAST [3]. Unique to the EAST MIR system,
which, is the ability to independently tune all 8 illumination frequencies. In addition toproviding unparalleled flexibility in the selection of plasma cutoff surfaces to monitor, thiscapability also allows the frequencies to be continuously varied so as to maintain optimum
focusing in the presence of time-varying density profile changes.
UC Davis also has an active mm-wave technology development program that aims torevolutionize the state-of-the art in transmitter and receiver systems for ECE-Imaging and MIR.
CMOS systems-on-a-chip (SOC) are being pursued for use as a compact, multi-frequency MIR
transmitter. Next generation receiver arrays feature “on-board” multiplication, amplificationand delivery of local oscillator drive, eliminating long, costly runs of overmoded waveguide andreducing low frequency noise. Pre-amplification of signals in-band with MMIC LNAs significantly reduces system noise temperature, while reliability is enhanced through the use of liquid
crystal polymer (LCP) hermetic packaging and modules. Wide bandgap GaN MMICs exhibit
larger breakdown voltage, allowing for very high power density, necessary for robust, low-noise
operation in harsh environments. Digital beamforming systems using high speed field
programmable gate array (FPGA) controllers as well as direct digital synthesis will provide
active beam steering and beam focusing/defocusing capabilities to MIR.
The plasma imaging program is supported by an aggressive program in synthetic modelling.
A robust MIR synthetic diagnostic platform has been developed by integrating the PPPL
full-wave reflectometer code and a diffraction-based optical code to simulate the full diagnostic
process using either experimentally determined perturbed density profiles or from plasma
simulations [4]. A similar synthetic diagnostic is also being developed for ECE-Imaging. This
work is supported by US DoE grants DE-AC02-09CH11466 and DE-FG02-99ER54531.
[1] C.M. Muscatello, C.W. Domier, X. Hu, G.J. Kramer, N.C. Luhmann, Jr., X. Ren, P.
Riemenschneider, A. Spear, B.J. Tobias, E. Valeo, L. Yu, Rev. Sci. Instrum. 85, 11D834702(2014).
[2] A.G. Spear, C.W. Domier, X. Hu, C.M. Muscatello, X. Ren, B.J. Tobias, N.C. Luhmann,
Rev. Sci. Instrum. 85, 11D834 (2014).
[3] C. Luo, B.J. Tobias, B. Gao, Y. Zhu, J. Xie, C.W. Domier, N.C. Luhmann, T. Lan, A. Liu,
H. Li, C. Yu and W. Liu, J. Instrum. 9, P12014 (2014).
[4] X. Ren, C.W. Domier, G. Kramer, N.C. Luhmann Jr., C.M. Muscatello, L. Shi, B.J. Tobias
and E. Valeo, Rev. Sci. Instrum. 85, 11D863 (2014).

- 47 -

O-7
Recent results from the H-1 heliac on MHD activity, fluctuations and RF
propagation
B.D. Blackwell, C.A. Michael, S.R. Haskey, A. Thorman, F.J. Glass, D.G. Pretty,
C. Nührenberga, A. Koniesa, M.Colea, M.J. Holeb, V. Moiseenkoc, B. Seiwald and J. Howard
Plasma Research Laboratory, RSPE, ANU, Canberra, ACT 0200, AustraliaaMax-Planck-Institut für Plasmaphysik, Greifswald, Germanyb
RSPE, ANU, Canberra, ACT 0200, Australiac
IPP, National Science Center "Kharkiv Institute of Physics and Technology", Ukraine

H-1 is a medium-sized (R ~ 1m, <a> ~ 0.15-0.2 m) 3 field period flexible heliac with access toa wide range of magnetic configurations. Plasma generated in H-1 (H/D/He mixtures, B~0.5,
ne~1018) by RF at ωcH shows a very complex dependence on configuration; both the electron
density and the nature of fluctuations vary in a manner correlated with the presence of low orderrational values of rotational transform. Results from both coherent, MHD range fluctuations
and broadband fluctuations in the range 1-200kHz are reported, and a new method forvisualising RF waves is described.
Coherent modes ~5-100kHz are detected by threemagnetic probe arrays, displaying Alfvénic
scaling in many cases. New results separating theeffects of magnetic field, rotational transform,
number and mass density will be presented. The
accompanying density fluctuations have been
imaged by tomography of three views of carbon
ion optical emission, revealing their 2D structure.
This is in good agreement with the mode structureof BAEs predicted by the CAS3D MHD code, Figure 1: Identification of the dominant M H D
induced by the strong shaping of H-1. Mode modes in H -1, n/ m = 5/ 4 and 4/ 3
identification using data mining techniques (Figure 1) and possible drive mechanisms will be
discussed. For example, the strong shaping produces a large n=6, m=2 component in the |B|
spectrum allows excitation of the 5/4 mode by populations ~ one order slower than VAlfvén.
Measurements of the structure of edge turbulence using twin Langmuir probes have revealed
poloidally extended structures in both density and potential at low frequencies. Their narrow
radial extent resembles zonal flows found in both Ar discharges in H-1 and other devices.
Fluctuation-induced flux has been evaluated for higher frequency modes, including BAEs andbroadband turbulence, and correlated with changes in overall confinement. Langmuir probes
have also been used to determine the configuration dependence of confinement in the vicinity of
the 7/5 resonance. When the resonance is located near the plasma edge in a magnetic hill region,
the plasma density exhibits strong dithering type behaviour as well a decrease in edge density,
characteristic of flattening in the 7/5 island or ergodic region.
The first images of propagating radiofrequency waves were obtained by an RF-synchronised
camera viewing the H-1 plasma under the RF antenna, illuminated by a localised supersonic gas
puff. The observed waves appear to have a wavelength smaller than the fast wave, and arestrongly affected by the position of the ion-cyclotron resonance layer. Observations will becompared with predictions of a finite element cylindrical RF code, and plans for a compact
Uragan-type Alfven heating antenna will be presented.

- 48 -

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

Friday,  6th  November,  2015  
  
  
  
  

  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

  

PL-4
Commissioning and First Plasma Operation at the Wendelstein7-X
Stellarator
H.P. Laqua, W7-X team,
Max Planck Institute for Plasma Physics, EURATOM Association, 17491 Greifswald, Germany

Wendelstein7-X is the first large Stellarator with super-conducting modular coils. Its design andconstruction has been a technologic, engineering and logistic challenge. 50 non-planar coils
and 10 planar coils had been produced and assembled with high precision and quality
requirements. Thermal insulation was fault-free placed in a narrow 3-dimesional space inbetween the cold coil casing and the plasma vessel. Even more the components inside theplasma vessel required precise positioning in the 3-dimensional environment. Most of thecomponents need a full steady state design for continuous heat loads and have to take into
account the additional load by microwave stray radiation of the electron cyclotron heating(ECRH) with a power of 10 MW for 30 minutes. High heat load components have beendeveloped and tested for steady state operation at a maximal power load of up to 10 MW/cm2.
The plasma start-up and long pulse operation will be sustained by a 10 MW ECRH system at afrequency of 140 GHz. The ECRH will be supported by NBI and ICRH for short pulse (<20 s)operation.
The main machine assembly, in particular closing the cryostat, had been finished in March
2014. The consecutive commissioning runs with in its schedule. First the cryo-vessel was
successfully pumped down and the coils cool down was performed within 4 weeks. When thecoil temperature was below 4°K the commissioning of the super conducting system started.
The nominal magnetic field strength of 2.5 T on axis could be achieved, which is the
pre-condition for ECRH plasma operation at the second harmonic electron cyclotron resonance.
The safety system for the coil current fast shut down was successfully tested. In parallel theplasma vessel was pumped down. Next the magnetic flux surfaces could be measured. Different
coil currents generate different magnetic field configurations. The characteristic magnetic flux
surfaces and island structure were found as theoretically predicted. Therefore major
misalignments could be excluded. This demonstrates already the feasibility of the modular
stellarator concept.
For the first test plasma operation (OP1.1 [1]) W7-X is equipped with a graphite limiter in eachof the five modules. A set of most important diagnostics is installed. The ECRH system iscommissioned. Six of seven gyrotrons with a power of up to 5 MW are in operation.
Further preparation work is ongoing now. The central safety system and radiation protection ispresently evaluated by the government authorities in order to receive the operation license soon.
[1] T. Sunn Pedersen et. al., ”Plans for ﬁrst plasma operation of Wendelstein 7-X”, submitted to
Nuclear Fusion- 49 -

I-15
Progress of Gyrokinetic Turbulence Simulation in Helical Plasmas
M. Nunami, A. Ishizawa, M. Nakata, S. Toda, T.-H. Watanabea , and H. Sugama
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japana Department of Physics, Nagoya University, Nagoya, 464-8602, Japan
nunami.masanori@nifs.ac.jp

Quantitative evaluations of anomalous transport ﬂuxes of particles and heat in magnetically conﬁned plasmas are critical issues for the design of fusion reactors. The gyrokinetic
approaches are powerful to analyze the transport phenomena, which are considered to be
driven by the drift-wave plasma turbulence. Numerical simulations based on the gyrokinetics, however, requires a huge computational resources for calculations of turbulent transport
in toroidal plasmas. Especially, in three-dimensional system such as Large Helical Device(LHD), the higher spatial resolutions are required to cover the complicated three-dimensional
ﬁeld structure, which are obtained from the equilibrium calculation corresponding to the experimental observations.
In the past few years, a number of gyrokinetic codes which are available for simulating the
turbulence phenomena in the three-dimensional ﬁeld conﬁgurations have been developed,
e.g., GENE-GIST[1] and GKV/GKV-X[2], and the codes were veriﬁed each other in thebenchmarking calculations. Using the codes, the turbulence simulations have been performed.
The ion temperature gradient (ITG) turbulence simulations, for example, showed signiﬁcantresults that the simulations reproduced the turbulent spectra and the transport coeﬃcient in
the LHD experiment, and the optimized ﬁeld conﬁgurations may reduce the transport levels[3, 4]. Recently, we extend the GKV/GKV-X code to include the electromagnetic eﬀects [5],and multi-species plasmas with precise collisions [6, 7].
On the other hand, it is strongly demanded to establish a reduced transport model thatcan quickly reproduce the nonlinear simulation results, because the gyrokinetic simulationsin helical systems tend to consume huge computational resources. Using linear instability
analyses, a primitive reduced model has been constructed for pure ITG turbulent transportin helical plasmas under the collisionless limit [8]. Linear growth rates of the instabilities andthe linear zonal ﬂow responses in the model, we applied the model to integrated transport
simulations, and it can predict the plasma proﬁles in LHD plasmas.

[1] P. Xanthopoulos et al., Phys. Plasmas 16 (2009), 082303.
[2] M. Nunami et al., Plasma Fusion Res. 5 (2010) 016.
[3] M. Nunami et al., Phys. Plasmas 19 (2012), 042504.
[4] T.-H. Watanabe et al., Phys. Rev. Lett. 100 (2008), 195002.
[5] A. Ishizawa et al., Nucl. Fusion 55 (2015), 043024[6] M. Nunami et al., Plasma Fusion Res. 10 (2015), 1403058.
[7] M. Nakata et al., Compt. Phys. Comm. (2015) in press.
[8] M. Nunami et al., Phys.Plasmas 20 (2013), 092307.

- 50 -

I-16
Kinetic physics of magnetic reconnection in collisionless plasmas
C. Z. Chenga, S. Inoueb, Y. Onoa, and R. HoriuchicaDepartment of Advanced Energy, University of Tokyo, Kashiwa, Chiba 277-8561, Japanb
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japanc
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Highlights of the collisionless plasma kinetic physics during driven magnetic reconnection arepresented. In particular, we explain the physical pictures of (1) how electron and ion dynamics
decouple and how the charge separation and electrostatic electric field are produced in themagnetic field reversal region (reconnection current layer and outflow exhaust) and around theseparatrix regions, (2) how electrons and ions gain energy in the reconnection current layer, (3)why the electron outflow velocity in the reconnection exhaust reaches super-Alfvenic speed andthe ion outflow velocity reaches Alfvenic speed and how the parallel electric field is produced,(4) how electrons are accelerated by the parallel electric field around the separatrix region, and(5) how ions gain energy when they move across the separatrix region into the downstream. We
also show that electrons and ions gain energy mainly from the inductive reconnection driven
electric field and less from the electrostatic electric field. Applications to understand spaceobservations and laboratory experiments will be discussed.

- 51 -

I-17
Geospace Exploration Mission: ERG
Y.Miyoshia, I. Shinoharab, T. Takashimab, K. Asamurab, N. Higashiob, H. Matsumotob,
T. Mitanib, S. Kasaharab, S. Yokotab, Y. Kazamac, M. Hiraharaa, Y. Kasabad,
A. Matsuokab, H. Kojimae, Y. Katohd, K. Shiokawaa and K. Sekiaa

Solar-Terrestrial Environment Laboratory, Nagoya University, Nagoya 464-8601, Japan
ISAS/JAXA, Sagamihara 252-5210, Japanc
ASIAA, Taipei , TaiwandTohoku University, Sendai 980-8578, Japan,e
RISH, Kyoto University, Uji 610-0011, Japanb

We present the geospace exploration project: ERG (Energization and Radiation in Geospace).
Relativistic electrons of the outer belt largely change associated with the solar wind variations,and often increase largely particularly during the storm recovery phase. Several wave-particle
interactions have been proposed as plausible acceleration processes. The inward radial diffusion
driven by the interaction between the MHD mode waves and drifted electrons around the Earthprovides significant accelerations via the betatron acceleration. The acceleration caused by thewave-particle interactions with whistler mode waves also contributes the large flux
enhancement of relativistic electrons. In order to investigate the acceleration mechanism ofrelativistic electrons of the outer radiation belt as well as the space storm dynamics, the ERG
project is going [1]. The ERG project consists of the satellite program, the ground-based
network observations, and the integrated study group. The ERG satellite will be launched in2016 that is a declining phase of solar cycle 24. The planned apogee altitude is about 4.5 Re andthe perigee is about 300 km. The ERG satellite is designed to have a comprehensive set ofplasma/particle sensors that can measure electrons from 12 eV to more than 10 MeV and ionsfrom 10 eV/q to about 200 keV/q with a mass discrimination. The satellite can also measure
field and waves from DC to 10 MHz (electric field) and to100 kHz (magnetic field) in order toobserve several plasma waves in the inner magnetosphere that are essential for the particle
acceleration, transportation, and loss. Besides these instruments, newly developed
wave-particle interaction analyser that can measure directly energy exchange process between
electrons and whistler mode waves [2], which provides essential data sets to understand thenon-linear wave particle interactions in the space plasma.
[1] Y. Miyoshi et al. AGU Monograph 199,2012.
[2] Y.Katoh et al., JPS, 1,015100, 2014.

- 52 -

I-18
Magnetized plasma modelling: common approach interesting for fusionand processing plasmas
T. Minea1, A. Revel1, S. Mochalskyy2, A. Lifschitz3
1

Laboratoire de Physique des Gaz et Plasmas, UMR 8578 CNRS, Université Paris-Sud, 91 405 Orsay Cedex,
France2
Max-Planck-Institut für Plasmaphysik, Garching, Germany
3
Laboratoire d'Optique Apliqueé, ENSTA/CNRS/Polytechnique UMR 7639, Palaiseau, France

Magnetized plasmas operate at low pressure and due to the efficient trap of electrons they lead to a highionization degree of the neutral gas, even if they remain partially ionized. Hence a large variety of sourcesof charged particles exploits magnetized plasmas to produce electrons and positive or negative ions, often(i) extracted from the plasma and further accelerated to higher energies for several applications or (ii) useddirectly in interaction with the electrodes, the substrate or plasma immersed surfaces for technologicalprocessing.
As an example for the first case, the future experimental reactor ITER will use the energy delivered to thecore plasma by two beams of neutral Deuterium particles of 16 A @ 1 MeV each. In addition to energy,
they should provide also momentum to the tokamak plasma, allowing to reach the H-mode (High advanced
confinement mode) by current drive. The neutralization reaction by simple stripping of the negative ions(D-) has a significantly high cross section at 1 MeV than the electron attachment on energetic deuterons.
Definitely, the ITER's negative beam injector (NBI) is designed to operate with negative ions extracted
from a low pressure-high density plasma through a complex (3D) magnetized structure.
Another example for the second category is taken among the recent technologies of ionized physical vapor
deposition (I-PVD) processing, known as High Power Impulse Magnetron Sputtering (HiPIMS) and used
for specific coatings. In spite of the large scale industrial mastering of the deposition process, the further
optimizations of the HiPIMS demand necessarily the fundamental understanding of the plasma phase andthe transport of particles in the reactor volume, at different moments.
Analogously, the NBI optimization requires a deeper understanding of the meniscus formation and the 3D
extraction of NI together with the co-extracted electrons, and the intrinsic relationship with the gas kinetics
in the accelerator, as well as in the plasma source volume and at the surface.
Fully 3D Particle-in-Cell numerical modelling coupled with the Monte Carlo treatment of collisions (PICMCC) can effectively simulate these types of complex situations involving time dependence, complex gas
kinetics, self-consistent 3D evolution of the plasma or beam particles in the presence of 3D magnetic fieldtraps. Even if these two cases are apparently very different, the same numerical approach proved theirefficiency giving valuable results on the mechanism of extraction of negative ions, the geometry andexpected extracted current of the beams produced, the transport of these beams and the role of the plasmas
created in the accelerator and in the neutralization region and provides information at nanosecond scale on
the plasma instabilities (‘spokes’, flares, etc.) governing the electron transport across the complex magnetic
structures related with co-called anomalous transport.

- 53 -

O-8
Self-Organization and Heating by Inward Diffusion
in Magnetospheric Plasmas
N. Sato, Z. Yoshida, and Y. Kawazura
Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa, Chiba 277-8561, Japan

In magnetospheric plasmas, spontaneous confinement is achieved by self-organization of
charged particles through the process of inward diffusion. The resulting creation of density
gradients appears as a violation of the entropy principle. Recent studies [1] have suggested thatfoliation of phase space resulting from topological constraints imposed by adiabatic invariants
is consistent with maximization of entropy on the associated symplectic leaves [1-2].
Consequently, description of the dynamical process leading to the formation of the plasma
clump requires the formulation of an appropriate diffusion operator on the distorted metric of
the foliation. This work is carried out in [3] where a Fokker-Planck equation for inward
diffusion is obtained. Here, with the aid of numerical simulations, we investigate the heating
mechanism associated with the above diffusion process and study the resulting temperature
anisotropy.
An interesting relationship between temperature anisotropy and conservation of the second
adiabatic invariant 𝐽∥ emerged. Our research suggests that time-scale and strength ofelectromagnetic fluctuations responsible for inward diffusion drastically change the behaviorof the Fokker-Planck equation in [3]: fast and strong diffusion destroys the second adiabatic
invariant (𝐽∥̇ ∝ 𝑣𝐸×𝐵 /𝜔𝑏 with 𝑣𝐸×𝐵 the 𝐸 × 𝐵 velocity generated by fluctuations and 𝜔𝑏
the bounce frequency). This results in preferential
heating of the perpendicular temperature as
particles climb up the magnetic field by preserving
the first adiabatic invariant. The opposite scenario
is observed for slow, weak diffusion, where no
preferential heating occurs. This situation isrepresented in figure 1.
From a theoretical perspective, we expect theratio between time-scale of fluctuations and periodof bounce oscillations to determine the definition of
Figure 1: spatially averaged plasma
the stochastic integral that has to be taken in thetemperatures versus time. ∗ : fast, strong
Fokker-Planck equation.
diffusion. #: slow, weak diffusion.

[1] Z. Yoshida and S.M. Mahajan, Prog. Theor. Exp. Phys. 2014 (2014), 073J01.
[2] N. Sato, N. Kasaoka, and Z. Yoshida, Phys. Plasmas 22 (2015), 042508.
[3] N. Sato and Z. Yoshida, J. Phys. A: Math. Theor. 48 (2015), 205501.

- 54 -

I-19
Development and Initial Cruise Operation of the Hayabusa2 Ion Engine
SystemK. Nishiyama, S. Hosoda, R. Tsukizaki, H. Kuninaka
Japan Aerospace Exploration Agency, 3-1-1 Yoshinodai, Chuo-ku, Sagamihara 252-5210, Japan

Hayabusa2 is the second asteroid sample return mission by JAXA [1]. The ion engine system(IES) for Hayabusa2 is based on that developed for Hayabusa with modifications necessary to
improve durability, to slightly increase thrust, and to reflect on lessons learned from Hayabusa
mission [2]. Hayabusa2 will rendezvous with a near-earth asteroid 1999 JU3 and will takesamples from its surfaces. More scientific instruments than Hayabusa including an impactor to
make a crater and landers will be on board thanks to the thrust enhancement of the IES. An
improved neutralizer with stronger magnetic field for longer life has been under endurance test
in diode mode since August 2012 and has accumulated the operational hours of 25800 h (>mission requirement: 14000 h) until August 2015. The IES flight model and the spacecraft wasdeveloped within 2.5 years (Figure 1). The spacecraft was launched from Tanegashima Space
Center in Kagoshima Prefecture on-board an H-IIA launch vehicle on December 3, 2014. The
IES was quickly checked out in orbit and cruise operation by ion propulsion for the electricdelta-V Earth gravity assist (EDVEGA) was almost completed in the first week of June 2014.
Accumulated operating times for four ion thrusters are 594 h, 11 h, 54 h and 592 h, respectively.
Trajectory correction maneuvers by IES and reaction control system (RCS) will be executed
after September if necessary before the planned earth swing-by on December 3, 2015.

Figure 1.
Asteroid explorer Hayabusa launched in 2003 (Left) and its successor
Hayabusa2 launched in 2014 (Right).
[1] H. Kuninaka and Hayabusa-2 Project, 30th International Symposium on Space Technology
and Science, 2015-k-61, 2015.
[2] K. Nishiyama, H. Kuninaka, Trans. JSASS Aerospace Tech. Japan 10, No ists28, (2012) pp.
Tb_1-Tb_9.

- 55 -

I-20
The origin and the impact of stochasticity resulting from edge localized
perturbations.
Michael Tendler
Fusion Plasma Physics, Alfvén Laboratory, Royal Institute of Technology,
SwedenAbstract
Stochastization of the magnetic fields resulting from overlapping magnetic
islands has been addressed because of the importance of the 3 D effects causedby perturbation fields and currents at the edge and in the scrape-off layer in atokamak. The main mechanism results from the emergence of the radial electron
stochastic conductivity in a braided magnetic field. This electron current isclosed by the ion radial current well known from biasing experiments andcausing a significant pump-out effect. The impact of the stochastization isdemonstrated to affect ELM´s events via RMP and/or filaments existing within
the SOL. The resulting electric fields from the balance of currents are strongly
affected changing their value and even the sign under certain conditions . The
impact on the momentum balance is also important because convective fluxes
emerge due to the axial jirxB torque.

- 56 -

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

Poster  Session  1  Tuesday,  3rd  November,  2015  
  
  
  

  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

  

P1-1
Change in Reynolds stress induced by end-plate biasing
in a cylindrical laboratory plasma
T. Kanzakia, Y. Nagashimab, S. Inagakib, T. Yamadac, Y. Miwaa, F. Kina, A. Fujisawaa,
N. Kasuyab, Y. Kosugad, M. Sasakib, M. Lesura, S.-I. Itohb and K. ItoheaInterdisciplinary Graduate School of Engineering Sciences, Kyushu University,
Kasuga, Fukuoka 816-8580, Japanb
Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japanc
Faculty of Arts and Science, Kyushu University, Fukuoka, Fukuoka 819-0395, Japand
Institute for Advanced Study, Kyushu University, Kasuga, Fukuoka 812-8581 Japane
National Institute for Fusion Science, Toki, Gifu 509-5292. Japane-mail address:kanzaki@riam.kyushu-u.ac.jp

Formation of the shear flow structure in magnetized plasmas is now widely recognized to
be important because it is accompanied by the reduction of turbulence-driven transport. Thus
prediction capability of flow structure is strongly required to control the turbulence transport.
The Reynolds stress is a candidate for the fundamental driving force for the self-consistentflow generation in magnetized plasmas by micro-scale turbulent fluctuations [1]. The
determination of the Reynolds stress requires specific diagnostics. In small laboratory
plasmas, probe measurement allows us to evaluate the ExB flow and its fluctuations. A
multi-channel Reynolds stress probe system has been developed in a linear device, PANTA[2]. Each probe has six tungsten tips. Four of them are simultaneously used to measure theradial and azimuthal electric field and the others are used for ion saturation current
measurement. To understand the flow formation mechanism, end-plate biasing experiments
have been performed in PANTA and change in the drift-wave characteristics during biasing
are found[3].
Here we report changes in radial profiles of the Reynolds stress induced by end-platebiasing. The drift wave amplitude is significantly reduced during the biasing [3] while few
changes in the azimuthal flow in the central region are observed. The impact of the end-platebiasing on the radial structure of the Reynolds stress will be discussed.
This work is partly supported by a grant-in-aid for scientific research of JSPS, Japan(15H02335, 15H02155, 23244113, 26420852) and by the collaboration programs of NIFS
(NIFS13KOCT001) and of the RIAM of Kyushu University and Asada Science foundation.
[1] P. H. Diamond, et al., Plasma Phys. Control. Fusion 47 (2005) R35[2] Y. Nagashima, et al., Rev. Sci. Instrum. 82 (2011) 033503[3] T. Yamada, et al., Nucl. Fusion 54 (2014) 114010- 57 -

P1-2
How to Identify Ion Temperature Gradient Mode in Cylindrical Plasma
Y. Miwaa, S. Inagakib, Y. Nagashimab, N. Kasuyab, T. Kobayashic, Y. Kosugad,
T. Yamadae, M. Sasakib, H. Arakawaf, F. Kina, K. Itohc and S.-I. ItohbaInterdisciplinary Graduate School of Engineering Sciences, Kyushu University,
Kasuga, Fukuoka 816-8580, Japanb
Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japanc
National Institute for Fusion Science, Toki, Gifu 509-5292. Japand
Institute for Advanced Study, Kyushu University, Kasuga, Fukuoka 812-8581 Japane
Faculty of Arts and Science, Kyushu University, Fukuoka, Fukuoka 819-0395, Japanf
Teikyo University, Omuta, Fukuoka, 836-8505, Japane-mail address:yudai_miwa@riam.kyushu-u.ac.jp

In fusion plasmas, a radial ion heat transport driven by Ion Temperature Gradient(ITG) modes is considered to be dominant [1]. For improvement of confinement,
comprehensive understanding of features of ITG turbulence is necessary. The ITG turbulence
has been studied theoretically [1] and substantial progress in the study has been made innumerical simulation [2]. In experiments, a few observations in tokamak indicated thatfluctuations propagating in the ion diamagnetic drift direction were the candidate for ITG
modes [3], and some spectrum analyses under a particular condition indicated an excitation of
the ITG mode in the linear plasma [4]. However, other evidences are needed for a full
identification of the ITG mode, such as the ion temperature fluctuation measurement. It isdifficult in core region of toroidal plasmas. The impact of ITG turbulence on heat transport
has not been fully verified as well. In linear plasmas, multi point measurement of these toconfirm corresponding to the density/ITG whole distribution is much easier.
In the basic linear plasma: Plasma Assembly for Nonlinear Turbulence Analysis(PANTA) in Kyushu University, we have succeeded in observing a fluctuation propagating in
the ion diamagnetic drift direction, which co-exists with drift waves propagating in theelectron diamagnetic drift direction. The linear wave analysis by using the fluid simulation
indicated possibility of excitation of ITG modes in PANTA and an optimum condition to
excite the ITG mode was obtained, which depends on not only ITG but plasma scales [5]. The
excitation region of a fluctuation (propagating in the ion diamagnetic drift direction) in thespace of plasma radius and magnetic field was in good agreement with the prediction by thesimulation. Here we report the details of excitation condition and identification of
spatiotemporal structure of the observed modes. The parallel wavenumber of these modes wasevaluated to be finite (order of the inverse of plasma length) from a phase difference between
probes on the same magnetic field line at distant locations. Cross-correlation technique wasapplied to experimental data observed by multiple Langmuir probes aligned along themagnetic field line. In addition, we have developed a conditional averaging method toestimate electron/ion temperature fluctuations by using the probes. Observation oftemperature fluctuations allows us to evaluate heat flux driven by these modes. Behaviors oftemperature fluctuations and heat flux will be discussed.
[1] W. Horton, Rev. Mod. Phy. 71, 735 (1999)[2] W. Dorland et al., Science 275, pp. 289-292 (1997).
[3] A. K. Sen et al., Phys. Rev. Lett. 66, pp. 429-432 (1991)[4] D.W. Ross et al., Phys. Fluids B 3 2251 (1991)[5] Y. Miwa et al,. Plasma Fusion Res. 8, 2403133 (2013)- 58 -

P1-3
Observation of radial scale length of fluctuations in PANTA
T. Tarumia, S. Inagakib, Y. Nagashimab, T. Yamadac, Y. Miwaa, F. Kina, A. Fujisawab,
N. Kasuyab, Y. Kosugad, M. Sasakib, M. Lesurb, K. Itohe and S.-I. ItohbaInterdisciplinary Graduate School of Engineering Sciences, Kyushu University,
Kasuga, Fukuoka 816-8580, Japanb
Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japanc
Faculty of Arts and Science, Kyushu University, Fukuoka, Fukuoka 819-0395, Japand
Institute for Advanced Study, Kyushu University, Kasuga, Fukuoka 812-8581 Japane
National Institute for Fusion Science, Toki, Gifu 509-5292. Japane-mail address : tarumi@riam.kyushu-u.ac.jp

Radial scale of turbulence in magnetized plasmas characterizes transports driven by
turbulence and thus a scaling law of the radial length of turbulence is of central importance.
The characteristic radial scale length of micro-fluctuations, e. g. the drift waves, is considered
to scale with the drift scale (the ratio of sound speed to ion cyclotron frequency) which is of
the order of an ion gyro-radius [1]. Many experiments, however, have shown that transportdoes not simply scale the gyro-radius [2]. A closer experiment–theory comparison for
turbulence structure size is strongly required to predict the turbulence driven transport in thelarge fusion plasma. While at the same time, the stationary small laboratory plasma allows
better diagnostics accessibility. Probe arrays have proven already successful as a diagnostic
for turbulence studies in fusion and magnetized low-temperature plasmas [3]. Thus we
studied the spatial structure of the turbulence in the low-temperature plasma by probe arrays.
In order to vary the drift scale in a wide range, we changed the magnetic field strength from500 G to 1500 G and ion mass number by using helium and argon as working gases in thelinear plasma device PANTA [4].
Here we report the experimental observations of the radial correlation length offluctuations. The correlation lengths of the drift waves will be compared with the drift scales.
We will also vary the plasma radius from 2 cm to 8 cm and discuss about its effect on radialtransport scaling.
This work is partly supported by a grant-in-aid for scientific research of JSPS, Japan(15H02335, 15H02155, 23244113) and by the collaboration programs of NIFS
(NIFS13KOCT001) and of the RIAM of Kyushu University and Asada Science foundation.
[1] W. Horton, Rev. Mod. Phy. 71 (1999) 735[2] C. C. Petty, et al., Phys. Rev. Lett. 74 (1995) 1763
[3] A. Latten, et al., Rev. Sci. Instrum. 66 (1995) 3254
[4] H. Arakawa, et al., Plasma Phys. Control. Fusion 52 (2010) 105009- 59 -

P1-4
Reconstruction of Cross-Sectional Structure of Reynolds Stressin End-Plate Biasing Experiment in PANTA
1,2

1,3

1,2

F. Kin, Y. Nagashima , T. Yamada , S. Inagaki , T. Kanzaki, Y. Miwa,
4
5
1,2
1,2
2
H. Arakawa , T. Kobayashi , N. Kasuya , M. Sasaki , M. Lesur ,
2,6
1,2
1,5
1,2
Y. Kosuga , A. Fujisawa , K. Itoh , S.-I. Itoh
Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-1 Kasuga-koen,
Kasuga 816-8580, Japan
1
Research Center for Plasma Turbulence, Kyushu University, 6-1 Kasuga-koen, Kasuga 816-8580,
Japan
2
Research Institute for Applied Mechanics, 6-1 Kasuga-koen, Kasuga 816-8580, Japan
3
Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
4
Teikyo University, 6-22 Misakimachi, Omuta 836-8505, Japan
5
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292. Japan
6
Institute for Advanced Study, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581
Japan

In magnetized plasmas, micro-scale plasma turbulence interacts with meso-scale structures
such as zonal flow and streamer. Evaluation of momentum transport between different scalesof fluctuations is important to understand the structure formation in plasma turbulence.
Spatial gradient of Reynolds stress is a main mechanism for flow generation by turbulence. In
magnetized plasmas, axial-scale length of plasma is assumed to be much longer thanazimuthal- and radial-scale lengths, observations of the dynamics of turbulence, flow andReynolds stress in the azimuthal cross-section allow us to identify the causal relation among
them. Direct measurement of Reynolds stress has been performed in linear plasma devices [1],and has also been performed in PANTA (Plasma Assembly for Nonlinear Turbulence
Analysis) with Reynolds stress probe [2]. For reconstructing two-dimensional patterns ofReynolds stress, cross-correlation analysis between azimuthal 64-channel probe array and theradial movable Reynolds probe is used. The azimuthal probe array sets 250 mm away from
Reynolds stress probe in magnetic field direction, and measures floating potential for
azimuthally. Conditional averaging methods are used to reconstruct azimuthal mode
structures. Hear we report the details of the reconstruction method and observed
2-dimentional structures of density and potential fluctuations, mean flows and the Reynolds
stress. We also carried out the end-plate biasing experiment in PANTA [3] to induce
perturbations to the plasma. During biasing, turbulence was suppressed and Reynolds stress
was drastically changed. The changes in two-dimensional patterns will be discussed.
This work is partly supported by a grant-in-aid for scientific research of JSPS Japan(15H02335, 15H02155, 23244113, 26420852) and by the collaboration programs of NIFS
(NIFS13KOCT001) and of the RIAM of Kyushu University and Asada Science foundation.
[1] C. Holland, et al., Phys. Rev. Lett. 96, 195002 (2006)[2] Y. Nagashima, et al., Rev. Sci. Instrum. 82, 033503 (2011)[3] T. Yamada, et al., Nucl. Fusion 54, 114010 (2014)- 60 -

P1-5
Evaluation of flow velocity with a frequency comb Doppler Backscatteringin PANTA
R. Minatoa, S. Inagakib, T.Mizokamia, T. Yamadac, Y. Nagashimab, T. Tokuzawad,
Y. Miwaa, F. Kina, A. Fujisawab, N. Kasuyab, K.Itohd and S.-I. ItohbaInterdisciplinary Graduate School of Engineering Sciences, Kyushu University,
Kasuga, Fukuoka 816-8580, Japanb
Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japanc
Faculty of Arts and Science, Kyushu University, Fukuoka, Fukuoka 819-0395, Japand
National Institute for Fusion Science, Toki, Gifu 509-5292, Japane-mail address:minato@riam.kyushu-u.ac.jp

A multi-channel Doppler backscattering diagnostics is the most promising to observe theflow structures and their dynamic behaviors simultaneously in magnetized plasmas [1]. We
have applied the microwave frequency comb technique to the reflectometry in the linear
plasma device PANTA and succeeded to observe density fluctuations at 29 different radii
simultaneously [2]. We also applied this technique to the Doppler backscattering diagnostics.
Typical operational parameters of PANTA are as follows; plasma diameter of about 120 mm,
axial magnetic field of 0.09 T, and filling argon pressure of 3 mTorr. Typical central densityand electron temperature are 1.0 × 10−19 m−3 and 3 eV, respectively. Reflectometer and probe
measurements indicated that the electron density gradient is steep in the radius of r = 30-40mm, and produces drift waves, which propagates in the nearly perpendicular direction to themagnetic field (in the electron diamagnetic direction) with finite axial wavelength. The
Doppler backscattering wavenumber is selected via an adjustable launch angle. In this study,
the incident and reflected wave signals are directly transferred to the digital storage
oscilloscope, which has a frequency band of 33/50 GHz (the sampling frequency is 80/160GHz), so the waveforms of the incident and reflected signals are detected in the form ofdigital signals with very high temporal resolution [2]. The Doppler frequency is calculated by
the delta-phase method [3]. Advanced digital signal processing techniques (e.g. Hilberttransform and convolution) allow us to improve the signal-to-noise-ratio of evaluating the
Doppler frequency. Results obtained with a Backscattering diagnostics will be compared withprobe measurements.
This work is partly supported by a grant-in-aid for scientific research of JSPS, Japan(15H02335, 15H02155, 23244113) and by the collaboration programs of NIFS
(NIFS13KOCT001) and of the RIAM of Kyushu University and Asada Science foundation.
[1] T. Tokuzawa et.al. ,Plasma Fusion Res.9(2014)[2] T. Mizokami et. al., 24th International Toki conference (2014)[3] S. Inagaki et.al., Plasma Fusion Res. 8 (2013).
[4] G. D. Conway et al ., Plasma Phys. Control . Fusion47(2005)- 61 -

P1-6
Helium volumetric recombining plasma formationfor energetic ion injection in a radio-frequency plasma device
DT-ALPHA
H. Takahashi, A. Okamotoa , T. Miura, D. Nakamura, P. Boonyarittipong,
S. Sekita, S. Kitajima
Dept. of Quantum Science and Energy Engineering, Tohoku Univ., Sendai 980-8579, Japana Dept. of Energy Engineering and Science, Nagoya Univ., Nagoya 464-8603, Japan

In magnetically confined fusion devices, heat flux mitigation by detached plasma formation
is important issue. Because reaction rate of the volumetric recombination becomes large in a
low electron temperature plasma, suﬃcient plasma-neutral interactions are required to maintain stable plasma detachment. On the other hand, energetic particles are exhausted from
confinement region and then transported in divertor region by edge localized modes in a highconfinement mode plasma. Therefore, understanding of the divertor plasma dynamics thatcoexists with intermittently inflowing energetic particles is recognized as an important subject in fusion related studies. A divertor plasma simulator with radio-frequency (RF) plasma
source oﬀers some advantages for such divertor plasma dynamics study because cylindricallywound RF antenna permits the penetration of the energetic particles.
Experiments were performed with the RF plasma device DT-ALPHA [1]. Plasma production
method is a 13.56 MHz RF discharge. Typical electron temperature produced in the DTALPHA device is several eV, so helium secondary gas is supplied into the device at thedownstream region to form helium recombining plasma [2].
To investigate the volumetric recombining plasma dynamics which coexists with energetic
ions by beam injection, radial distribution of the volumetric recombination is important because beam components pass around central axis of a cylindrical target plasma. In a low
electron temperature and high electron density plasma, electrons lose their energy throughenergy exchange between ions. Therefore, electron-ion temperature relaxation time τTei isimportant to induce volumetric recombination [3]. τTei is dependent on the electron temperature and density. In the DT-ALPHA device, the electron temperature and electron density
are controlled by changing RF heating power or neutral pressure distribution. Spectroscopic
and Langmuir probe measurements indicate that τTei aﬀects helium volumetric recombining
plasma formation. In this presentation, we will discuss about helium volumetric recombining
plasma formation required for beam injecting experiment based on the electron-ion temperature relaxation.
This work is partly supported by JSPS KAKENHI Grant numbers 22740357, 26420848 andby JSPS Fellows Grant number 264331.
[1] A. Okamoto, et al., Plasma Fusion Res. 7 (2012) 2401018[2] H. Takahashi, et al., Fusion Sci. Technol. 63 1T (2013) 404[3] N. Ezumi, et al., J. Nucl. Mater. 241-243 (1997) 349- 62 -

P1-7
Development of linear plasma device
for proof of principle of energetic ion injection
S. Tsubota, A. Okamotoa, K. Shimizu, D. Nakamura, T. Kobayashi, H. Takahashi, S. Kitajima
Department of Quantum Science and Energy Engineering, Tohoku University, Sendai 980-8579, Japana
Department of Energy Engineering and Science, Nagoya University, Nagoya 464-8603, Japan

In the study of magnetically conﬁned fusion plasma, conﬁnement of the fusion products isimportant. One of the possible methods to investigate conﬁnement performance of fusion
products experimentally is using energetic ions produced by neutral beam injection [1].
However, the above method has some restrictions for energetic ion production due to ﬁxed
beam line and ionization point. Moreover it is difﬁcult to apply this method to smaller devices
and lower density target plasmas. We have been developing a compact ion injector, which
consists of hydrogen storage material containing hydrogen molecules. It is expected toaccelerate hydrogen ions by using the electron sheath potential between the plasma and thesurface of positively biased electrode. Using the injector inside the last closed ﬂux surface,
more ﬂexible experiments are expected. It has been observed that emission of hydrogen atomline increased according to increase in bias voltage of the electrode in the small standard
heliac device (Tohoku University Heliac) [2].
The linear plasma device, in which a target plasma for proof of principle of theabove-mentioned method is produced, has been developed. The plasma is produced by theelectron cyclotron resonance (ECR) heating using a microwave of frequency 2.45 GHz.
Maximum injection power is 6 kW. Total length and inner diameter of the device are 2.2 m
and 0.2 m, respectively. Two groups of magnetic coils are located. One produces a magnetic
beach configuration (B ~ 0.1 T) for ECR heating in a plasma production region [3], and theother produces uniform conﬁguration of 0.3 T in a measurement region. Components for theplasma production region have been constructed, and the plasma was successfully produced.
We measured the electron temperature and density of the helium plasma in the plasma
production region using Langmuir probe (typically Te ~ 10 eV, ne ~ 1 × 1017 m−3) andobtained their radial proﬁles. Dependence of the Te and ne on plasma heating power, magnetic
ﬁeld strength, and neutral pressure was also investigated. In the measurement region, we havea plan to produce the magnetic ﬁeld by a pulse operation. Therefore, we developed a circuit
consists of capacitors and inductors for a pulse forming network of current source (~1.2 kA).
In preliminary experiments the ﬂat top of the magnetic ﬁeld (~10 ms) was successfully
produced, and the current in the magnetic coils in the plasma production region induced by
the pulse operation in the measurement region was investigated quantitatively.
This work is supported by Grant-in-Aid for Scientific Research (KAKENHI), 24246152.
[1] R. K. Fisher et al., Rev. Sci. Instrum. 81, 10D307 (2010).
[2] A. Okamoto et al., Rev. Sci. Instrum. 85, 02B302 (2014).
[3] M. Tanaka et al., J. phys. Soc. Jpn. 60 1600 (1991).

- 63 -

P1-8
Observation of Plasma Dynamics in TPD-sheet IV
	
  

	
  

	
  

	
  

T.Onda, S. Kajitaa, A. Tonegawab, T. Iijimab, and N. Ohno
Graduate School of Engineering, Nagoya Univ, Furo-cho,Chikusa-ku, Nagoya, 464-8603, Japana
EcoTopia Science Institute, Nagoya Univ, Furo-cho,Chikusa-ku, Nagoya, 464-8603, Japanb
Department of Physics, School of Science, Tokai Univ, 4-1-1 Kitakaname, Hiratukai,259-1292, Japan

Sheet plasmas have been used in various fields of researches such as a divertor simulator infusion plasmas and, negative ion production [1, 2]. For ex., previously studies, negative ion
density in a sheet plasma was controlled with aim of reducing the heat flux to the divertor
plates [3]. Furthermore, the temperature and density associated with the generation of negative
ion were investigated [4]. Although expensive experiments have been conducted in sheet
plasmas up to now, the dynamic behavior of sheet plasma has not been fully understood yet. In
order to utilize the sheet plasma more efficiently and meaningfully, it is of interests toinvestigate the dynamic behavior and understand the physics of the sheet plasma. In this work,
we have observed dynamic behaviors of a sheet plasma with a high-speed camera. The dynamic
behavior of sheet plasmas and corresponding releases of blob-like plasmas to the periphery of
the sheets are shown.
The experiment was performed in the linear divertor plasma simulator TPD Sheet-IV. The
plasma was divided into two regions: a plasma source region and an experimental region fordiverter simulator [2]. In order to observe the dynamic behavior of sheet plasma structures, ahigh-speed camera (ULTRA CAM HS-106E : NAC Image Technology, Inc.) was used at aframe rate of 100000 fps. A heat-resistant glass was installed at the end of TPD Sheet-IV toobserve the plasma in parallel to the magnetic field. A Langmuir probe was installed from theupper side of the device and measured the ion saturation current at the periphery of the plasma.
We observed the dynamic behavior of a pure helium plasma or helium argon mixture plasma.
Additional hydrogen gas was injected in the experimental region.
Figure 1 (a) and (b) shows fast framing camera images of He and Ar mixture plasmas andpure He plasma, respectively, observed with the high-speed camera. It is seen in Fig.1(a) thatthe plasma column moves back and forth horizontally at a frequency of about 4 kHz in He/Ar
mixture plasma. However, in case of pure He plasma, the plasma column did not exhibit such a
dynamic behavior. In addition, intermittent transport which is similar to plasma blob wasobserved. The temporal evolution of Isat in the peripheral plasma supported the fact that blob
like plasma was released at the same frequency of ~4 kHz.
This work was supported by JSPS KAKENHI Grant Number (25289337). This work is
partially supported by NIFS/NINS under the project of Formation of International Network for
Scientific Collaborations and NIFS Collaboration Research Program NIFS14KUGM094[1] A. Ando, T. Kuroda, et al. : Rev.Sci. Instrum.,Vol 61, No 1, January (1990)442-444[2] A. Tonegawa, K. Kumita, et al.: Jpn. J. Appl. Phys., Vol. 45, No. 10B (2006)8212-8216[3] M. Ono, A.Tonegawa, et al.: Journal of Nucl. Mater 337–339 (2005) 261–265[4] S. Matumoto, et al.: JPS Conf.Proc. 1, 015042(2014)

Fig. 1 Experimental setup

Fig. 2 A snapshot of image captured by the high-speed
camera:(a)He and Ar plasma (b)He plasma- 64 -

P1-9
Experimental and Numerical Studies of Hall Eﬀect on Merging
Formation process of a Field-Reveresed Configuration
Y. Kaminou, X. Guo, M. Inomoto, Y. Ono, and R. Horiuchia
Graduate School of Engineering, the University of Tokyo, 7-3-1 Yayoi, Bunkyo-ku, Tokyo, Japana National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanCounter-helicity spheromak merging[1] is one of the formation methods of a Field-Reversed
Configuration(FRC). In counter-helicity spheromak merging, two spheromaks with opposing
toroidal fields merge together, through magnetic reconnection events and relax into a highbeta FRC, which has no or little toroidal field. This process contains magnetic reconnection
and a relaxation phenomena with energy conversion from magnetic energy to kinetic orthermal energy, and the Hall eﬀect has some essential eﬀects on these process because the
X-point in the magnetic reconnection or the O-point of the FRC has no or little magnetic
field. In this poster, we conducted 2D/3D Hall-MHD simulations and experiments of counterhelicity spheromak merging. We found that the Hall eﬀect enhances the reconnection rate,
and reduces the generation of toroidal sheared-flow. The suppression of the “slingshot eﬀect”
may aﬀect the relaxation process. We will discuss their details in the poster.

[1] M. Inomoto et al, Phys. Rev. Lett., 97 (135002) 2006

- 65 -

P1-10Generation of Energetic Electrons during Spherical Tokamak Merging in
UTST
T. Ushiki, X. Guo, H. Furui, K. Matsuyama, H. Koguchia , M. Inomoto
Graduate School of Frontier Sciences, The University of Tokyo, 5-1-5, Kashiwanoha, Kashiwa, Chiba
277-8656, JapanaNatonal Institute of Advanced Industrial Science and Technology, 1-1-1, Umezono, Tsukuba, Ibaraki
305-8586, Japan

Spherical tokamak (ST) merging method has been studied as one of the CS-less start-up
schemes in START, MAST (UKAEA), TS-3/TS-4 (Univ. Tokyo). In merging start-up
high-power plasma heating through magnetic reconnection is expected to form high beta STs in
short period, but a small number of experimental studies have been carried out on the electron
acceleration/heating during reconnection in ST merging. In this paper, experimental
investigation of electron acceleration during reconnection in the presence of strong guide field
will be presented.
In recent experimental result in MAST, extremely localized electron heating was observed
near the X-point during magnetic reconnection [1]. The heating mechanism is unclear, but
recent numerical work with 3-D PIC model suggests that electrons are effectively accelerated in
toroidal direction by reconnection electric field in presence of the strong guide field [2]. These
accelerated electrons may contribute to the electron heating observed in MAST. The goal of thisresearch is to verify the electron acceleration during magnetic reconnection with strong guidefield.
Surface barrier detector (SBD) was employed to observe soft x-ray emission by electron
bremsstrahlung during magnetic reconnection.
2 µm thickness polycarbonate foil was equippedin front of the SBD as X-ray absorption filter toeliminate photon with low energy (< 200eV).
Sharp soft x-ray burst was observed only during
magnetic reconnection with good correlation
with reconnection electric field. This suggeststhat electron acceleration by toroidal electricfield accounts for the soft x-ray emission. Fig.1
shows dependence of soft x-ray intensity on
toroidal guide field and toroidal electric field.
Soft x-ray intensity showed clear dependency
on both toroidal electric field and guide field.
Fig.1 Dependence of soft X-ray
These results suggest that the efficiency of the(polycarbonate 2μm) intensity on toroidal
electron acceleration could be roughly evaluatedguide field and toroidal electric field.
by the effective electric field Et (Bt/Bp), whichis employed to predict the electron acceleration
in torus break down near a null point [3].

[1] Y. Ono. Et al, Plasma Phys and Control. Fusion 54 (124039) 2012.
[2] P. L. Pritchett and F. V. Coroniti, J. Geophys. Res. 109 (A01220) 2004.
[3] B. Lloyd. Et al, Nuclear Fusion. 31 (2031) 1991.

- 66 -

P1-11
Characterization of an atmospheric pressure plasma
by a Gerdien condenser
M.C.C. Lacdan and M. Wada
Graduate School of Science and Engineering, Doshisha University, 1-3 Tatara Miyakodani,
Kyotanabe-shi, Kyoto-fu 610-0394, Japan

A diagnostic instrument based on the Gerdien condenser [1] principle is being developed as a
new characterization method for laboratory-produced atmospheric pressure plasma. The
Gerdien condenser instrument is capable of determining the ion density and mobility, thus ion
species composition of the plasma. In this study, a miniaturized Gerdien condenser has been
designed and fabricated to measure parameters of atmospheric plasma produced by an RF
excited atmospheric pressure plasma pen.
The condenser has a 1.0 cm diameter 6 cm long cylindrical collector electrode mounted in abias electrode of 1.7 cm inner diameter. The entire condenser is mounted in an aluminum maderectangular shield box. An atmospheric pressure plasma pen produces a plasma by 13.56 MHz
RF power source. The plasma produced with 30 W discharge power was analyzed by theGerdien condenser while supplying Ar and N2 gases at 5 and 3 liters per minute flow rates,
respectively.

current (A)In Fig. 1 are shown typical current-voltage (I-V) characteristics obtained by applying voltage
sweep from 0 V to 12 V at the Gerdien condenser's bias electrode and measuring the collectedcurrent by the inner electrode. The flow rate of fan attached at the end of the condenser was
set at 1.88, 3.75, 5.63
and 7.5 x 10-4 m3/s.
The figure clearly
3E-10
indicates
thesaturation current of
2.5E-10
thecondenser
increases with the
2E-10
flow
rate
that
determines the intake
1.5E-10of atmospheric ions by
the condenser. The
1E-10
values of ion mobility
1.88
calculated from the
3.75
5E-11
condenser
I-V
5.63
characteristics
are
7.5
0
identified to be those
0
2
4
6
8
10
12
of O2+, Ar+, Ar++ and+voltage (V)N2 . The determined
ion densities are in therange from 1.93 x 105 Figure 1. Current-voltage characteristics of atmospheric pressure plasma
pen for varying fan flow rate of 1.88, 3.75, 5.63 and 7.5 x 10-4 m3/s.
to 2.24 x 106 cm-3.
[1] J.A. Chalmers, Atmospheric Electricity 2nd edition, (Pergamon Press, Oxford, 1967) p.
53-53, p. 128-130.

- 67 -

P1-12
Supersonic Gas Injection for High Density Operation and AssociatedDevelopment of 140 GHz Interferometer in RELAX RFP
S. Nakanobo, S. Masamune, A. Sanpei, G. Ishii, Y. Aoki, T. Nagano, M. Higuchi, K. Jikuhara,
R. Tsuboi, H. Himura, T. Akiyama1, H. Okada2, T. Mizuuchi2
Kyoto Institute of Technology, Matsugasaki, Sakyo-ku, Kyoto 606-8585, Japan
1
National Institute for Fusion Science, Oroshi-cho, Toki 509-5292, Japan
2
Institute for Advanced Energy, Gokasho, Uji 611-0011, Japannakano14@nuclear.es.kit.ac.jp

Reversed-Field Pinch (RFP) is one of the magnetic confinement systems for high-temperature, high-beta fusion plasmas. In the RFP, as the aspect ratio (A) is lowered, theself-induced bootstrap current tends to increase, which may lead to a reduction of the externalsource for current drive.
RELAX [1] is a low-aspect-ratio RFP whose major radius R is 0.5 m and the minor, a, 0.25
m. The objectives include geometrical optimization of the RFP configuration, so, the aspectratio A was designed as low as 2. The equilibrium analysis [2] showed that we could expect
sizable fraction of the bootstrap current with beta of ~30% at Ip ~100kA with Te of ~200-300 eV
in A=2 RFP, the details depending on the equilibrium profiles. Figure 1 shows the central
electron poloidal beta βpe(0)=ne(0)Te(0)/(Bp2(a)/2μ0) vs. electron density ne normalized to theGreenwald density nG achieved in RELAX to date, where the central electron temperature anddensity were measured with Thomson scattering [3]. We should note that since Bp>>Bt in the
RFP, the poloidal beta is almost equal to the total beta. Figure1 shows that the beta increases
with density, whose upper bound is ~30% of nG. It istherefore important to achieve higher density region tosee if higher beta could be realized there.
We have developed a supersonic gas injection
system for higher density operation in RELAX. The
system consists of a fast-acting electromagnetic gas
valve with a Laval nozzle to produce directed supersonic
Fig.1 Electron poloidal beta vs. density
gas flow with Mach number of up to 5 (M=5). A 140
GHz heterodyne interferometer was also developed tomeasure ne in higher regime than before [4]. The system
consists of two oscillators and an IQ detector, which
provide orthogonal components of the phase sift arisingfrom the difference in optical path lengths for plasma
and reference legs. In Fig. 2, a gas-puff fueled discharge
is compared with a non-fueled case in the initial
experiments. The electron density is almost doubled
near the current flat top at ~1.5ms into the discharge.
Details of the fuelling efficiency will be discussed alongwith the results from further optimized cases.
[1] S. Masamune et al., IAEA FEC25, EX/P3-52 (2014).
[2] A. Sanpei et al., J. Phys. Soc. Jpn. 78 013501 (2009).
[3] R. Ueba et al., PFR 9, 1302009 (2014).
[4] M. Sugihara et al., PFR 5, S2061 (2010).

- 68 -

Fig.2 Results from initial gas-puff
experimentP1-13Evidence of increased ion temperature by slow wave heating in RT-1
M. Nishiura, Z. Yoshida, Y. Yano, Y. Kawazura, T. Mushiake, H. Saitoh, M. Yamasaki,
A. Kashyap, N. Takahashi, M. Nakatsuka, A. Fukuyamaa
Graduate School of Frontier Sciences, The University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa
277-8561, Japana
Department of Nuclear Engineering, Kyoto University, Nishikyo-ku, Kyoto 615-8540, Japannishiura@ppl.k.u-tokyo.ac.jp

The ring trap 1 (RT-1) device realizes a "laboratory magnetosphere" created by a levitatedsuperconducting ring magnet. The physical processes in the vicinity of a magnetic dipole arestudied for magnetosphere dynamics and advanced thermonuclear fusions.
The electron cyclotron resonance (ECR) heating with 8.2 GHz and ~50 kW produces theplasmas with hot electrons in the range of a few ten keV. We reported that the local electron
beta exceeds 1 in RT-1 plasmas. In such situation, the ions still remain cold at a few ten eV,
because the heating source is dominated by Coulomb collisions between cold electrons andions. Heating ions is expected to access high ion beta state and to improve the plasma
confinement theoretically. Therefore the ion cyclotron range of frequencies (ICRF) heatingwith 2 - 4 MHz and 10 kW was prepared in RT-1. Based on the results of the TASK-WF2 code,the ∩ shape loop antenna was designed for a slow wave excitation, and was implemented in the
RT-1.
In the ICRF heating experiments, a base plasma was sustained by 8.2 GHz-ECRH for 1 s,
and an ICRF heating was applied with the delay of 0.2 ms from the plasma startup. We
observed the clear increase in diamagnetic signals and impurity ion temperature (C III) in
helium plasmas at the neutral gas pressure of 3 mPa, if the ICRF power of ~ 10 kW iscomparable to the ECRH one. The increases of ion temperatures for C III and He II wereobserved near the levitation coil (the coil did not levitate in this case), where the ion cyclotron
layer for H+ is located. The result is the first time in a magnetosphere plasma device, although
the heating effect has a possibility of electron heating by a fast wave that results in the ion
heating by Coulomb collisions. The results during the ICRF heating will be discussed in detail.

- 69 -

P1-14
Anisotropy in tail component of Hα in RT-1
Y. Kawazura, Z. Yoshida, M. Nishiura, Y. Yano, H. Saitoh, T. Nogami, T. Mushiake, M.
Yamasaki, A. Kashyap, N. Takahashi and M. Nakatsuka
Graduate School of Frontier Sciences, The University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa
277-8561, Japan

Magnetospheres are self-organized plasma confinement device commonly seen in the space.
Inhomogeneous magnetic field causes inward diffusion which gives rise to planetward density
gradient. The ring trap 1 (RT-1) device simulates laboratory magnetosphere by the levitated superconducting magnet. Plasma is produced by electron cyclotron resonance heating.
The inward diffusion in RT-1 was proved by the interferometric measurement of the peaked
electron density profile [1].
Recently the ions in RT-1 were found to beenergized by betatron acceleration concomi(a)tant with inward diffusion [2]. The heating
power was higher than that by the relaxation with the electrons. In this study, we
decomposed the emission line profile of Hαdipole field magnet
with double Gaussian distribution to obtain
the radial profile of hot component of neutral hydrogen. The primary loss mechanism
for ions, in RT-1, is considered to be charge
exchange with neutral particles, whose mean(b)free path is longer than the device size. The
tail component of Hα emission may inherit
the information of protons since the spontaneous emission occurs much faster than thetime scale of thermal motion of the neutralhydrogen particles. Figure (a) shows theradial profile of the ratio of line integrated
emission intensity (hot/cold) for two different filling gas pressure operations. For the
low filling gas pressure, the emission intensity of the hot component surpassed that of the cold component near the dipole magnet,
while little increase of intensity ratio was seen for the high gas pressure. At the same time,
we observed temperature anisotropy estimated by Doppler broadening of tail component only
for low filling gas pressure operation (Fig. (b)), which is consistent with the preceding ion
energy transfer model.
[1] H. Saitoh et al., Nucl. Fusion 51, 063034 (2011).
[2] Y. Kawazura et al., arXiv:1507.03692.

- 70 -

P1-15
Studies on power dependence of plasma production and ion heatingby a single helical antenna for a simple thruster
Y. Nakatsuka, F. Kurashita, S. Nakamoto, H. Takeno
Department of Electrical and Electronic Engineering, Kobe University,
1-1, Rokkodai-cho, Nada-ku, Kobe 657-8501, Japan

Recently, technologies for long distance and long time missions are required in the spaceengineering such as a farther space exploration or a sample return, so development of an
efficient electrical propulsion engine is necessary. The conventional electrical propulsion
engine cannot control a thrust and a specific impulse individually because they are inversely
proportional to each other under the condition of the constant input power. It also has adurability problem because electrodes are abraded by plasma.
The VASIMR engine was researched and developed in NASA and AD ASTRA Co. It cancontrol a thrust and a specific impulse individually by separating a plasma production section
and an ion heating section. A radio frequency (RF) power is inductively used, so the problem of
the durability on electrodes can also be avoided [1]. However, the device is not necessarily
efficient on the aspects of size and structure as it uses two RF power supplies and two antennas
for individual control of plasma production and ion heating.
For downsizing of the device, a new method was proposed in which plasma production and ion
heating were performed simultaneously by using one helical antenna [2]. The helical antenna
has a bi-directional excitation characteristic. For a right helical antenna, R wave, so-called fast
wave, is excited in the –z direction under the magnetic field in the +z direction. On one hand, L
wave, so-called slow wave, is also excited in the +z direction simultaneously. The R wave has
right-rotating field and a higher cut-off frequency, so high density plasma could be produced.
The L wave has left-rotating field and could couple with ion motion, so acceleration of ions
could be expected in the ion cyclotron resonance frequency. The high density plasma
production and the ion heating are corresponding to a strong thrust and a high specific impulse,
respectively, thus it is expected that a single helical antenna with one RF power supply cancontrol a thrust and a specific impulse individually by controlling a magnetic field. This
scenario was qualitatively confirmed experimentally [3], however, the quantitative examinationis necessary to design an actual engine. In this paper, we will show an RF power dependence ofplasma production and ion heating in this scheme.
The power dependence of plasma production and ion heating is not considered to be a simple
one. Even if the magnetic field is in the ion heating condition, a part of the input power isconsumed for plasma production. Moreover, increase of plasma production power results in an
increase of plasma density which may make ion heating complex as more power is required forthe higher density plasma heating and the propagation characteristic of L wave will be changed
according to the density increase. The heating characteristic to RF input power would becomecomplex, and a thrust and a specific impulse would be controlled by combined adjustment ofmagnetic field and RF input power.
This paper would clarify those heating characteristics experimentally. The calibrated powermonitor system was installed in the experimental device, and the experiments are in progress.
The detailed experimental results and discussion will be shown in the conference.
[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. Hayashi, et al., Trans. Fusion Sci. Tech. 63(1T) (2013) 389.

- 71 -

P1-16
Characteristics of Magnetic Fluctuations during Magnetic
Reconnection in Counter-helicity Spheromak Merging Experiment
R. Yanai, A. Kuwahataa and M. Inomotoa
Graduate School of Engineering, The University of Tokyo, Bunkyo, Tokyo, Japana Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa, Chiba, Japan

Magnetic reconnection is an important phenomenon in magnetized plasmas because it leads
to sudden release of magnetic energy and global changes of magnetic field structure. Fluctuations generated during magnetic reconnection are likely to aﬀect reconnection speed andenergy conversion process [1,2] as well as energy/helicity transport. We investigated magnetic
fluctuation spectra during TS-4 counter-helicity merging experiment in which both poloidal
and toroidal fields reconnect (no guide field case). The fluctuations had diﬀerent characteristics in the downstream region and at the reconnection X-point. The fluctuation power with
0.5-5.0 MHz frequency component concentrated on outboard-side downstream region (Fig. 1(a)), whereas the fluctuation power with 7-20 MHz frequency component was localized aroundthe X-point (Fig. 1 (b)) and propagated almost radially. Since the lower hybrid frequency
near the X-point was lower than 15 MHz, this higher frequency mode detected near the Xpoint is supposed to belong to the whistler mode. Dispersion relation of the fluctuations near
the X-point indicated that the angle between the wave vector and magnetic field was about
60-80 degrees. The fluctuation power spectral density (PSD) observed in the outboard-side
downstream region and near the X-point are shown in Fig. 1 (c). Since the observed fluctuation has higher frequencies than the ion cyclotron frequency and the reconnection structureis smaller than the ion inertia length, the turbulence could be expressed in electron MHD
framework. The PSD near the X-point clearly shows inverse cascade occurring in the imbalanced turbulence driven by the whistler waves[3]. These fluctuation characteristics could
play a key role in vanishment of the helicity during counter-helicity merging.
(a)

(c)

(b)

Fig. 1: Spatial distribution of magnetic fluctuation power of (a) 0.5-5 MHz and (b) 7-20 MHz
components during counter-helicity spheromak merging. Orange dots indicate measurement points
and cross mark represents the X-point. (c) Power spectral density (PSD) of the toroidal component
fluctuations measured at diﬀerent radial point of (blue) near X-point and (cyan) outboard side.

[1] H. Ji et al., Phys. Rev. Lett. 92 (2004) 115001.
[2] M. Inomoto et al., Phys. Plasmas. 20 (2013) 061209.
[3] H. Kim and J. Cho, Astrophys. J. 801 (2015) 75.

- 72 -

P1-17
Role of Kinetic Alfvén Wave in Energy Transportation in Magnetopause
Abstract
The transportation of energy is the vital subject in space physics. Magneto hydrodynamic waves
are known to play very important role in plasma transportation. In the present work, we havestudied the localization of kinetic Alfvén wave (KAW), which is three dimensionallypropagating. These localized structures might be a possible candidate for transported energy at
small scales in dissipation range of turbulent spectra. For this scenario, governing dynamical
equation of KAW has been derived taking ponderomotive nonlinearity into account.
Furthermore, these equations are solved numerically using pseudo spectral method inmagnetopause regime to analyse the localized structures of KAW and corresponding turbulent
spectrum. Spectrum shows steeper scaling followed by inertial range and this steepening is dueto the transportation of energy from large to small scale, which eventually contributes to heatingof plasma. The present mechanism may be useful for explaining the plasma heating and particle
acceleration in magnetopause.The present study is correlated with the observations made by
Chaston et al. [2008] using THEMIS data.
Reference
Chaston, C., et al. (2008), Turbulent heating and cross-field transport near the magnetopause
from THEMIS, Geophys. Res. Lett., 35, L17S08,doi:10.1029/2008GL033601.

- 73 -

P1-18
Numerical study of spectral line shapes in high-density He plasmas
H. Kawazome, S. Nambaa
Department of Information Engineering, National Institute of Technology, Kagawa College, 551
Kohda Takuma-cho, Mitoyo 769-1192, Japana Graduate School of Engineering, Hiroshima University, 1-4-1 Kagamiyama, Higashihirosima
739-8527, Japan

In order to determine electron temperature and density in plasmas, spectral line intensity
ratio of appropriate transitions provides a useful method for plasma diagnostics, in which theexperimental ratio is compared with the value calculated by using a collisional-radiative (CR)model. However, in high-density plasmas, such as atmospheric thermal plasmas and divertor
plasma of magnetically confined devices, an optical depth becomes significantly thick. For
example, although He arcjet plasmas have been studied extensively for fundamental researchof arcjet thrusters and atomic/molecular physics, it is found that the radiation trapping ofresonance transitions alters the population distribution. Consequently, we cannot apply theintensity ratio method for determinations of plasma temperature and density. In this study,
therefore, coupled rate equations incorporating the radiation transport are solved to derive
the population densities and analyze spectral line shapes of the resonance spectra. The
transitions considered are 1s2 1 S - 1s2p 1 P, 1s2 1 S - 1s3p 1 P and 1s2 1 S - 1s4p 1 P, and therespective wavelength are 58.4 nm, 53.7 nm and 52.2 nm. In the calculation, a slit function
is not taken into account. The plasma is a cylinder shape having a radius of 0.25 cm and alength of 0.5 cm and divided by four regions. The spatial profiles of the electron temperatureand density are given by T e(l) = T0 (1−(l/0.5)2 )2 eV ne = n0 (1−(l/0.5)2 )2 cm−3 , respectively,
where T0 and n0 are 0.5 eV and 1.0 × 1013 cm−3 , respectively. The ion density is equivalent
to that of electron density, and gas temperature is set to electron temperature. The atomic
density (=1.0 × 1015 cm−3 ) and temperature (Gaussian: 300 K) are constant throughout thevolume. Figure 1 shows the time evolution of calculated line profiles for 1s2 1 S - 1s2p 1 Ptransition. As clearly seen, the profiles change to flat top shape with time, because the line
1.2

Normalized Intensity (arb. units)-4

t=10 s1.0-6

t=10 s-8t=10 s

0.8

0.6

0.4

0.2

0.0
5.13335

5.13345

5.13355

15

Frequency (10 Hz)

Figure 1: Time evolution of spectral line profiles for 1s2 1 S - 1s2p 1 P transition.
center has the highest absorption probability. Details of the results will be discussed.

- 74 -

P1-19Doppler reflectometer system for measuring plasma rotation velocity
in GAMMA 10
J. Kohagura, T. Tokuzawaa, M. Yoshikawa, Y. Nakashima
Plasma Research Center, University of Tsukuba, 1-1-1 Tennodai, Tsukuba, Ibaraki 305-8577, Japana
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japankohagura@prc.tsukuba.ac.jp

Doppler reflectometry is currently used as a powerful technique to measure the perpendicularvelocity of density fluctuations, the radial electric field, and the perpendicular wave number
spectrum in many fusion plasma devices [1-3]. In GAMMA 10 a Doppler reflectometer has
recently been installed to measure perpendicular rotation velocity of density fluctuations in
the cylindrical plasma at the central cell as shown in Fig. 1. The Doppler reflectometer has anantenna system with a launching/receiving scalar feed antenna and focusing mirrors. It gives aquasi-Gaussian beam with beam waist ω of ~5 cm at the plasma edge. By rotating one of themirrors the tilting angle of the incident microwave can be controlled against the normal of
cutoff layers. A frequency synthesizer is used in the range 11.5–18 GHz as the stablemicrowave source of x-mode probing beam for GAMMA 10 plasma having typical peakdensity ~2×1018 m-3. Simple direct conversion microwave circuit is built using I-Q mixer
which combines the reference signal from thesynthesizer and the back-scattered signal fromplasma and produces in-phase (I=Acosφ) and
quadrature (Q=Asinφ) signals. The I and Q
signals are sampled at 1MHz by a fast
digitizing oscilloscope. Spectrum analyzer haspoor time resolution but it helps to confirm thefrequency spectrum of the back-scatteredsignal directly. The first preliminary results ofDoppler shifted complex spectra and radial
profiles of the perpendicular velocity ofdensity fluctuations are presented for ICRF
start-up plasma with additional ECH. The
rotation of fluctuations during additional ECH
period is found to become opposite direction Fig. 1 Schematic drawing of the Doppler
reflectometer system in GAMMA 10.
comparing to the case of ICRF heating alone.

[1] M. Hirsh et al., Plasma Phys. Contral. Fusion 43 (2001) 1641.
[2] G. D. Conway et al., Nucl. Fusion 46 (2006) S799.
[3] T. Tokuzawa et al., Plasma Fusion Res. 9 (2014) 1402149.

- 75 -

P1-20Co-extraction of Low Energy Ar+ and C2H2- Ions for Space Charge Effect
Mitigation and Increased Beam Intensity
C.L.S. Mahinay, M. Wadaa
Department of Physics, School of Science and Engineering, Ateneo de Manila University, Quezon City
1108, PhilippinesaGraduate School of Science and Engineering, Doshisha University, Kyotanabe, Kyoto 610 -0321,
Japan

Low energy ion beams, especially negative ion beams, have fairly low intensities due to thespace charge effect. Negative ions generated from a target surface by plasma sputtering
typically yield very low intensities. To significantly increase the beam intensity of negative ion
beams, Cs is introduced into the plasma volume which lowers the work function of the target
surface and increase negative ion yield [1]. However, Cs is difficult to handle and is a hazardous
substance. To increase the intensity of positive ion beams, electrons are introduced to neutralize
the beam and mitigate the space charge effect [2]. In this study, positive and negative ion beams
are co-extracted from a hot cathode filament-type DC plasma source as a means to increase theintensities of either beam. The positive ions consists mainly of Ar+ ions and the negative ionsare C2 H2 - ions. Energies do not exceed 800eV. A Wien filter mass analyzer is used to obtainthe mass spectra of the ion beam and a 127 o sector electrostatic energy analyzer (ESA) is used
to measure the ion energy distribution function. The ESA and Wien filter can be coupled tomeasure the ion energy distribution function of specific ion masses. The co-extraction of thepositive and negative ions increased the beam intensity of the Ar + ions to a maximum of 10-8A
from 10-9 A. The C2 H2 - beam intensity also increased from 10 -11 A to 10-9 A. Hence, we have adevised a simple way of increasing the intensity of either low energy positive or negative ion
beams by means of co-extracting both beams simultaneously. This method of co-extraction isa much safer alternative than using Cs in increasing the negative ion intensity especially at low
energies.
400Vt 500Vt
600Vt
700Vt 800Vt
400Vext 400Vext 400Vext 400Vext 400 Vext

1.4E-08
1E-08

8E-09

Postive and negative
coexctraction
Positive ion extraction0-5E-106E-09

intensity (A)intensity (A)1.2E-08

4E-09

2E-09

-1E-09-1.5E-09

0-2E-09Positive and
Negative ion
coextraction

Negative ion
extraction-2.5E-09

Fig. 1. Comparing intensities from both coextracted ion beam and positive only ion beam.

Fig. 2. Comparing intensities from both coextracted ion beam and negative only ion beam.
A geometric trend line is fitted to the negative
ion extraction intensities[1] J.R. Hiskes, A.M. Karo, P.A. Willmann and W.J. Stevens, Physics Letters A, 68 (1978)
221.
[2] V. Dudnikov and A. Dudnikov, Rev. Sci. Instrum. 73 (2002) 995.

- 76 -

P1-21Synthesis of hydrogenated amorphous carbon nitride using DC graphite
hollow cathode discharge
Catherine Joy M. Dela Cruz, Sidney M. Palardonio, and Henry J. RamosaaPlasma Physics Laboratory, National Institute of Physics,
University of the Philippines Diliman, Quezon City, Philippines

Hydrogenated amorphous carbon nitride (a-C:H:N) was synthesized on Si substrates in aplasma-enhanced chemical vapour deposition (PECVD) facility operating in direct current
graphite hollow cathode discharge (HCD) mode. Methane percentage, filling pressure anddischarge voltage of the CH4/N2/Ar mixture plasma were varied. Prior to deposition, opticalemission spectroscopy (OES) was used to determine the presence of active species in plasmathat are needed for carbon nitride synthesis, such as CN radicals, N2+ first negative, N2 second
positive and Ar II. Electron temperature was obtained using the slope of the Boltzmann plot of
ten Ar I lines. The optical emission spectra has showed that carbon nitride synthesis is favoured
with high discharge voltage, 6% methane and 1.0 Torr filling pressure. Changes induced in theelemental composition, surface morphology, crystallographic structure, and structural disorderin the deposited a-C:H:N were analysed by FTIR, SEM, XRD and Raman spectroscopy,
respectively. As an application, surface free energy was determined using the Owen Wendt
method.
For all the samples, FTIR showed the presence of a strong broad band peak in 3000-3500 cm-1
corresponding to OH, medium strength peaks on 2800-2900 cm-1 and 2150-2250 cm-1
corresponding to (sp3) C-H and (sp) CN triple bond, respectively. The intensity for C-H bondincreases while the CN triple bond decreases with increasing methane percentage. Raman
spectra analysis showed that ID/IG intensity ratio decreases with increasing methane percentage
for 0.6 and 0.7 Torr filling pressures. XRD and SEM confirmed the amorphous properties of
a-C:H:N deposits. There is an increase in the surface free energy for the treated samples ranging
from 37.5-41.9 compared to the untreated sample with 35.3.

- 77 -

P1-22
Study on operating principle of Cockcroft-Walton circuit
to produce plasmas using high-voltage discharge
T. Fukuyama, K. SugiharaaaaFaculty of Education, Nagasaki University, 1-14 Bunkyo-machi, Nagasaki 852-8521, Japan
Graduate School of Education, Ehime University, 3 Bunkyo-cho, Matsuyama, Ehime 790-8577, JapanThe basis of operating principle of Cockcroft-Walton (CW) circuit [1] as shown in Fig. 1 inorder to produce high-voltage is studied using SPICE which is simulator of electrical circuit.
CW circuit experimentally contributed to breakthrough of nuclear physics to be applied to thedevelopment of particle accelerator [2]. High-voltage discharges obviously play an importantrole not only in producing plasmas but also a wide variety of fields. However, studies on
operating principle of CW circuit are not enough, though improvement and application of it
have been studied [3, 4].
As a result of our studies, we have found that: part of electrical charge stored in capacitors
remains, output value of voltage can be expressed in recurrence formula, and more time isneeded in order to boost concerning circuit which has higher number of steps as shown in Fig. 2.
Furthermore, we have studied combination between frequency of alternating current andcapacitance in capacitor in order to boost effectively as shown in Fig. 3.
[1] J.D.Cockcroft and E.T.S.Walton, Proc. Roy. Soc. Lond., A. 136 (1932) 619.
[2] J.D.Cockcroft and E.T.S.Walton, Proc. Roy. Soc. Lond., A. 137 (1932) 229.
[3] N.Marium, D.Ismail, K.Anayet, N.Khan, and M.Amran, Amer. J. Appl. Sci. 3 (2006) 2178.
[4] K.Sur, and E.Bloodworth, IEEE Trans. Circuits Syst. 37 (1990) 432.

Fig. 1: CW circuit which has k steps％− log10 C◆ 1 step■ 5 steps▲ 10 steps

log10 fFig 3: Frequency of power source and capacity (Ifcombinations of frequency and capacity enter in thedomain surround in a plot, CW circuit can work)

periods

Fig. 2: Boost rate of CW circuit with 1, 5,
and 10 steps- 78 -

P1-23Nonthermal and geometric effects on the dual-mode surface waves
in a Lorentzian dusty plasma slab

Myoung-Jae Lee
Department of Physics
Hanyang University, Seoul 04763, South Korea
mjlee@hanyang.ac.kr
Kyu-Sun Chung
Department of Electrical Engineering
Hanyang University, Seoul 04763, South Korea
kschung@hanyang.ac.kr
Young-Dae Jung
Department of Applied Physics and Department of Bionanotechnology
Hanyang University, Ansan, Kyunggi-Do 15588, South Korea
ydjung@hanyang.ac.kr

The nonthermal and geometric effects on the propagation of the surface dust acoustic
waves are investigated in a Lorentzian dusty plasma slab. The symmetric and antisymmetric dispersion modes of the dust acoustic waves are obtained by the plasma
dielectric function with the spectral reflection conditions the slab geometry. The
variation of the nonthermal and geometric effects on the symmetric and the antisymmetric modes of the surface plasma waves is also discussed.

- 79 -

P1-24High density plasma sustainment in microwave plasma source with slot
antennas
S. Kogoshi, Y. Kanesaka, H. Araya, Y. Takahashi, S. Nakatui, N. Katayama
Department of Electrical Engineering, Faculty of Science and Technology, Tokyo University of Science,
2641 Yamazaki, Noda, Chiba, 278-8510, Japankogoshi@ee.noda.tus.ac.jp

It has been shown experimentally that TM waves and near field from the antennas can sustainhigh density plasma in a microwave plasma source with slot antennas, in which surface waves
was not excited judging from mode numbers. It is well known that surface wave plasma (SWP)
can sustain high density (and low electron) plasma [1]-[3]. A resonant absorption model has
been proposed as the mechanism of the high density plasma sustainment of SWP [4]. The
model suggests that high density plasma (and low electron temperature) would be sustained if
there exists oscillating E-field parallel to a density gradient, its density is higher than the cut-off
density, and the space between the cut-off density point and the sheath edge is wide enough forelectron plasma waves to be excited. It has shown that a microwave plasma source with aresonant cavity for a TM mode was able to sustain high density plasma when it was satisfied
with above conditions [5]. Here, we show that the same microwave plasma source as thereference (5) except for the air gap length (see Fig.1), which was different from the resonantcavities for TM waves. New air gap lengths, which are corresponding to near field (L=2 mm)
and far field (L=15 mm, TM mode) irradiation, can also sustain high density plasma. Figure 2
shows that when plasma density reached cut-off density, the increasing rate of plasma density
becomes more rapid and the electron temperature starts to drop. It suggests that the energy
absorption mechanism changed from normal one to the resonant absorption at the point. The
picture of the plasma taken from the bottom is shown in Fig.3. The mode pattern suggests theplasma was not sustained by SWP.
7

Wave guideDielectric
Plasma(120x120mm)

ne (x1011 cm-3 )/ Te (eV)

Air gap (length L)

▲:ne
□:Te

6
5
4
3
2
1
0
0

Fig.1 Schematic
experimental setup.

50

100

150

200

Input power (w)Fig.2 Dependence of ne and Te on
input power (P=0.6 Torr, L=2 mm).

Fig.3 Picture of plasma(Pin=200 W, P=0.6 Torr
L=2 mm).

[1] K. Komachi, J. Vac. Sci. Technol. A11(1993) 164.
[2] M. Nagatsu, G. Xu, I. Ghanashev, M. Kanoh, and H. Sugai, Plasma Sources Sci. Technol.
6(1997)427.
[3] J. Kudela, T. Terebessy, and M. Kando, Appl. Phys. Lett. 76(2000)1249.
[4] T. Omaru, F. Komiyama, and S. Kogoshi, Jpn. J. Appl. Phys. 43(2004)2690.
[5] S. Kogoshi, Y. Yoshioka, N. Katayama, and Y. Kudo, Jpn. J. Appl. Phys. 48(2009)08HB01.

- 80 -

P1-25
Formation of Diamond-Like Carbon (DLC) on Siliconusing Low Energy Ion Beam Deposition
byJuan Paulo N. Gonda, and Dr. Christian Lorenz S. Mahinay
This study focuses on the synthesis and yield of a Diamond-Like Carbon thin film deposited on
silicon slides using low energy ion beams , with acetylene as the carbon source. The plasma source
used is a tungsten filament with current passing through. The characteristics of the plasma and ion
beam generated, such as discharge voltage, discharge current, extraction voltage and ion current, arevaried and related to the yield of DLC on the surface of the silicon. The yield is to be determined byusing an SEM.
As of writing, only argon plasma has been generated. Using the argon plasma, the ion andplasma characteristics were tested and related to one another. It has been observed that as discharge
current and extraction voltage increases, the ion current increases as well, indicating an increase in thenumber of argon ions flowing. This could lead to a greater DLC yield when using carbon ions.
Another aspect tested was how the pressure of the vacuum chamber affected the generation ofions. It was determined through tests that the optimum pressure for this particular setup is at 5e-4 torr as
at that pressure. Less than that amount would require more current in the filament to produce the samedischarge current, and more would also show the same effects.
The extraction voltage(EV) was used to determine if what was being harvested was the desired
positive ions. As the EV increased, the ion current changed from negative to positive, meaning that the
faraday cup was catching positive ions. The ion current also increased gradually, indicating that if ahigher yield is to be desired, a greater EV is the optimum setting.
Further tests will be run using an acetylene argon mixture to produce plasma to coat the slides.
Also, tests will be run to see how the changes in characteristics relate to the yield deposited on thesurface.

- 81 -

P1-26
Parallel electric fields in nonlinear magnetosonic waves
in a multi-ion-species plasma
M. Toida
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Theory and particle simulations have revealed that nonlinear magnetosonic waves can strongly
accelerate particles with various nonstochastic mechanisms [1]. The electric field parallel tothe magnetic field in a magnetosonic shock wave plays crucial roles in some of the mechanisms.
For example, the parallel electric field can cause trapping and acceleration of electrons in an
oblique shock wave. In a plasma containing electrons, positrons, and ions (e-p-i plasma), theparallel electric field can accelerate positrons.
∫
The parallel electric field E∥ and its integral along the magnetic field, F = − E∥ ds, were
quantitatively analyzed with theory and electromagnetic particle simulations [2]. The theory
and simulations showed that F can be large (E∥ can be strong) when the external magnetic
field is strong. The value of F can be estimated as eF ∼ ϵ2 mi vA2 in small-amplitude pulses
in a cold plasma, where ϵ is the amplitude of the pulse, mi is the ion mass, and vA is the
Alfvén speed. This work was also extended to e-p-i plasmas [3].
In this study, we theoretically analyze E∥ and F in nonlinear magnetosonic waves in a plasma
containing two species ions. In such a plasma, there exists two magnetosonic modes, which
we call high-frequency mode and low-frequency mode. Nonlinear behavior of these modes canbe described by Korteweg-de Vries equations for the frequency regions, ω ≪ Ωi of the lowfrequency mode and Ωi ≪ ω ≪ (Ωi Ωe )1/2 of the high-frequency mode [4, 5]. It is found thevalue of F for the high-frequency-mode pulse can be estimated as eF ∼ ϵ2 mi vA2 , whereas thatfor the low-frequency-mode pulse as eF ∼ ϵ2 (me mi )1/2 vA2 . Thus, F for the high-frequencymode pulse is (mi /me ) times as strong as F for the low-frequency-mode pulse. It indicatesthat E∥ of the high-frequency-mode pulse may influence particle acceleration in a multi-ionspecies plasma. We also derived the parallel electric field of nonlinear magnetosonic pulses in
e-p-i plasmas for the region Ωi ≪ ω ≪ (Ωi Ωe )1/2 , which was not obtained in Ref. [3] where
the expansion method for the region ω ≪ Ωi was used.

[1] Y. Ohsawa, Phys. Reports 536 (2014) 147[2] H. Takahashi, Y. Ohsawa, Phys. Plasmas 14 (2007) 112305[3] H. Takahashi, M. Sato, Y. Ohsawa, Phys. Plasmas 15 (2008) 082309[4] M. Toida and Y. Ohsawa, J. Phys. Soc. Jpn. 63 (1994) 573[5] M. Toida and Y. Kondo, Phys. Plasmas 18 (2011) 062303- 82 -

P1-27

  
  
  
  
  
  
  
withdrawal  - 83 -

P1-28Design study on double layer type interface structure
for heat exchanger of Fusion DEMO blanket
M. Kondo1), T. Tanaka2), Y. Hishinuma2), N. Suzuki3), Y. Kawamata3), H. Chikaraishi2), T. Muroga2), A. Sagara2)
1) Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 Japan,
2) National Institute for Fusion Science, Toki, Gifu 502-5292, Japan,
3) Tokai university, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292, Japan
Heat exchanger
tube

The design of heat exchanger for the liquid breeders,(a)which can achieve a large coefficient of heat transfer and asmall permeability of tritium (T), is important issues for thedevelopment of fusion DEMO reactors. The tritium
permeability in the heat exchanger can be improved by thecoating with tritium permeation barrier (TPB) by metal organic
decomposition (MOD) method or metal organic chemical(2) Thermal splay layer(1) Ceramics coating
vapor decomposition (MOCVD) method. Several ceramicby MOD or MOCVD
coatings having high thermodynamic stability, such as Y2O3,
(1µm)
(b)
Er2O3 and ZrO2, are the candidates of the TPB material for theliquid breeder blanket. The function of the TPB must be kept
during the life time of the heat exchanger. Therefore, thesecoatings should be protected in the liquid breeders by so-called
multi- layer structure as shown in Fig.1 (a). The thickness ofthe outer layer corresponds to the corrosion margin, and thethinning of the outer layer does not influence on the TPB
performance. The purpose of the present study is to design the
double layer type interface structure, which achieves thefunction of TPB and the corrosion resistance in the same time
during the long- term operation. The allowable thickness of the
double layers was evaluated based on the estimation of thecoefficient of overall heat transfer. The online monitoring
SUS316L
technology for this type of functional layers must be
Resin
Al2O3
established. The electrochemical impedance spectroscopy
thermal splay
Al2O3 MOD coating(EIS) is proposed as the online monitoring method and theFig. 1 (a) schematic of double layer
response for the degradation and the layer thinning by the
structure, (b) Surface cross section of
corrosion was simulated by Z-view code. The double layer
specimen of TEST ID 5 observed bystructures were preliminary fabricated on SUS316L steel by the
FE-SEM
MOD coating method and the thermal splaying method. The
fabrication conditions are presented in Table 1. Figure 1 (b) shows the results of surface cross sectional
observation for the specimen of test ID 5 by FE-SEM. It was found that thermal splay layer of Al2O3 with the
thickness of approximately 20µm was successfully fabricated on the Al2O3 MOD layer.
Table 1 Fabrication conditions of double layer structure
Substrate
Substrate
temperature
Thermal splay layer
Average size of
powders [µm]Spray distance [mm]Plasma output [kW]Repeated times
Photo afterthermal splay

Test ID 1
Test ID 2
Test ID 3
Disc type specimen (Φ20mm x t7mm)
SUS316L
Al2O3 MOD/ Y2O3 MOD/
SUS316L
SUS316L
30ºC - 60ºC 30ºC - 60ºC
30ºC-60ºC

Test ID 4
Test ID 5
Bullet type (Φ20mm x length 40 mm)
SUS316L
Al2O3 MOD
/ SUS316L
30ºC- 60ºC
30ºC-60ºC

Y2O3 MOD/ SUS316L
30ºC-60ºC

Al2O3
30

Al2O3
30

Y2O3
30

Y2O3
30

Al2O3
30

Y2 O 3
30

130
44
7

130
44
7

130
44
7

130
44
2

130
44
7

130
44
2

- 84 -

Test ID 6

P1-29Thermal quench of ion-beam induced luminescence of Er2O3
D. Kato, H.A. Sakaue, T. Tanaka, I. Murakami, T. Muroga, and A. Sagara
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Er2O3 is a prospective material for electric insulation and tritium permeation barrier ofadvanced blanket systems with liquid breeders and coolants, e.g., Li, Fliebe, and Li-Pb [1, 2].
It is undertake to develop optical methods for in-situ characterization of radiation-induced
defects in Er2O3 by using energetic ion bombardment. The Er2O3 crystals exhibit an ion-induced
luminescence band in 640-690 nm which is identified as 4f11 4F9/2 → 4I15/2 transition of Er3+ at
C2 cation sites. It has been demonstrated that the luminescence intensity is reduced significantly
by the energetic ion bombardment with plasma splay coating samples [3] and sintered bulk
samples [4]. The intensity reduction can be associated with decrystallization of surfaces by theenergetic ion bombardment.
In the present work, we addressed for the first time measurements of the ion-induced
luminescence with a heated sample up to 873 K. As shown in Fig. 1, it is found that theluminescence intensity vanishes above 670 K. Excited electrons in the upper levels of Er3+ ionsare dissipated non-radiatively due to interaction with ambient thermal phonons at hightemperatures. Indeed, a characteristic rate of the non-radiated depopulation of the upper levels
deduced from the emission line broadening increases with temperatures. However, this drastic
thermal quench of the luminescence is hardly to reconcile with the measured rate of the nonradiative depopulation.

Fig.1. Temperature dependence of the luminescence intensity induced by an Ar+ ion-beam (35keV, 10μA). Solid squares and circles stand for measurements ramping up the temperature, and open squares
and circles for those ramping down the temperature.

[1] T. Muroga et al.: J. Nucl. Mater. 367–370 (2007) 780.
[2] D. Levchuk et al.: J. Nucl. Mater. 367–370 (2007) 1033.
[3] T. Tanaka et al.: J. Nucl. Mater. 417 (2011) 794.
[4] D. Kato et al.: Plasma Fusion Res. 7 (2012) 2405043.

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P1-30Hydrogen solubility in molten salt FLiNaK mixed with nano-Ti powder
J. Yagi, A. Sagara, T. Tanaka, S. Takayama, T. Muroga
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan

Molten salts including FLiNaBe, FLiBe, and FLiNaK have attracted strong interests ascandidate liquid breeding material for a magnetic fusion reactor due to the low electric
conductivity, light density, and chemical stability. Their major concerns are the very lowhydrogen solubility (i.e. risk for tritium leakage) and the enhanced corrosivity by impurities
such as moisture, oxygen, and generated TF in neutron field. Considering these points, theconcept of mixing a hydrogen-absorbing metal powder with the molten salt has been proposed
by A. Sagara [1]. A hydrogen-absorbing metal such as Ti or Zr increases the effective hydrogensolubility of the molten salt. If the mixing fraction of the metal powder is small, the advantagesof the molten salt (low conductivity, light density, chemical stability, etc.) would be retained. J.
Yagi has already confirmed the 5 order increase of effective solubility [2] using FLiNaK mixed
with Ti powder whose grain size is below 38 μm. However, there was no increase for the
powder exposed to air [2]. The aim of this paper, therefore, is to make clear the effect of surface
contaminations by comparing with ideally clean powder. Fortunately, The nano-Ti powder(0.1wt%) was directly electrocrystallized in FLiNaK by I’MSEP, Japan [3].
In this work, practical hydrogen solubility of nano-Ti powder mixed FLiNaK (Fig. 1) isinvestigated using the device shown in Fig.2. Detailed results will be presented and discussed in
the conference.

Fig.1. nano-Ti powder mixed FLiNaK

Fig.2. Experimental setup[1] A. Sagara, H. Tamura, T. Tanaka, N. Yanagi, J. Miyazawa, T. Goto, R. Sakamoto,J. Yagi,
T. Watanabe, S. Takayama, the FFHR design group, Fusion Engineering and Design, 89(2014) 2114-2120[2] J. Yagi, A. Sagara, T. Watanabe, T. Tanaka, S. Takayama, T. Muroga, Fusion Engineeringand Design, in press[3] Y. ITO, T. Nishikiori and M. Tokushige, Plasma-Induced Discharge Electrolysis for
Nanoparticle Production, Molten Salts Chemistry: From Lab to Applications, 269-286 (2013)- 86 -

P1-31Fabrication and characterization of
MOD ceramic coatings on coolant duct materials
T. Tanaka, T. Chikadaa,Y. Hishinuma, J. Morimotob, T. Muroga
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japana
Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japanb
The Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan

Studies of ceramic coating on low activation ferritic/martensitic (RAFM) steels havebeen conducted mainly for suppression of tritium permeation from blanket components.
However, the coating technologies are also considered important for coolant ducts to minimize
tritium leakage during the transport to a hydrogen recovery system. Coating fabrication on
candidate duct materials and characterization are performed in the present study.
Coatings of Er2O3 were fabricated on stainless steel 316, Inconel 600 and
HASTELLOY C-276 substrates by the metal organic decomposition (MOD) method. After
painting the MOD organic liquids on the substrate surfaces, the oxide ceramic coatings wereobtained by baking the substrates in air for 1 hour at 650 ºC. Depth profiles of atomic
concentrations and energy spectra obtained by X-ray Photoelectron Spectroscopy (XPS)
showed that Fe2O3 and Cr2O3 were produced on all the three substrate surfaces under ~50 nm
thick Er2O3 coating layers. In our previous studies, the production of Fe2O3 on substratesurfaces degraded hydrogen permeation barrier performances and corrosion resistanceproperties of the ceramic coatings on RAFM steel substrates [1]. Also in the present fabrication,
growth of Fe2O3 was found to break the coating layer (stainless steel 316) or make the coating
layer rough (Inconel 600) in scanning electron microscope (SEM) observation. On the
HASTELLOY substrate, influence of the Fe2O3 production on the coating layer was moderatecompared with the others. However, since Fe2O3 is reduced by hydrogen at high temperatures,the production should be suppressed in all the substrates to obtain stable coating performances.
In the previous studies, preprocess to suppress the production of Fe2O3 in the coating
baking process has been examined for RAFM steel substrates and preferable coating
performances have been obtained [2]. Similar preprocess was applied to a HASTELLOY
substrate. The substrate was heated up to 700 oC in a vacuum of <7 x 10-3 Pa and heat treatment
was performed at 700 oC in air of 5 Pa for 30 minutes. Atomic concentrations obtained by XPS
after the heat treatment showed that the surface was covered with ~200 nm thick Cr2O3 layer
and the production of Fe2O3 was suppressed successfully. This indicates that the preprocess canbe applied to the Ni-based alloy substrates and coatings with superior performances could beobtained on the Cr2O3-covered substrate surfaces.
Regarding coating fabrication on rounded duct surfaces, test fabrication of a MOD ZrO2
coating was performed on an outer surface of an Inconel 600 pipe of 26 mm in outer diameter.
In the present study, the MOD liquid was painted on the pipe surface by silk screen printing andsmooth coating was successfully obtained after baking in an Ar flow.
Optimization of the preprocess conditions to suppress the Fe2O3 production in coating
baking on the coolant duct materials and characterization of the obtained coatings are being
performed at present.
[1] D. Zhang et al., Fusion Sci. Technol. 60 (2011) 1576–1579.
[2] T. Tanaka, T. Muroga, Journal of Nuclear Materials 455 (2014) 630–634.

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P1-32
Investigation of the selective heating of the metal powder with hydrogen in
the molten salts using by microwave
S. Takayama, J. Yagi, T. Tanaka, A. Sagara
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

In this paper, we will discuss with experiments the selective heating behaver of the molten salts
and metal powders absorbed hydrogen in a single-mode cavity at 2.45GHz microwave. The newdeveloped experimental device has the structure that two single-mode cavity intersected
perpendicularly. In the crossing of two single-mode cavities, one becomes in the maximum
electric field of microwave, and another becomes in the maximum magnetic field of microwave.
For that reason, the sample at the crossing can control an inputting electric field and magnetic
field separately.
In FFHR design studies, the molten-salt Flinabe (LiF-NaF-BeF2) is the first candidate for aself-cooled liquid breeding blanket due to its superior safety (e.g., low MHD pressure loss, low
reactivity with air, low-pressure operation, and low hydrogen solubility) [1]. In particular, lowhydrogen solubility is advantageous for keeping the tritium inventory low and for simplifying
the tritium recovery system. However, due to the high equilibrium pressure of tritium in theliquid (a few kPa at 600◦C for 3 GW fusion), the following two aspects have been major issues:
reduction of tritium permeation through a heat exchanger, and the use of low radio-activation
vanadium alloys that have high mechanical strength up to and over 700◦C but high hydrogensolubility (i.e., hydrogen storage metal). In order to resolve those two issues simultaneously,
Sagara proposed to mix hydrogen-absorption metal powders (e.g., Ti, Zr, and V) in molten salt
to effectively increase hydrogen solubility, where the MHD pressure loss is negligible, and also
proposed to use selective heating with microwave in order to recover hydrogen from powders[1]. Yagi confirmed experimentally that a hydrogen-soluble Ti powder mixed with FLiNaK can
increase the effective hydrogen solubility for more than 5 orders [2].
On the other, microwave has characteristics of the selective heating, internal heating,volumetric heating and rapid heating. For the reasons, since 1960th, we have been successfully
applied microwave energies for cooking foods, drying woods, heating rubbers, medicines andtextiles, sintering of ceramics etc. During the last decade, the use of microwave sintering
technology for production of sintered composite materials with components of high electric
conductivity such as metals has intensively been investigated. It is well known that metallic
materials reflect microwaves rather than absorb them. However, it has been demonstrated thatsintering of metal powder compacts by using microwave heating is possible. Systematic
experimental investigations on microwave sintering of metal powder compacts in a single mode
microwave applicator operating at 2.45 GHz revealed that heating may strongly depend on thesample position in the electromagnetic mode. In addition, selective heating occurs in different
absorption ratios of materials each other. In selective heating, the local temperature difference ismade in materials and the thermally non-equilibrium state occurs in macro-state. We expect the
selective heating of metal powders in molten salts using by the magnetic field of microwave.
[1] A. Sagara, H. Tamura, T. Tanaka, N. Yanagi, J. Miyazawa, T. Goto, R. Sakamoto, J. Yagi,
T. Watanabe, S. Takayama, the FFHR design group, Fusion Engineering and Design, 89(2014) 2114-2120[2] J. Yagi, A. Sagara, T. Watanabe, T. Tanaka, S. Takayama, T. Muroga, Fusion Engineeringand Design, in press- 88 -

P1-33
Performance evaluation of a three-surface-multi-layered channel
with a reinforced metal layer
D. Isshikia, S. Itoa, T. Murogab, H. Hashizumeaa

Tohoku University, 6-6-02-1 Aramaki Aza Aoba, Aoba-ku, Sendai 980-8579, Japanb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japandisshi@karma.qse.tohoku.ac.jp

Reducing MHD pressure drop is one of the most important R&D issues in liquid metal fusion
blankets. A three-surface-multi-layered channel which is a rectangular channel and consists ofa thin metal layer and a ceramic insulating layer on three of the four inner surfaces of thechannel, is an advanced concept to reduce the MHD pressure drop [1]. The three-surface-multilayered channel reduces the MHD pressure drop to an acceptable value if the thickness of theinner metal layer is less than 0.02 mm. However, this thickness is too small to sustain strengthof the inner metal layer. To solve this problem, we have proposed a new design for the channel
in which the inner metal layer is partially reinforced by attaching reinforcing structures asshown in Fig. 1 [2]. The study demonstrated that the attached reinforcing structure not only
increases rigidity but also increases pressure drop which, however, can change flow field andenhance heat transfer. Therefore, this study evaluates the effect of the geometry of reinforcing
structures on the pressure drop, rigidity of inner metal layer, and heat transfer performance by
3D finite element analysis depending on geometry of the reinforcing structure. And then we
will discuss optimized geometry of the inner metal layer. Additionally, we fabricate prototypeof the channel.
Fig. 2 illustrates numerical model used in this study to consider a reinforced three-surfacemulti-layered channel. In this figure, the width, thickness, and the angle of the reinforcing
structures are denoted as w, h, and . Navier-Stokes equation with Lorentz force and charge
conservation equation are solved to evaluate pressure drop under conditions of a magnetic fieldof 1 T and an average velocity of 0.10 m/s. To evaluate rigidity of inner metal layer, weconsider it as cantilever and evaluate the maximum displacement. We also analyze temperaturefield at inflow condition of a heat flux of 1 MW/m2 from the first wall, and obtain the maximum
temperature to discuss heat transfer performance.
As a result, when w = 2.0 mm, h = 0.140 mm,  = 34 degree, the pressure drop is only 1.16 times
larger but the rigidity is 128 times larger than an unreinforced channel. Detail of results
including heat transfer performance and fabrication of the prototype will be presented in the full
paper.

Fig. 1 Reinforced three-surface-multiFig. 2 Assumed geometry of the reinforcing
layered channel
structure attached inner metal layer[1] H. Hashizume, Fusion Engineering and Design, Vol. 81, (2006), 1431-1438.
[2] Y. Inage et al., Proceedings of NTHAS8, (2012), N8P1107.

- 89 -

P1-34
Adhesion strength of Er2O3/buffer stacked MOCVD coating for advanced
breeding blanket
Y. Hishinuma, T. Tanaka, Y. Tanakaa, K. Matsudaa T. Muroga and A. Sagara
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japana
University of Toyama, 3190, Gofuku, Toyama 930-8555, JapanThe electrical insulator and tritium permeation barrier are key components to realize liquid
metal (Li or Pb-Li) and molten-salt typed breeding blanket systems. We investigated the Er2O3
coating as the one of the candidate material for the electrical insulator and tritium permeation
barrier. Because Er2O3 coating has higher electrical resistivity and barrier effect of hydrogen
permeation compared with the other oxide materials. Recently, we succeeded to form thehomogeneous Er2O3 coating on the large area, for example the interior surface of the 600 mm
length stainless tube, using MOCVD process [1]. Furthermore, the double-stack structured
Er2O3 coating layer with oxide buffer material was also easily formed using MOCVD process.
It is well known that the crystal growth direction can be controlled by the crystal orientation of
the substrate. Generally, the physical soundness of coating material was evaluated by the
hardness, abrasion and adhesion properties. The scratch test is often used to evaluate theadhesion strength of hard coating and car paint layer. The schematic image of the scratchmechanism is shown in fig.1. The coating material is actually broken by the diamond scratch
stylus, and the adhesion strength is obtained by the shearing stress on the interface of coating
and substrate shown in fig.2.
In this study, we carried out the scratch test to the single and double-stacked coating, and theadhesion strength comparisons between single and double-stacked coating were investigated asfunctions of thickness and crystallinity of Er2O3 coating. Furthermore, the thermal cycling up to973 K in Ar atmosphere was also carried out, and the changes of microstructure and adhesion
strength will be reported.
Acknowledgement
This study was mainly performed and supported by the Fusion Engineering Project in NIFS
(UFFF026)[1] Y. Hishinuma, T. Tanaka, T. Tanaka, T. Nagasaka, Y. Tasaki, A. Sagara and T. Muroga,
Fusion Engineering and Design, 86, 2011, 2530.

Fig.2 The configuration and interaction between
the diamond scratch stylus and surface of thecoating layer

Fig.1 The schematic image of the scratchmechanism on the scratch test.

- 90 -

P1-35
Proposal of the maintenance method of breeder blankets of LHD-type
helical fusion reactor FFHR
T. Goto, J. Miyazawa, H. Tamura, T. Tanaka, N. Yanagi, A. Sagara and the FFHR Design
Group
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Conceptual design activity of LHD-type helical fusion reactor FFHR has been advanced under
the Fusion Engineering Research Project in NIFS. One of the important design issues is themethod of construction and maintenance of the in-vessel components with a three-dimensional
shape. In particular, breeder blankets should be replaced periodically due to the neutron damageof the structural material. To enable the replacement of the breeder blankets with a short time,
module structure of the breeder blankets should be designed with a consideration of themaintenance method.
In the latest conceptual design of FFHR, molten salt (FLiBe or FLiNaBe) is considered as abreeder material. Because electric conductivity of the molten salt is significantly lower than thatof other liquid breeder candidates (liquid metals), flexible design of the flow channels ofbreeder blankets can be adopted. Therefore, we proposed Toroidally-Sliced, HELically-Linked(T-SHELL) breeder blanket system. The helical shape breeder blanket is cut by the plane ofconstant toroidal angle. By considering the mass of each module, division per 2-3 degree isassumed. With such a narrow range of the toroidal angle, the degree of torsion of the blanket is
not so large and the shape of each module is similar to the bananas-like blanket modules oftokamak reactors. Because large areas between superconducting coils of FFHR can be used as aport structure, most of these toroidally-sliced modules can be extracted by a simple movementwith radial and vertical directions. In the presentation, the detailed shape of each module isgiven and a baseline replacement method will be discussed.

- 91 -

P1-36
Observations of gas stream from fusion test device
for the design of exhaust detritiation system
M. Tanakaa,b,, N. Suzukia, H. Katoa, T. Kondoa and LHD Upgrade Teama. National Institute for Fusion Scienceb. SOKENDAI (The Graduate University for Advanced Studies)322-6 Oroshi-cho, Toki 509-5292, Japane-mail address: tanaka.masahiro@nifs.ac.jp

The development of nuclear fusion system is remarkably progressed in recent years. Owing toadvance of the plasma performance, the deuterium gas is used as the operation gas in a fusion
test device. Under the deuterium plasma experiments by large fusion test device equipped with
NBI heating systems, a small amount of tritium is produced by D-D reactions in the vacuum
vessel. Then, a part of tritium is exhausted from the vacuum vessel and the rest is retained in thefirst wall and the diverter tiles. From the viewpoints of radiation protection and public
acceptance, exhausted tritium from the vacuum vessel may be recovered by an exhaust
detritiation system. The exhaust gas streams are classified into: (1) plasma discharge operation;(2) the regeneration of cryopumps and ice pellet injection systems; (3) rough pumping for main
and NBI vacuum vessels; (4) a purge gas in the vacuum vessel during the maintenance. For
example, the plasma discharge for a few seconds is operated intermittently with a cycle per
several minutes. Thus, the average gas flow rate would be very low and the concentration of
gaseous hydrogen isotopes in the exhaust gas may be higher. On the other hands, the cryopump
systems would be regenerated with the purging of nitrogen gas over a period of hours once a
week. Thus, the gas flow rate under the regeneration operation and the total gas volume would
be much larger than that under the plasma discharge operation. In the case of rough pumping,
the flow rate is large but transitory. To design of the exhaust detritiation system, therefore, thebehavior of exhaust gas stream from large fusion test device under each operation has to beunderstood quantitatively. In this study, we observed the gas stream from the large fusion test
device under the various operation conditions such as plasma discharges, the regeneration of
cyopump system and the rough pumping of the vacuum vessels. Then, the conceptual design of
exhaust detritiation system, which has enough capacity to treat the various gas streams, will beshown.

- 92 -

P1-37Development of Soft-X Ray Computed Tomography Diagnostic System
for MHD Dynamics Study in RELAX RFP
Y. Aoki, A. Sanpei, H. Makizawa, M. Higuchi, T. Nagano, S. Nakanobo, R. Tsuboi,
H. Himura, S. Masamune
Kyoto Institute of Technology, Department of Electronics, Kyoto 606-8585, Japanaoki13@nuclear.es.kit.ac.jp

The reversed-field pinch (RFP) is one of the magnetic confinement systems for high-betafusion plasmas. One of the topics is a spontaneous transition of RFP configuration fromaxisymmetric to non-axisymmetric helical structure dubbed quasi-single helicity (QSH) state.
In this QSH state, improved confinement was realized in the helical core, either inside a large
magnetic island or helically deformed nested flux surfaces having helical axis, with increased
electron temperature and/or higher density [1].
RELAX is an RFP machine with major radius R = 0.5 m and the minor radius a = 0.25 m.
The aspect ratio A = 2 in RELAX, whose research objectives include geometrical optimizationof the RFP configuration [2]. Generally speaking, the q value on axis increases as the aspectratio is lowered, while the edge value does not change so much because the field reversal
surface (q = 0) is close to the edge. Therefore, if we could avoid the resonance near the core,
relatively large space is available in the core region without major resonance; the single islandof a core resonant tearing mode can grow without interacting neighboring islands associatedwith the tearing modes. This picture has been demonstrated experimentally in RELAX in whichtransition to QSH occurs at lower current density and lower Lundquist number than in othermachines such as RFX-mod and MST [3].
When the plasma is in equilibrium, the Soft-X ray (SXR) emissivity contours coincidewith magnetic flux surfaces because both the electron temperature and density can be regarded
as surface quantities. Therefore, the SXR computed tomography (CT) is a powerful tool toidentify the thermal or magnetic structure inside hightemperature plasmas. We are developing a new SXR CT
system consisting of two 20-channel photodiode arrays.
The cross-sectional view of the diagnostic system is shownin Fig.1 where two arrays are set at the top and bottom at asingle toroidal location. Our interest includes time
evolution of the SXR emissivity contours at transition toand back-transition from the QSH in RELAX; the process
was studied by magnetic diagnostics and SXR camera, thelatter providing SXR images with 10-μs time resolution.
We expect to obtain time evolution of the magnetic flux
surfaces with 1-μs time resolution (typical sampling time), Fig.1 Poloidal cross-section ofwhich would be useful to study the dynamic processes of RELAX with lines of sighttransition or self-organization associated with QSH.
covered by two photodiodes.
[1] D. Escande, “What is a reversed field pinch?” 2013 <hal-00909102>.
[2] S. Masamune et al., J. Phys. Soc. Jpn. 76, 123501 (2007).
[3] K. Oki et al., PFR 7, 1402028 (2012); Fusion Sci. Technol. 63, 1T, 386 (2013).

- 93 -

P1-38Studies on the effect of radio frequency field
in a cusp-type charge separation device for direct energy conversion
M. Hamabe, H. Izawa, H. Takeno, S. Nakamoto, K. Ichimuraa, Y. Nakashimaa
Department of Electrical and Electronic Engineering, Kobe University,
1-1, Rokkodai-cho, Nada-ku, Kobe 657-8501, Japana
Plasma Research Center, University of Tsukuba, 1-1-1, Tennodai, Tsukuba 305-8577, Japan

Direct energy conversion can be applied to D-3He fusion power generation in which high plant
efficiency can be expected because the majority of the produced energy in D-3He reaction iskinetic energies of charged particles. This power generation method is also advantageous in theenvironmental aspects because harmful high energy neutrons are hardly generated. For directenergy conversion, charged particles escaping from the reactor should be separated from eachother in energy and electrical polarity. For this purpose, application of a cusp-type direct energy
converter (CuspDEC) is expected to be used [1].
Previous studies show that cusp magnetic field is effective for particle separation to low density
plasma [2]. Although the effectiveness was also examined to high density plasma, separation
performance was insufficient for high efficiency direct energy conversion and some methods toimprove separation performance were required. As one of these methods, using a radio
frequency (rf) electric field was proposed. The ponderomotive force is well known as a
non-linear effect of high frequency electric field [3], and is expressed by
FNL  (p /  ) 2   0 E 2 / 2 , where ω is the angular frequency of rf, E 2 is time-averaged squareof the rf electric field, and other symbols are as in Ref. 3. This force influences electrons muchstronger than ions. An rf field with an appropriate distribution may have a function to assist
charged separation in a CuspDEC.
In this study, we experimentally examine the effect of rf field in a CuspDEC simulator
following to the previous reports [4,5]. Two ring-type electrodes were newly installed in thesimulator, and electrostatic rf field was formed near the electron collector and the collector
current was measured. The collector current increased or decreased according to theenhancement of field. We could not find difference of the variation to the change in rf frequency
although the strength of the ponderomotive force depends on rf frequency as shown in theequation above. On one hand, we found clear difference of the variation to the change inmagnetic field strength. We also found some differences on the field curvature and on therelative phase difference between rf voltages of the two electrodes. The detail of theexperimental results with discussions will be presented in the conference.
This work is partially supported in part by a Grant-in-Aid for Scientific Research (C)(25420255) from Japan Society for the Promotion of Science and by the bilateral coordinatedresearch program (NIFS13KUGM082) between PRC, University of Tsukuba, NIFS, and Kobe
University.
[1] H. Momota, et al., Proc. 14th IAEA 3 (1993) 319.
[2] Y. Yasaka, et al., Nucl. Fusion 49 (2009) 075009.
[3] F. F. Chen, Introduction to Plasma Physics and Controlled Fusion, 2nd Ed. 1, Plenum Press(1984) 305.
[4] H. Takeno, et al., 23rd ITC (2013) P1-53.
[5] H. Takeno, et al., PLASMA2014 (2014) 18PB-031.

- 94 -

P1-39Studies on the axial position of the decelerator
in traveling wave direct energy converter
T. Katsura, T. Wakaizumi, H. Takeno, K. Ichimuraa, Y. Nakashimaa
Department of Electrical and Electronic Engineering, Kobe University,
1-1, Rokkodai-cho, Nada-ku, Kobe 657-8501, Japana
Plasma Research Center, University of Tsukuba, 1-1-1, Tennodai, Tsukuba 305-8577, JapanIn D-3He fusion, most of fusion energy is carried by kinetic energy of created protons, so directenergy conversion can be applied. Traveling wave direct energy converter (TWDEC) was
proposed as an energy recovering system for these protons [1], which was composed of amodulator and a decelerator. Incident protons are velocity modulated by radio frequency (RF)field in the modulator and are density modulated in the downstream. Then, they are decelerated
by traveling RF field and their kinetic energy is converted into electricity in the decelerator.
The axial position of the decelerator is one of the important device parameters for energy
conversion efficiency and device size. The best position for conversion efficiency is considered
to be the bunching position in which ion density is the highest. In the previous study, the authors
conducted numerical calculations and simulation experiments by setting a decelerator at thebunching position [2]. The highest experimental record of the efficiency was obtained, however,
the optimum axial position was not confirmed experimentally.
In this study, we investigate the dependence of deceleration efficiency on the axial position of
the decelerator. By using numerical calculations and simulation experiments, we examine theefficiency by changing the distance between the modulator and the decelerator (L), and discuss
the optimum position of the decelerator based on the results.
We examined the dependence by numericalcalculations. Figure 1 shows deceleration
efficiency versus L. The conditions are thesame as those in the previous simulation
experiment. As shown in the figure, thehighest deceleration efficiency of 61.5 % is
obtained for L = 22 cm, and this is muchhigher than 55 % for the bunching length of L
= 11.9 cm. Therefore, the bunching position
is not necessarily the optimum position of thedecelerator. This may be because that thedecelerator is designed based on the constantdeceleration scheme [3]. The corresponding
simulation experiments are in preparation.
Fig. 1 Deceleration Efficiency versus distance
We will present the results and discussion in
between the modulator and the decelerator.
the conference.
This work is supported in part by the bilateral coordinated research program(NIFS13KUGM082) between PRC, University of Tsukuba, NIFS, and Kobe University..
[1] H. Momota, LA-11808-C, Los Alamos Natl. Lab. (1990) 8.
[2] Y. Togo, et al., 24th ITC (2014) P6-33.
[3] Y. Togo, et al., Plasma and Fusion Research 10 (2015) 345013.

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P1-40
Improvement of charge separation by increase of magnets
in a PM-CuspDEC simulator
T. Watanabe, Y. Kitahara, H. Takeno, S. Nakamoto, K. Ichimuraa, Y. Nakashimaa
Department of Electrical and Electronic Engineering, Kobe University,
1-1, Rokkodai-cho, Nada-ku, Kobe 657-8501, Japana
Plasma Research Center, University of Tsukuba, 1-1-1, Tennodai, Tsukuba 305-8577, Japan

Cusp-type direct energy converter (CuspDEC) was proposed as a charge separation device for
D-3He fusion power generation [1]. It separates charged particles based on the difference ofLarmor radii. The authors proposed an application of this function to divertor plasma to reduce
thermal loads on a divertor plate [2]. For the application to divertor plasma, the device should
be miniaturized as the divertor region is not so wide. The authors proposed use of permanent
magnets (PMs) and composition of CuspDEC with them (PM-CuspDEC) [2].
Some core units of PM-CuspDEC were examined experimentally, showing a good performance[3]. Based on the results, a PM-CuspDEC simulator, composed of two magnetic coils and two
cylindrical hollow ferrite magnets, was constructed, which was assumed to be applicable to
GAMMA 10. By using a test plasma source, the charge separation performance of the simulator
was examined, however, it was not sufficient [4]. In this paper, an improvement of theperformance of the simulator is treated by increasing the number of PMs.
Figure 1 shows results of numerically calculated electron orbits. Cross sections of two magnetic
coils and PMs are shown in r-z plane. Thin curves are field lines, and electrons are expected to
be deflected in the cusp field in the downstream of thePMs although ions go through the region. Orbits ofelectrons incident at (r,z) = (0.1 cm, -20 cm) are
shown by thick (magenta) curves.
In Fig. 1(a) and (b), two and three PMs are installed
corresponding to the conditions of the former
simulator and an improved one, respectively. The
incident energy of the electron (Ein) also differs: 14
eV and 20 eV for (a) and (b), respectively. Although
the electron in (a) goes straight, that in (b) is deflectedin spite of its higher energy. In other calculations,electrons with much higher Ein are also deflected
when four PMs are installed, which is preferable for(a) Ein = 14 eV
efficient separation. An increase of PMs would
improve separation performance of the simulator.
The corresponding experimental results anddiscussion will be presented in the conference.
This work is supported partially by the NIFS
collaboration research program (NIFS13KUGM082),
and by a Grant-in-Aid for Scientific Research (C)(25420255) from JSPS.
[1] Y. Tomita, et al., NIFS Series (1993) NIFS-252.
[2] H. Takeno, et al., Trans. Fusion Sci. Tech. 63(2013) 131.
[3] K. Nishimura, et al., 9th APFA (2013) TP-33.
[4] Y. Tomita, et al., 24th ITC (2014) P6-32.

- 96 -(b) Ein = 20 eV
Fig. 1 Orbits of electron.

P1-41Development of high strength W/V/Au/ODS-Cu joint using HIP process
H.Notoa, T.Yamadab, Y.Hishinumaa, T.MurogaaaNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanb
Nagoya University, Graduate School of Engineering, 509-5292 Gifu, Japane-mail address of submitting author:noto.hiroyuki@nifs.ac.jp

The divertor with higher heat removal capabilities and heat resistance to severe heat andparticle load are one of the key components to operate fusion reactors. In the designing ofhelical DEMO reactor FFHR-d1, the steady thermal load is assumed to exceed 20 MW/m2. Oneof the candidate material combination for fabricating the divertor is tungsten (W) and copper(Cu) alloy because W and Cu have promising properties for the divertor application. W has highmelting point, low hydrogen isotope retention, low sputtering yield, and so on. Cu alloy has
higher thermal conductively compared with the other metal materials. In fact, copper alloy(Cu-Cr-Zr) pipes are tried to join with W by conventional brazing on the ITER project.
However, tough ness of the copper alloy drastically decreased at a high temperature over 450 ˚Cdue to instability of strengthening precipitated particle. On the other hand, oxide
dispersion-strengthened copper alloys (ODS-Cu) such as GlidCop® (Cu-0.3wt%Al2O3) has
superior mechanical properties at high temperatures than Cu-Cr-Zr because the metal has
thermally-stable oxide particle, and thus can be an advanced candidate for the divertor
components.
Therefore, the establishment of the bonding process between W and ODS-Cu is an essential
technology to succeed the divertor system using ODS-Cu. W and Cu are almost insoluble witheach other and they have very different coefficient of thermal expansion (CTE). Thus, the insert
materials between W and ODS-Cu is expected to be effective to the bonding.
In this study, vanadium (V) and gold (Au) are selected as interlayers to the joint between W
and ODS-Cu. These layers have advantage for the W/ODS-Cu bonding such as high solubility,
plasticity and thermal conductivity, and low activation. V and Au have extremely different
melting points, these metal expected to use as a new bonding material based on diffusion
reaction. The new bonding proceeded with two steps. The first step is the W/V bonding bysolid-phase diffusion reaction. The second step is the W-V/Au/ODS-Cu bonding using the Hot
Isostatic Press (HIP). The HIP process condition was 950 ˚C for 1 hr in argon atmosphere of 200
MPa. The melting point in the interface between Au and ODS-Cu is decreased by the eutectic
reaction, and referred as the Transient Liquid Phase bonding.
The smooth interfaces without reaction layers and large defects were observed in the W/V and
Au/ODS-Cu interface. However, the partial-coverage by reaction layer (approx.20μm) thatdiffer from each inset metals (V,Au) wereobserved in the V/Au interface region.
The homogeneous hardness profiles
showed around the hardening area at
V/Au interface. In addition, the overallbending strength of the
W/V/Au/ODS-Cu joints using HIP
process and the depressant method of
hardening area at V/Au interface will
Figure 1. OM image of the cross section are of V/Au/GlidCop
were reported.
joint, showing hardness profile.

- 97 -

P1-42Development of Oxide Dispersion Strengthened-Copper Using MA-HIP
MethodT. Yamada a, H. Noto b, Y. Hishinuma b, H. Nakamura a, b, T. Muroga bab

Graduate School of Engineering, Nagoya University, 509-5292, Japan
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
yamada-tetsuya14@ees.nagoya-u.ac.jp

The divertors of fusion reactors are subject to a high thermal heat load. In the helical
DEMO reactor FFHR, the steady thermal load is assumed to exceed 20 MW/m2. Thus
development of heat sink materials having a high heat removal capability is required.
For this purpose, copper alloys have a large advantage in thermal conductivity.
Although copper alloys are susceptible to neutron irradiation and thus were notconsidered as DEMO candidates, the reduced neutron fluence in the recent FFHR
design has led to revisiting them as heat sink materials. In ITER, use of precipitation
strengthened (PS)-Cu (CuCrZr) is being considered. However PS-Cu has issues suchas instability of microstructure at high temperatures. On the other hand, dispersion
strengthened (DS)-Cu is known to have higher stability in microstructure and strength
at high temperature. The conventional DS-Cu has been produced by internal oxidationand extrusion, which, however, can cause inhomogeneity and coarsening of dispersed
particles and formation of anisotropic microstructure.
In this study, we research into improvements of DS-Cu using Mechanical Alloying(MA) followed by Hot Isostatic Pressing (HIP) method. MA can disperse particles
finely and homogeneously, and HIP can form dense microstructure. In this study,
effects of additive elements (Al, Zr) and MA conditions (rotation speed and duration)
are investigated. Cu-1wt.% Al (Cu-Al) and Cu-1wt.% Zr (Cu-Zr) are alloyed using
planetary ball mill, with rotation rates of 50, 200 and 250 rpm for 1 to 40 hours. During
MA, the state of the powder is observed with Optical Microscopy (OM), a Scanning Electron
Microscopy (SEM) and X-Ray Diffractometer (XRD). HIP is carried out at 950C for 1 hourin an argon atmosphere of 150MPa. Vickers hardness tests are also carried out.
Examination of particles after MA shows thatthe original particles start to combine with eachother around 200-250 rpm for 1 hour, which isconsidered to be the starting point of alloying.
An examination with XRD shows the peak of Al,
which is present before the MA treatment,
disappear after the MA treatment. HIPedmaterials following MA show reduced grain size
and increased hardness relative to those of theunalloyed copper HIPed in the same condition.
Large precipitates or inclusions are not observed
either in matrix or on grain boundaries.
The present study has demonstrated that MA-HIP of Cu with Al or Zr addition can
produce homogeneous structure with refined grains.

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P1-43
Real-Time Display System of the Equilibrium Reconstruction for LongTime Discharge in QUEST
M. M. Alam, K. Nakamuraa, M. Hasegawaa, F. Xiab, O. Mitaraic, K. Tokunagaa,
K. Arakia, H. Zushia, K. Hanadaa, A. Fujisawaa, H. Ideia, Y. Nagashimaa,
S. Kawasakia, H. Nakashimaa, A. Higashijimaa, T. Nagataa
IGSES, Kyushu University, 6-1 Kasuga-koen, Kasuga, Fukuoka, 816-8580 Japan
RIAM, Kyushu University, 6-1 Kasuga-koen, Kasuga, Fukuoka, 816-8580 Japanb
CFS, SWIP, P.O. Box 432, 610041 Chengdu, P. R. Chinac
Tokai University, Kumamoto, 862-8652 Japana

Cost effective, efficient, precise and faster real-time display system design and development has
become an important issue of QUEST (Q-shu University Experiments with Steady-State
Spherical Tokamak) for advanced operations. For the construction of real time display system,
data acquisition is one of the most important issues. Most of the cost is required for theconstruction of the real-time data acquisition system. But due to having the promise of costeffectiveness for the construction of real-time display system, data acquisition has been carriedout from reflective memory. In QUEST, it is reported that a distributed feedback control system
has been developed with a reflective memory [1]. So for the design and develop of the real-time
display system of the equilibrium reconstruction for long time discharge in QUEST, thereflective memory of the distributed control system will be used.
For the equilibrium reconstruction, CCS (Cauchy-Condition Surface) method will be used. The
CCS method has suitable precision for equilibrium reconstruction in conventional tokamak.
But for spherical tokamak good precision can be obtained by CCS method if we solve theinverse problem of flux loop measurement by using least square method [2]. On the other handto remove noise we can introduce singular value decomposition on CCS method so that the
CCS method can reproduce plasma boundary shape more precisely and faster for the real-timesystem by using the data from magnetic sensors [3]. The plasma equilibrium reconstruction
parameters such as position, minor radius, elongation and triangularity will be calculated byusing this CCS method and will be carried out in real-time operating system (LabVIEW). The
RTOS (Real-time Operating System) will ensure precise timing and high reliability. Finally, thereal-time data profiles and plasma shape will be displayed in real-time I/O and third party PC.

[1] Makoto Hasegawa et al., Fusion Eng. and Des. 95 (2015).
[2] K. Nakamura et al., J. Plasma Fusion Res. SERIES, 9 (2008).
[3] K. Nakamura et al., IEEE Trans. on Plasma Sci. 42 (9) (2014).

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P1-44
W7-X ECRH plant - Control System Design Features
H. Braune1, V. Erckmann1, H. Laqua1, S. Marsen1, F. Noke1, F. Purps1, N. Schneider1,T. Schulz1, T. Stange1, P. Uhren1 and the W7-X ECRH teams at IPP, IGVP Stuttgart and KIT
Karlsruhe
1

Max-Planck-Institut für Plasmaphysik, Garching and Greifswald, Germany

Electron Cyclotron Resonance Heating (ECRH) is the main heating system for steady-state
operation and will be the only available heating system during the first operation phases of thestellarator W7-X [1]. The ECRH is going to be used for different tasks at the W7-X such as wallconditioning, controlled plasma start-up from the neutral gas up to steady state plasma control[2] and performance optimization by plasma profile shaping. The ECRH plant is equipped with
10 gyrotrons with a frequency of 140 GHz and a microwave output power of 1 MW each [1].
The microwave beams are guided quasi optical via cupper coated mirrors toward the W7-X
vessel. The prospective W7-X operation requires an ECRH system as a whole, only controlled
by some high level parameters such as heating power, launch direction, polarisation and timing.
The main challenges and developed solutions will be discussed.

Challenges and Requirements
Up to 10 gyrotrons have to be controlled simultaneously as well as 10 different wave guides andtwo additional remote steering launchers apart from the 10 ordinary front steering launchers.
Each gyrotron is equipped with a superconducting magnet, high voltage power supplies,
collector sweep systems, different cooling circuits filled with water as well as with oil. Response
times between 1 µs and 100 ms are necessary in order to control the different auxiliaries reliable.
The control system has to provide the interface between the ECRH-plant and the W7-X main
control system as well as the W7-X data acquisition system.
The support of error diagnostics for each ECRH component part is a further mandatory control
system function.

Solutions
The human-machine interface is based on the Siemens WinCC software and has been redesigned
in order to meet the W7-X requirements. Each part of the ECRH installation such as magnet,
high voltage power supplies, water cooling system etc. is controlled separately by Siemens S7
PLC [3] and the controller software has been completed with several functions that relive the
ECRH operators from many periodical tasks.
The ECRH remote control system has been connected to the central W7-X control system in a
reliable and easily maintainable way.
An ECRH fast interlock system monitors all devices within timescales less than 5 µs, which areable to achieve a dangerous state in particular arcs and gyrotron failures.
[1] V. Erckmann et al. "ECRH and W7-X – an intriguing pair" Radio Frequency Power in
Plasmas, Proc. 20 Topical Conference, Sorrento, Italy, AIP Conference Proceedings 1580
(2013) page 542[2] M. Preynas et al. "Characterization of plasma start-up using ECRH in preparation of
Wendelstein 7-X operation" Conference Proceedings 18th Joint Workshop on ECE and
ECRH, 22-25.04.2014,Nara, Japan[3] H. Braune et al. "Architecture of central control system for the 10 MW ECRH-plant at
W7-X" Fusion Engineering and Design 82 (2007) 677–685- 100 -

P1-45Experimental Study on W-Band Oversized Surface Wave Oscillator Driven
by Weakly Relativistic Electron BeamMin Thu San, K. Ogura, K. Yambe, Y. Annaka, S. Gong, J. Kawamura, T. Miura,
S. Kuboa, T. Shimozumaa, S. Kobayashia, K. Okadaa
Graduate School of Science and Technology, Niigata University, Niigata 950-2181, Japana
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan

High power microwave sources utilizing high current relativistic electron beam such as
backward wave oscillator consist of slow-wave structure (SWS). The SWS reduces the phasevelocity of electromagnetic waves below the vacuum speed of light so that the electrons in therelativistic beam can interact with the wave. To increase the operation frequency, SWS isoversized with the diameter larger than the free-space wavelength of output electromagnetic
wave by several times or more. And microwave generators with oversized SWS have attracted
much attention as a compact terahertz wave source [1], [2]. In oversized SWS cases, the slow
wave becomes a surface wave concentrated on the SWS wall. And the interaction may result inbackward wave or travelling wave operation depending on the group velocity of the surface
wave. At high electron energy, the group velocity of this wave is positive and the device
operates in the manner of a travelling wave tube (TWT), at low electron energy, the groupvelocity is negative and operates like a backward wave oscillator (BWO). And at anintermediate energy, called the Bragg condition or π point, the group velocity of the evanescentwave vanishes. The overall operation from BWO to TWT has been studied theoretically and it
has been shown that there is no operation near the π point [3]. Up to now, the theoreticalprediction has not been validated against experimental results.
In this work, we studied a W-band oversized surface wave oscillator driven by weakly
relativistic electron beams of less than 100kV. The overall operations from BWO to TWT
including π point operation were studied experimentally. Rectangular corrugations are used as
the SWS having surface waves with upper cutoff frequency about 100GHz. Parameters of
SWS: average radius R0, corrugation amplitude h, corrugation width d and periodic length z0,
determine dispersion characteristics of the structure. To get the upper cutoff frequency about
100GHz, the structure parameters of R0 = 15.1mm, h = 0.3mm, d = 0.3mm and z0 = 0.5mm areused. Uniformly distributed annular electron beams generated by a disk-type cold cathode are
injected into the oscillator. In the study, the operations of the oscillator were investigatedexperimentally by injecting the electron beams into various structure lengths of 40, 80 and 120
periods. Driven electron beam energies are of 10~80keV. The operations of the oscillator
depend on the energy level and the structure length. The experiments showed that the operationsof the oscillator are mainly concerned with the structure length.

[1] V. L. Bratman, B. S. Dumesh, A. E. Fedotov, P. B. Makhaliv, B. Z. Movshevich, and F. S.
Rusin, IEEE Trans. Plasma Sci. 38 (2010) 1466.
[2] S. Gong, K. Ogura, S. Nomizu, K. Yamazaki, K. Yambe, S. Kubo, T. Shimozuma, S.
Kobayashi, and K. Okada et al., accepted for publication in IEEE Trans. Plasma Sci.
[3] H. L. Andrews, C. H. Boulware, C. A. Brau, and J. D. Jarvis, Phys. Rev. ST Accel. Beams 8,(2005) 050703.

- 101 -

P1-46
Experimental Study on Slow Cyclotron Interaction under Anomalous
Doppler Effect in W-Band Backward Wave Oscillator
Y. Annaka, K. Ogura, J. Kawamura, T. Miura, T. Kawamura, S. Takanashi,
S. Gong, Min Thu San, and K. Yambe
Graduate School of Science and Technology, Niigata University, Niigata 950-2181, Japan

Backward Wave Oscillator (BWO) has been studied as one of high power microwave
sources. BWO has a slow-wave structure (SWS) to reduce the phase velocity of microwave tothe beam velocity and is driven by axially streaming electron beams. The electron beams arepropagating along the direction of an axial magnetic field and have four beam modes: slow
and fast space charge modes and slow and fast cyclotron modes [1]. The microwave radiation
may occur near the intersection point between the slow beam modes and slow-wave modes in
SWS. In this paper, we refer the intersection of slow space charge mode to “Cherenkov
interaction” and that of slow cyclotron mode to “slow cyclotron interaction”. The latter occurs
at the anomalous Doppler shifted cyclotron frequency [2]-[4]. The resonance condition(). Here,between the slow wave and the beam is given by
and k are respectively the angular frequency and the wave number of slow wave,
is thevelocity of electron along the magnetic field,
is cyclotron angular frequency, and S isharmonic number. At, Cherenkov interaction takes place. At
slow
cyclotron interaction occurs due to the anomalous Doppler effect. It has been pointed out by
numerical analyses that the slow cyclotron interaction provides high efficiency operations andthe potential for large frequency up-shift with moderate magnetic field and beam energy.
In this work, we examine experimentally microwave generations by the slow cyclotroninteraction in a W-band BWO. We use rectangularly corrugated waveguide as SWS withcorrugation amplitude, period lengthand corrugation width. The upper cutoff frequency of surface wave is about 100 GHz. We prepare two
SWSs. They have different periods of 40 and 80. We also compose 120 periods SWS of 40
and 80 periods SWSs. BWOs with various length SWS are driven by non-relativistic beams
less than 30 keV. To help electrons go through SWS, the magnetic field of 0.8 T is provided
by 10 solenoid coils. In this magnetic field, electron cyclotron frequency is 22.9 GHz.
Frequency of electromagnetic field can be predicted from the resonance condition anddispersion characteristics of SWS. The microwave from BWOs is detected by two detect
systems, F-band and D-band. The cutoff frequency of F-band is about 74 GHz. D-band isabout 91GHz. We investigate the slow cyclotron interaction in BWOs based on the frequency
of detected microwave by comparing numerically obtained dispersion characteristics andresonance conditions.

[1] K. Ogura, K. Yambe, K. Yamamoto, and Y. Kobari, IEEE Trans. Plasma Sci. 41 (2013)
pp.2729-2734.
[2] T. H. Kho and A. T. Lin, Phys. Rev. A, 38 (1988) pp.2883-2888.
[3] K. Ogura, M. R. Amin, K. Minami, X. D. Zheng, Y. Suzuki, and W. S. Kim, Phys. Rev. E,
53 (1996) pp.2726-2729.
[4] G. S. Nusinovich, M. Korol, and E. Jerby, Phys. Rev. E, 59 (1999) pp.2311-2321.

- 102 -

P1-47
Depth of Influence on the Plasma by Beam Extraction in a Negative
Hydrogen Ion Source for NBI
Shaofei Geng, Katsuyoshi Tsumoria, Haruhisa Nakanoa, Masashi Kisakia, Katsunori Ikedaa,
Masaki Osakabea, Ken-ichi Nagaokaa, Yasuhiko Takeiria, Osamu Kanekoa, Masayuki Shibuyaa
SOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi, Toki, Gifu 509-5292,
Japana)National Institutes for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 Japan

Experimental results obtained with single-Langmuir probe and laser photodetachment
measurements in beam extraction region of a cesium-seeded negative ion source for NBI are
described. It is unclear how negative ions are extracted from plasmas at the extraction region,
and we investigated the response of charged particles to applied external field. The profiles ofprobe saturation currents and H- ion density in the direction normal to the plasma grid (PG)
surface are measured by scanning the tip positions. By comparing the results before and during
beam extraction, the charged particle responses due to the beam extraction are analyzed.
During beam extraction, probe saturation current increases since H- ions are extracted andelectrons flow into the extracting region for compensation. The maximum increment of theprobe saturation current due to electron flow appears at 14-24 mm apart from the PG surface.
Meanwhile, the maximum decrement of the local H- ion density is in the same region. In theregion far from the plasma grid, probe saturation current increment is low as well as thedecrement of the local H- ion density. The extrapolation of the profile suggests that the depth of
the influence on the plasma by beam extraction is around 40 mm from the plasma grid surface.

- 103 -

P1-48
Power and polarization monitor developed for ECRH and ECCD on LHD
R. Makinoa, S. Kuboa,b, K. Kobayashib, T. Ii Tsujimuraa, S. Kobayashia, T. Shimozumaa,
Y. Yoshimuraa, H. Igamia, H. Takahashia,c, T. Mutoha,caNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
Nagoya University, Nagoya 464-8603, Japanc
SOKENDAI (The Graduate University for Advanced Studies), 322 Oroshi-cho, Toki 509-5292, Japan

Polarization states of injected EC waves significantly affect the mode excitation purity,
therefore, power absorption and current drive in plasmas. A power and polarization monitor has
been developed for electron cyclotron resonance heating (ECRH) and electron cyclotron current
drive (ECCD) on the Large Helical Device (LHD).
The power and polarization monitor is installed at the miter-bend in the 77GHz ECRH
transmission line system. The monitor is composed of a miter-bend mirror with coupling holes,
orthomode transducer, heterodyne systems and ADC with FPGA (Field Programmable Gate
Array). A part of EC wave is picked-up from main corrugated waveguide to square
sub-waveguide through the coupling holes in a miter-bend. The EC waves are separated to two
orthogonal polarization by orthomode transducer. Two orthogonal waves are down-converted
to a few hundred MHz by heterodyne systems. The IF signals are detected by ADC with FPGA
(Field Programmable Gate Array). Amplitudes and phase difference of the electric field of thetwo orthogonal polarizations are obtained by applying fast Fourier transforming to the IF signal
data. Polarization angle α and polarization ellipticity β can be calculated from amplitudes andphase difference of the electric field of the two orthogonal polarizations.
To investigate the performance of the power and polarization monitor, polarization scan
experiments were performed on LHD. Amplitudes and phase difference of two orthogonalelectric field, which are needed for calculation of the polarization angle α and polarization
ellipticity β, were successfully detected. We will present the configuration of the monitor andresults of performance tests.

- 104 -

P1-49

  
  
  
  
  
  
  
withdrawal  - 105 -

P1-50Absorption Behavior of Nitrogen in Nitrogen Hot Trap
E. Wakai, H. Tanaka, Y. Ito, H. Sakasegawaa, T. Higashib, A. Suzukib, T. Teraib, J. Yagic, K.
Nakaniwa, H. Kondo, F. Groescheld, H. Tanigawaa, M. Sugimotoa, S. Ohiraa
Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195, JapanaJapan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japanb
The University of Tokyo, Tokyo 113-8654, Japanc
National Institute for Fusion Science, Toki, Gifu 509-5292, Japand
Karlsruhe Institute of Technology, Karlsruhe, Germany
e-mail address: wakai.eiichi@jaea.go.jp

The main function of the International Fusion Materials Irradiation Facility (IFMIF) is togenerate the intense fusion neutrons by injecting the deuteron beams accelerated to highenergy onto the liquid lithium flow. Guiding the lithium flow along the concave back plate ata speed of 15 m/s is required to increase the pressure in the lithium flow by centrifugal force,to avoid boiling by the heat input about 1 GW of the deuteron beams, and to remove heat by
the lithium flow circulation under lithium purification control by N hot trap, H hot trap, andcold trap. In this study, the absorption behaviour of nitrogen using N hot trap and the trapsystem are mainly evaluated in IFMIF/EVEDA project, and the fabricated nitrogen hot trapsystem is also presented.
The validation test of the N hot trap using Fe-5Ti getter under nitrogen gas atmosphere was
performed at 500oC or 450oC. Almost 70% of the introduced nitrogen gas into the N hot trap
container was seemed to be absorbed by the Fe-5Ti pebble during this test for 115 h. That is,about 18 g of nitrogen in the N hot trap container have been absorbed by the Fe-5Ti pebble of
95 kg and by another area in the N hot trap container. The surfaces of the pebbles, which were
as-received Fe-5Ti pebble and its tested one, were observed by a scanning electron
microscopy, and the distribution of nitrogen and the elements were analysed by EPMA. In thechemical analysis of nitrogen content in the specimen, the 65 wppm of nitrogen wasquantified for the Fe-5Ti pebble tested under 500oC for 115 hours. That is, about 6.2 g ofnitrogen was absorbed by the Fe-5Ti pebble of 95 kg. Accordingly, the rest 11.8 g of nitrogen
would be absorbed by another area, for example, by the inner wall of N hot trap container.
Moreover, it was succeeded to maintain the temperature of the hot trap at 500oC or 450oC in a
total time about 800 hours. The other results will be also presented.
The getter materials for nitrogen hot trap system in the lithium facility of IFMIF are proposed
to use the getter materials of Fe-Ti alloy or pure Ti, and the nitrogen hot trap system for Li
purification of the lithium facility was also evaluated in this study.

- 106 -

P1-51
Materials Test Plan and Design Summary in PIE Facility of IFMIF
E. Wakai, T. Kikuchia, T. Yutani, M. Soldainib, M. Shingalac
Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195, JapanaJapan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japanb
CEA, Paris, Francec
IFMIF/EVEDA Project Team, Rokkasho, Aomori 039-3212, Japane-mail address: wakai.eiichi@jaea.go.jp

Under the frame of the Broader Approach Agreement between Europe and Japan, theintermediate engineering design of the International Fusion Materials Irradiation Facility(IFMIF) plant has accomplished in the middle of 2013 in the IFMIF/EVEDA (Engineering
Validation and Engineering Design Activity). The IFMIF plant has five specific facilities suchas Accelerator Facility (AF), the Lithium Target Facility (LF), the Test Facility (TF), the Post
Irradiation Examination Facility (PIEF), and the Conventional Facilities (CF). The mainfunction of the IFMIF plant is to give the demanded design database for the licensing, and theoperation and maintenance of the Fusion DEMO reactor, and it is achieved from the materials
data set obtained from the high, medium, and low flux test modules (HFTM, MFTM, and
LFTM) of the IFMIF.
For materials to be tested in IFMIF, databases on irradiation properties should alreadypre-exist. The candidate (structural) materials to be tested in IFMIF should undergo a rigorous
pre-selection procedure, including evidence on radiation resistance to be proven in otherirradiation campaigns including fission reactor, multi-ion beam facilities and spallation
sources. Expected contributions will include: (a) Engineering design for fusion DEMO reactor,(b) Definition of performance limits, (c) Completion and validation of (existing) databases,(d) Selection or optimization of materials, (e) Validation of fundamental understanding ofradiation response of materials, and (e) Tests of blankets and functional materials prior to orcomplementary to Test Blanket Modules (TBMs) of the Internal Thermonuclear Experimental
Reactor (ITER).
The contents of KIT (A. Moeslang) report for test matrix of IFMIF, a comprehensive design repot(CDR) for IFMIF and a report of the Specification Working Group of the IFMIF/EVEDA teamare very important, however, the details of information for the schedule and test plan, andmethodology of PIE in IFMIF in order to produce database for the design of fusion DEMO reactor
is not necessarily enough in them. Further requirement and guidelines from users will be needed
to prepare and construct them. The main function of IFMIF facility is, off course, to produce thedatabase for the design of fusion DEMO reactors, while this facility may be also useful formaterials and biological researches of neutron scattering with continuous neutron beams with highintensities. In this study, the materials tests and the engineering design for the post irradiation

examination (PIE) facility of the IFMIF was evaluated.

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P1-52
Installation and commissioning of components for IFMIF/EVEDA
Prototype Accelerator
R. Ichimiya, J. M. AyalaA), Ph. CaraC), R. HeidingerC), A. JokinenC), A. Kasugai, J. Knaster A),
H. KobayashiB), K. Kondo, S. Ohira, Y. Okumura A), G. Phillips C), K. Sakamoto and
M. Sugimoto
Japan Atomic Energy Agency (JAEA), Rokkasho Fusion Research Institute, Rokkasho, Aomori, Japana
IFMIF/EVEDA Project Team, Rokkasho, Aomori, Japanb
High Energy Accelerator Research Organization (KEK), Tokai, Ibaraki, Japanc
F4E, Fusion for Energy, BFD Department, Garching, Germany

International Fusion Materials Irradiation Facility (IFMIF) is a projected
accelerator-driven-type neutron source for qualifying fusion reactor materials, characterized
by its beam current frontier accelerator producing two sets of 125 mA/CW deuterium ion
beams up to 40 MeV. Under the framework of the Broader Approach (BA) agreement
between Japan and Europe, Engineering Validation and Engineering Design Activities(EVEDA) for IFMIF were launched in 2007 to validate the key technologies on which the
IFMIF design is based. In order to realize such a high current CW accelerator, Linear IFMIF
Prototype Accelerator (LIPAc) has been developed and is constructed in Rokkasho, Japan.
LIPAc components are to a major part developed and contributed by European Institutes(CEA, CIEMAT, INFN, SCK-CEN) and consist of an injector, a RFQ accelerator, asuperconducting RF linac (SRF), a Low, a Medium and High Energy Beam transport and abeam dump. Its target is to produce 125 mA/CW deuterium ion beam up to 9 MeV adequate
to validate the 40 MeV deuteron beam of IFMIF. The accelerator building of LIPAc, central
control system and other facilities are developed and contributed by JAEA. Commissioning
and operation are taken by JAEA with tight collaboration with European Institutes, Fusion
For Energy (F4E) and Project Team (PT). In order to achieve the target performance, various
technologies including limit design have been implemented and also thus special care isrequired when installation and commissioning them. Therefore, we are developing
installation/commissioning procedures and improving them towards the optimum. In thispaper, we will introduce the design scheme of LIPAc and discuss the installation andcommissioning scheme/plan.

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P1-53
Evaluation of dark current profile for prediction of voltage holding
capability on multi-aperture multi-grid accelerator for ITER
R. Nishikiori, A. Kojima, M. Hanada, M. Kashiwagi, K. Watanabe, N. Umeda, H. Tobari,
M. Yoshida, M. Ichikawa, J. Hiratsuka, Y. Yamanoa and NB Heating Technology Group.
aJapan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japan
Saitama University, 255, Shimo-Okubo, Sakura-ku, Saitama 338-8570, Japan

Stable high energy beam acceleration is required in neutral beam injectors for JT-60SA and
ITER where 22- 40 A D- ions are accelerated up to 0.5-1MeV by a large electrostatic accelerator
with multi-aperture multi-grid (MAMuG). One of critical issues for such high-energy
high-current beam acceleration is the high voltage holding which is dominated by vacuum
discharges. In order to improve and design the voltage holding capability of MAMuG
accelerators based on the physics understanding, the physics of vacuum discharge is being
investigated. The past results suggest that vacuum discharge occurs beyond the threshold of thedark current due to the field emitted electrons from cathode to anode [1]. Therefore, analyticalestimation of the dark current plays an important role to evaluate the voltage holding capability.
The dark current can be derived from F-N theory [2] where electric field enhancement factor
β is included. Though, β could only be evaluated from the experiment previously. Therefore, themethod to decide β without experiment was considered.
β can be evaluated from the F-N plots with measured dark current. This time dark currents
were measured at three different areas in the MAMuG accelerator to compare β in different
electric field. Fig.1 shows schematic view of each measured areas and the electrical field
distribution between the electrodes. As a result the effective electric field βEave, where Eave isaverage electric field, were found to be almost constant for different areas although the β islargely different as shown in fig. 2. By applying βEave ~103 kV/mm β can be evaluatedanalytically, leading to the analytical prediction of the dark current without the measurements.
This prediction will enable the estimation of voltage holding capability by evaluating thethreshold of the dark current.

fig. 2 Schematic view of measured areas
and the electrical field distribution between
the electrodes.

fig. 1 Comparison of measured β and effective
electric field at each area.

[1] A. Kojima, M. Hanada, A. Hilmi et al., Rev. Sci. Instrum. 83, 02B117 (2012).
[2] R.H. Fowler, L.W. Nordheim, Proc. Roy. Soc. London A, 119 (1928) p. 3.

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P1-54
Variation in secondary electron collection and energy recovery
concerning electrode angle in a high energy ion penetration scheme
T. Wada, S. Konno, S. Nakamoto, H. Takeno, Y. Furuyamaa, A. Taniikea
Department of Electrical and Electronic Engineering, Kobe University,
1-1, Rokkodai-cho, Nada-ku, Kobe 657-8501, Japana
Graduate School of Maritime Sciences, Kobe University,
5-1-1, Fukae-minamimachi, Higashinada-ku, Kobe 658-0022, Japan

D-3He nuclear fusion has been expected as an energy source for next generation because mostof the energy obtained in this reaction is kinetic energy of charged particles and direct energy
conversion can be applied. An energy recovery device from produced protons by using
traveling wave (TWDEC) was proposed [1], but some protons were not decelerated andefficiency of the device was limited. In order to recover those protons passing through the
TWDEC, an additional device was proposed to be settled in the downstream, which wascalled secondary electron (SE) direct energy converter (SEDEC) [2]. In SEDEC, a lot of
metal foil electrodes are aligned in the direction of the proton beam. The incident protons
penetrate the electrodes where SEs are emitted. Collectors settled on both sides of the foil
electrodes catch the SEs and those energy is recovered by appropriately biased voltage.
A series of experiments were performed. In an
SEDEC with a simple structure, SEs arrived at not
the collectors, but the anteroposterior electrodes [2].
A magnetic field perpendicular to the proton beamby permanent magnets (PMs) was introduced, but
magnetic mirror effect in front of PMs disturbed
SEs arrival at the collectors [3]. Finally, PMs werealigned on one side and an appropriate distance tothe collector plate was taken as shown in Fig. 1. In
this structure, the amount of the recovered energy
exceeded that without PMs [4].
Fig. 1 Structure of SEDEC with PMs
Although the final structure succeeded in thealigned on one side.
improvement of SE collection, optimization of thestructure has not been studied yet. The structure has many parameters. Concerning foil
electrodes, parameters such as thickness, interval between electrodes, angle to the incident
beam, etc. can be considered. In this paper, we investigate variation in SE collection on the
aspect of angle to the incident beam as a first step for optimization of the structure.
The angle of the electrode to the incident beam affects SE collection in many points of view.
It changes equivalent thickness of the foil. The amount of SE emission may change according
to the incident angle to the foil. The distribution of initial velocity of SEs may also change.
The SE motion is disturbed by the electrode and may change because the angle to themagnetic field is changed. The experiments of SE collection by changing the angle of the foil
electrodes are in progress, and detail results and discussion will be presented in theconference.
[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] Y. Takeshita, et al., 9th APFA (2013) TP-36.
[4] S. Hagihara, et al., Plasma and Fusion Res. 10 (2015) 3405025.

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P1-55
Numerical Study on New Steering Grid of Negative Ion Source for NBI
K. Tsumori, M. Kisaki, O. Kaneko, Y. Takeiri, M. Osakabe, K. Ikeda, O. Kaneko,
K. Nagaoka and NIFS NBI group
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japane-mail: tsumori@nifs.ac.jp

New steering grid (SG) for a beam accelerator equipping multi-slot grounded grid (MSGG)
has been investigated using beam-trajectory simulation. A long circular shape with 13 mm(width) and 10 mm (height) is adopted for the SG to correct the asymmetric focal
characteristics of MSGG [1, 2]. Extracted hydrogen/deuterium-negative-ion (H-/D-) beamscratches inner wall of the narrower opening of the SG in the high current density of H- beam.
Scratched beam produces secondary electrons, which damage the grounded grid.
The new SG has φ13 mm circular apertures with so-called kerbs surrounding each horizontal
row of the apertures. The kerb has a similar shape of the slot of MSGG, and the beamlet
trajectories are controlled by sifting the vertical and horizontal centerlines of the kerb. In theprevious SG without the kerbs, coulomb repulsion between beamlets was corrected only bydisplacing the aperture center of the SG to inner direction. On the other hand, the beamlet
interactions are compensated with shifting the edge of kerbs in the new SG. This makes thedesign for arrangement of the aperture much easier.
The SG is made of feromagnetic material, nickel-plated iron, to short the magnetic circuit
induced with electron deflection magnets imbedded in extraction grid (EG), and is installed at
the exit side of the EG. By reduce the return magnetic field, the field strength on the opposite
side, plasma-grid side, of the EG increases and more electron suppression is expected in thecase of D- extraction. Influences of the feromagnetic SG to the other magnetic field and thatto the beamlet trajectories are indicated with simulation result.
[1] K. Tsumori, O. Kaneko, Y. Takeiri et al., Proceedings of the 10-th International
Symposium on Production and Neutralization of Negative Ions and Beams, AIP
Conference Proceedings 763 (2005) pp.35-46.
[2] K. Tsumori, Y. Takeiri, O. Kaneko et al., Rev. Sci. Instrum, 79 (2008) pp. 02C107-1 02C107-3- 111 -

P1-56Fabrication and property investigation of metal hydride blocks
for neutron shielding in a fusion reactor
T. Tanaka, H. Mutaa, M. Yoshinob, Y. Hishinuma, T. Muroga, A. Sagara
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan
Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871, Japanb
Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603, Japana

Hydride materials such as ZrH2 and TiH2 have been adopted as superior neutron shielding
materials in neutronics designs of fusion reactors. However, for investigation of a detailed
shield design, material studies on fabrication and properties of the hydride blocks, andhydrogen desorption behaviors in accident conditions are required.
In neutronics benchmark experiments using a DT neutron source, the minimum size ofblocks used to construct an irradiation assembly is ~50 x 50 x 50 mm3 [1]. Therefore, test
fabrication of blocks from commercially available TiH2 powder has been performed with thedimensions of ~25 x 25 x 5 mm3 and ~25 x 25 x 12.5 mm3 in the present study. Cold-pressedblocks were prepared by using a die at 80-100 MPa and following densification were performedby cold isostatic pressing (CIP) at 200 MPa to apply pressure to the blocks from all directions.
The densities of the TiH2 blocks after the CIP process were ~75 % of the theoretical density of
3.75 g/cm3. These densities are almost same as those of small thin TiH2 discs (10 mm x ~2 mmthick) fabricated by cold-pressing using a die at 200 MPa [2]. If a CIP apparatus with 400 MPa
is used, the densities of the TiH2 blocks is expected to be ~85 % of the theoretical density from
relation between pressures and densities obtained in our previous fabrication of small
cold-pressed TiH2 discs [2]. Test fabrication of ZrH2 blocks and study on a densification
method are being conducted at present.
Regarding behaviors of the hydride materials in accident conditions, oxidation of the TiH2
and ZrH2 powders in an air flow is examined by thermogravimetry and differential thermal
analysis (TG/DTA) up to 1000 oC. Composition in the flowing air was analyzed simultaneously
by quadrupole mass spectrometer (QMS) to examine desorption of hydrogen under theoxidizing condition. The TG data showed that oxidation started from ~300 oC in both thepowders and significant increase in weight were observed at 600-900 oC in the TiH2 powder
and at 400-900 oC in the ZrH2 powder. In the QMS data, hydrogen desorption from the TiH2
powder was detected in the range of 400-600 oC and this is considered almost independent fromthe oxidation process. In the ZrH2 powder, hydrogen desorption was detected in the range of
350-800 oC and might be affected by the oxidation process. After the heating up to 1000 oC in
air, the flat surface of the ZrH2 powder packed in a specimen cup became significantly rough.
Deformation of the hydride blocks under oxidation conditions is required to be examined
especially for ZrH2 for discussion of a shield design.
[1] T. Tanaka et al., Fusion Science and Technology 60 (2011) 681-686.
[2] H. Muta et al., Plasma and Fusion Research 10 (2015) 3405021.
The present study is being conducted under supports of KAKENHI 25289344 and NIFS UFFF0251.

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P1-57
Improvement of weight density and thermal conductivity of hydride
neutron shielding material using metal coated powder
H. Muta, T. Tanakaa, R. Nishikane, Y. Ohishi, K. Kurosaki, Y. Hishinumaa,
S. Yamanaka, T. MurogaaaOsaka Universty, 2-1 Yamadaoka, Suita, Osaka 565-0871, Japan
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
muta@see.eng.osaka-u.ac.jp

To reduce a fusion reactor size, a thinner radiation shield layer with high density neutron
shielding materials must be installed. Metal hydrides are promising as the shielding materials
because they can include higher density of hydrogen atoms than liquid hydrogen [1]. However,
it is known that the bulk non-crack metal hydride is difficult to fabricate due to the brittlenessand the large volume change in the hydrogenation. Previously, to simplify the fabrication
process, we tried to make dense hydride pellets by cold-pressing from their powders [2]. Morethan 80 % of theoretical density and acceptable thermal conductivity could be obtained by thesimple process. In the present study, to improve the density and the thermal conductivity, metal
coating on the hydride powders were examined.
Commercial -titanium hydride (TiH2) (Aldrich, 99 % purity) was used as a starting
material. The powders were coated by copper (Cu) by using electroless Cu plating process atroom temperature. The plating solution was C-200LT (Japan Pure Chemical Co., LTD). After
plating, the powder was washed by ethanol. Then the powder was pelletized by cold-pressing
with pressure of 400 MPa. The density was calculated from the weight and dimensions. The
thermal diffusivity, , was measured by laser flash method using Netzsch LFA457 to calculate
the thermal conductivity, , from the equation of  =  Cp d. Cp and d are heat capacity andsample density, respectively.
Figure 1 shows SEM image of the Cu-coated TiH2 pellet. Small particles of Cu are
distributed in gap of the large TiH2 particles. Due to the Cu coating, the relative density of thepellets increases from 83 % [2] to > 90 %. The Cu particles appears to act as lubiricant, whichmight promote the particle displacement at the cold-pressing. The thermal conductivity isshown in figure 2. In spite of the increase of sample density, the thermal conductivity is almost
same as that for the pure TiH2 pellet because of decrease of the thermal diffusivity.

12

-1

-1

Thermal conducitivty (Wm K )

14

10
8

Cu-coated TiH2
400 MPa

6
4

TiH2[2]200 MPa
400 MPa910 MPa

2
0
300

350

400

450

500

550

600

650

Temperature (K)Fig. 1 SEM image of Cu-coated TiH2 pellet.

Fig. 2 Thermal conductivity of the TiH2 pellets.

[1] T. Tanaka et al., to be published in Fusion Science and Technology.
[2] H. Muta et al., Plasma and Fusion Research 10 (2015) 3405021.
The present study is being conducted under supports of KAKENHI 25289344, 26420871 and NIFS14KEMF060.

- 113 -

P1-58Environmental Effects on V-4Cr-4Ti Alloy during Irradiation
H. Watanabe1, T. Nagasaka2, T. Muroga2,2

Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
1
National Institute for Fusion Science, 509-5292 Gifu, Japan

In the V-4Cr-4Ti alloy, it is well established that interstitial impurities such as oxygen andnitrogen play an important role in radiation effects such as microstructure change, irradiation
hardening and embitterment. Among them, oxygen is regarded as the most effective atom in
the alloy. Furthermore, the alloy is considered to be used as a fusion blanket structural
material that face to a vacuum and helium gas, which may contain a low partial pressure ofoxygen. Recently, our ion irradiation experiments on V-4Cr-4Ti alloys at 473-973K revealed
that oxygen pick-up from irradiation environment (namely, vacuum) is very minor up to thelevels of 1 dpa. But at higher dose levels, the effects might cause degradation of mechanicalproperties due to enhanced formation of titanium oxide precipitates during irradiation. To
understand the radiation induced phenomena related with oxygen, such as oxide formation in
the matrix, not only oxygen level in the alloy but also oxygen pick-up from radiation
environment must be considered. To reduce the oxygen concentration in the V-4Cr-4Ti alloy,
oxygen gettering using Zr foil and Y addition on V-4Cr-4Ti alloy were a well-known
technique. The objective of the present study is, therefore, to understand the detailed
mechanisms of oxygen pick-up from irradiation environment which related with titaniumoxide.
For neutron and ion irradiation, NIFS HEAT 2 and welded joint samples by YAG laser,
Y doped V-4Cr-4Ti alloys were used in this study. The samples were irradiated with 2.4MeV
copper ion irradiation with a tandem accelerator at Kyushu University. The TEM specimens
were sliced from welded materials and irradiated at 573 and 873K up to the dose of 12 dpa.
After irradiation, the specimen was electro-polished by a back thinning method, and the area
near the peak damage region (at about 700 nm) was observed. These samples were also
irradiated in JMTR and BR2 in the temperature range of 563-873K. After the irradiation,
conventional TEM (CTEM) observation and detailed chemical analysis by Cs corrected
microscopy were performed by using a JEOL ARM200FC located in a radiation controlledarea in Kyushu University.

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P1-59
ACT2: a high heat flux test facility using electron beam for fusion
application
Y. Hamaji, M. Tokitani, R. Sakamoto, H. Tamura, A. Sagara
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

ACT2 (Active Cooling Teststand 2), a high heat flux test facility has been upgradedfrom ACT facility and started operation. A newly installed electron gun and data acquisition
systems of ACT2 enables powerful and flexible experiments for R&D activities of plasma
facing components (PFCs) for DEMO and beyond such as FFHR-d1[1]. The beam profile andheating performance were characterized by measurements using a pyrometer, thermocouples,
fast visible camera and graphite probes.
Thermo-mechanical properties, heat removal performance and reliability of plasma
facing components (PFCs) are crucial for long-term operation in future fusion reactors such as
FFHR-d1. High heat flux testing with reactor relevant heat flux with actively cooled PFCs isessential for R&D of PFCs and fundamental researches of material response to high heat flux. Ahigh flexibility of electron beam control is also required to test components in various scale andshape. The original ACT facility was established in 1994[2]. Nominal power and flexibility ofbeam control system of ACT was not sufficient for resent requirement for heat flux test facility.
ACT was upgraded into the ACT2 to satisfy these demands. Technical characteristics of ACT2
are listed in Table 1. Focused electron beam scans defined area using high frequency deflectionlenses. 300 kW beam power and 60 degree of deflection angle of ACT2’s electron gun allow toapply reactor relevant heat flux (~30 MW/m2) into defined area up to 100 × 100 mm2. Bothsteady state and transient (~ms) heat load are capable. Cyclic heat loads for thermal fatigue test
are achieved by automatic routinely beam transfer between defined areas with each dwell time.
In order to remove the effect of inhomogeneous heating at the edge of scanned area, awater-cooled “beam limiter” made of copper with specific aperture is located above the sample
surface. The edge of scanned area is scraped by the limiter. Absorbed heat flux is calculated
from electron current through the sample and irradiated area defined by limiter’s aperture. To
characterize electron beam of ACT2 facility, beam profile and local heat flux is measured by 2
mm diameter graphite probes that are embedded in a target tile at a high time resolution. Beamprofile was obtained as Gaussian-like peaked profile with the spot size about 23 mm. Peakvalue of local heat flux during each pass of scanned beam was also obtained.
[1][2]

A. Sagara, T. Goto, et al., Fusion Eng. Des. 87 (2012) 594–602.
Y. Kubota, N. Noda, A Sagara, O. Motojima, Vacuum 47 (1996) 955–957.
Table 1. Technical Characteristics of ACT2
Parameter
Max. beam power
Accelerating voltage
Max. beam current
Beam deflection frequency in x direction
Beam deflection frequency in y direction
Max. heated Area
Beam spot diameter
Typical flow rate of the cooling water

Value
300
40
7.5
500
5000
500 × 500
~2042

- 115 -

unit
kW
kV DC
A
Hz
Hz
mm
mm
l/min.

P1-60Conceptual design of a new liquid metal divertor REVOLVER-D
J. Miyazawaa,b, T. Gotoa,b, T. Ohgob, N. Yanagia,b, T. Murasea, H. Tamuraa, A. Sagaraa,b,
and the FFHR Design Groupab

National Institute for Fusion Science,322-6 Oroshi, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi, Toki 509-5292, Japan

A divertor system with easy maintenance and low radioactive wastes is desired for a fusion
reactor. It is also important that the divertor can handle a high heat load exceeding 20 MW/m2.
A liquid metal divertor is one of the promising candidates. Especially, a continuous flow of
the liquid metal of the order of ~1 m/s will be effective to remove the high heat load in steadystate. According to these demands, a new liquid metal divertor system named the
REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) is proposed.
The REVOLVER-D is composed of the vertical free-surface stream (VFSS) as shown in Fig.
1, where two VFSS of the molten solder (60 % tin + 40 %lead) with and without a ball-chain inside are shown. The
landing points of the two VFSS were different althoughthese were ejected from the same nozzle at the same flowrate with an identically slanted angle, showing a reduced
velocity in the case with the ball-chain. Instead of thesolder, pure tin (Sn) is considered to be the first candidate
for the REVOLVER-D because of its low melting
temperature, low vapor pressure, low material cost, highsafety (low toxicity and no explosive reaction with water),and high nuclear stability.
The top-view of the REVOLVER-D showing its nozzle
layout is shown in Fig. 2. In this example, 52 nozzle units
drawn by large circles are used. Each of the nozzle unit
has 10 nozzles drawn by small circles, to generate 520
VFSS in total. Ball chains or something equivalent tothem will be installed on each nozzle, to slow down the
VFSS against the gravity. As seen in the figure, the Fig. 1. Photos of VFSS w/ andmagnetic field lines slanted through cannot penetrate the w/o a ball-chain inside. The
REVOLVER-D. Many magnetic field lines hit the VFSS height of the VFSS is ~170 mm.
inside the REVOLVER-D and the heat load is absorbed
Plasma Side
there. This is why we call this the “effectively volumetricdivertor”. On the other hand, the neutral particles
generated inside the REVOLVER-D can easily move tothe vacuum pump side, since the conductance in thisdirection is larger than those of the other directions.
In the case of a helical fusion reactor FFHR-d1, it iscalculated that 10 REVOLVER-D set at the inner side of
the torus to intersect with the ergodic layer can correct
more than 80 % of the magnetic field lines directed
toward the divertor region. Applying the REVOLVER-D,
it becomes possible to design a helical fusion reactor
Vacuum Pump Side
without complicated full-helical tungsten divertors. Easy
maintenance and reduced radioactive wastes together
with the high-safety of the REVOLVER-D are attractive Fig. 2. An example of the
REVOLVER-D nozzle layout.
for the fusion reactor.

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P1-61
Experimental simulations for detached plasma with Super-X divertor
using a liner divertor simulator TPD-Sheet IV
Y. Tanakaa, T. Iijimaa, T. Hasea A. Tonegawaa, K. N. Satob and K. KawamuraaabTokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa, 259-1292, Japan
Chubu Electric Power Co. Inc., 20-1 Kitasekiyama, Ohdaka-cho, 459-8522, Japan
E-mail: tanachu09@gmail.com

In the magnetic confinement fusion reactor for high power and long pulse operation, enormous
heat flux (exceeding 10 MW/m2) is expected to flow onto divertor plates from core plasma. In
order to reduce this heat load, the divertor geometry on stationary detached plasma formationmust be realized. In addition, the neutral particle flowback into the core plasma is necessary to
suppress by the divertor geometry. The Super-X divertor has been proposed by DEMO divertor
concept, investigated using simulations. Its optimization needs detail information on confinedneutral particles and their physical process. An experimental simulation on the V-shaped,
long-leg and baffled long-leg target for closed divertor has been conducted by a liner divertor
plasma simulator TPD-Sheet IV [1]. Experimental simulations of divertor geometries for theformation of detached hydrogen sheet plasma using the linear divertor plasma simulator
TPD-Sheet IV. In order to understand the basic mechanism of detached plasma, we havecarried out an experiment using two target geometries (V-shaped target, super-X target). In thisexperiment, the ionization and recombination rate are discussed by using thecollisional-radiative (CR) model. The electron density and temperature were measured using aLangmuir probe. Ionization and recombination ratio (PIon, PRec) are discussed fromexperimental data of Te and ne, using the collisional-radiative (CR) model. The results expectthat a Super-X divertor with a Super-X target effectively enhances plasma detachment.

[1] S. Tanaka, et al., Fusion Sci. Tech. 63 (2013) 420- 117 -

P1-62
Effects of gamma-ray irradiation on electronic and non-electronicequipment of the Large Helical Device
K. Ogawaa,b, T. Nishitania, M. Isobea,b, M. Satoa, M. Yokotaa, H. Hayashia
T. Kobuchia, T. Nishimuraa, S. Imaicab

National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (The Graduate University for of Applied Advanced Studies), School of PhysicalSciences, 322-6 Oroshi-cho, Toki 509-5292, Japanc
Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603, Japankogawa@nifs.ac.jp

Deuterium operation will be started from 2017 on the Large Helical Device (LHD). In
deuterium campaign, measurement and control equipment of the LHD must be survive underradiation environment. The LHD is controlled by means of many semiconductor integrated
circuits. The circuit is regularly replaced latest highly-integrated one. The circuit becomes
weaker against radiation year by year because of higher integration. Water sealing gaskets
made of polytetrafluoroethylene are used on the LHD to serve the cooling water. Radiation
resistance polytetrafluoroethylene should be studied to avoid the leakage of water. Neutrons
are mainly generated by deuterium reactions, on the other hands, gamma-rays are generated
by materials activated by neutrons. Equipment of the LHD is irradiated with gamma-rays,
regardless of the presence or absence of plasma discharges. Both transient effect andpermanent effect can be caused by irradiation on equipment. Permanent effect is more serious
because the damage cannot be removed forever. So, understanding of permanent effect of
gamma-ray irradiation on equipment being used for the LHD is very important for the
deuterium operation.
Gamma-ray irradiation on the programmable logic controllers (PLCs), PC, flow meterscomposed of a light-emitting diode and a photodiode, isolation amplifiers, media converters,
and water sealing gaskets are done at the 60Co irradiation facility of Nagoya University. The
integrated dose is from 100 to 500 Gy. We found that analogue to digital and digital toanalogue converter module of PLC does not work after 100 Gy irradiation, and Ethernet
connection of the PC is failed after 300 Gy irradiation. Flow meters, isolation amplifiers andmedia converters work after 100 Gy irradiation. In this presentation, detailed results of
gamma-ray irradiation on electronic equipment and water sealing gaskets will be reported.

- 118 -

P1-63
Desorption of non-metal impurities from Pb and Pb-Li alloy
M. Kondo1), Y. Nakajima2), T. Tanaka3), T. Nozawa4), T. Yokomine5)
1) Research laboratory for nuclear reactors, Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku,
Tokyo 152-8550 Japan
2) Department of Nuclear Engineering, Tokai university 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292,
Japan3) National Institute for Fusion Science, Toki, Gifu 502-5292, Japan
4) Japan Atomic Energy Agency, 2-166 Omotedate, Obuchi, Rokkasho, Aomori 039-3212
5)Kyoto University, Kyotodaigaku-Katsura, Kyoto, JapanTo TPD-MS

Liquid lithium lead alloy (Pb-Li) is one of the promising candidates of the tritium breeders for fusion DEMO
reactors. The study on the fabrication methodology of the alloy having high purity has been studied [1]. However,the control of some non-metal impurities such as hydrogen, oxygen, nitrogen, water and carbon dioxide dissolvedin the alloy is an important issue. The low oxygen concentration in the alloy should be kept to suppress thedepletion of Li by its selective oxidation. The hydrogen in the alloy should be removed to keep the solubility oftritium. Though the effect of the nitrogen on the chemical characteristics of the alloy is not made clear, nitrogenalso should be removed from the alloy to suppress the loss of Li due to the selective nitriding of Li in the alloy. The
information on these non-metal impurities such as hydrogen and oxygen in the alloy is quite limited. The control of
the non-metal impurity in the Pb-Li alloy is not made clear so far. The purpose is to investigate the potentiality of
the high temperature desorption method for the removal of the non-metal impurities from the Pb-Li alloy.
The temperature- programmed desorption (TPD) experiments were performed at the test conditions summarizedin Table 1. Figure 1 shows the test section connected to the Temperature Programmed Desorption and Mass
Spectrometer (TPD-MS). The test cylinder was connected through some stainless steel tubes, the cold trap andvalves into the TPD-MS. The valve (A) was used for the evacuation of the TPD-MS system. The valve (B) was
opened and the temperature of the test cylinder was raised at a temperature rising rate of 2.5K/min, starting at
308.5 K and finishing at 774.5 K.
Valve (B)
Large desorption of CO2, O, C and H2O from Pb
Valve (A)samples (types A and B) even below the melting point.
30cm
The result for Pb-Bi test indicated that the desorption
behavior of the non-metal impurities was close to that
Filterfrom the Pb samples. The standard Gibbs free energies
for the formation of oxides indicated that PbO is morestable than H2O and CO2 in these cases. The molecules
Test cylinder
were attached on the surface and dissolved in the alloy.
Outer diameter: 19.1mm
Inner diameter: 15.6mm
TC
Hydrogen and oxygen were desorbed from the Pb-Li
10cm
alloy around 700K. The desorption of H2O and CO2
from Pb-Li alloy was much smaller than that from Pb
HeaterCold trap:and Pb-Bi samples. This is because Li in the alloy react
room temperature
Samplewith H2O and CO2, and LiH, LiOH and LiO2 are formed(Immersion to water tank)(100g)and dissolved in the alloy. Some part of hydrogen and
Fig. 1 Test section of TPD system
oxygen must be dissolved in the fabrication procedureof the alloy due to the chemical reaction between Li andH2O in Pb [1]. It was indicated that hydrogen and oxygen can be removed from the Pb-Li alloy by TPD method.
Table 1 Test conditions
Test material
PuritySample quantity [g]Pb (type A) [1] 99.999%
100
Pb (type B) [1] 99.9%
100
Pb-Bi
100
Pb-17LiLi (99%) + Pb (99.9%)100
Pb-17LiLi (99%) + Pb (99.9%, Baking)1
Pb-15Li [1]Li (99%) + Pb (99.999%)
100
Pb-28Li [1]Li (99%) + Pb (99.999%)
100
TiH2
Calibration of hydrogen desorption 0.011
Reference [1] Y. Nakajima, M. Kondo, T. Nozawa, Study on fabrication method of lithium alloys with metal grain,
Fus. Eng. Design in press (2015).

- 119 -

P1-64
Conceptual design of temporally storage area in hot cell for fusion DEMO reactor
M. Kondoa, Y. Someyab, M. Tsujic, S. Yanagiharad, T. Kato, Y. Sakamotob, K. Tobitab, S. MatsudaaaTokyo Institute of Technology, 2-12-1, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550, Japan
Japan Atomic Energy Agency, Rokkasyo-mura, Kamikita-gun, Aomori 039-3212, Japanc
Tokai University, 4-1-1, Kitakaname, Hiratuka, Kanagawa, 502-5292, Japand
Research Institute of Nuclear Engineering, University of Fukui, Fukui 914-0055, Japanb

The maintenance scenario and the waste management of fusion DEMO reactor having a fusion output of
1.35GW have been proposed. In the maintenance procedure of the reactor, used in-vessel components were
removed from the reactor vessel for their replacement. Then, the low-level radioactive wastes (LLW) of
approximately 6000 t are generated. The reuse and the recycle of the blanket components can reduce the quantityof the LLW by 60% in total. However, these components are radioactivated and having a large decay heat. Theirdismantlement is available by the remote handling (RH) only after their temporally storage for a certain period in
the hot cell till their radiation level decreases to allowable level for the RH. Therefore, the temporally storage ofthe blanket components in the hot cell is the important procedure for the waste management of the fusion reactors.
However, the information for the temporally storage of the used blanket components in the hot cell is quite limited.
The used blankets are removed as multi module segments from the vacuum vessel (VV) though the vertical port.
The segments are the back plate integrated blanket modules. These segments must be stored temporally in the hotcell after the removal of residual tritium by thermal detritiation process. During the temporally storage of thesegments, their decay heat must be removed. The segments must be kept at the adequate temperature to suppress
the effect of the heat treatment on their reuse and recycle. In the present study, it is proposed that the used blanket
segments are stored dispersively in the plural compartments of the temporally storage area in the hot cell. The
motivation for this system is to achieve the reduction of the risk, the simplification of the hot cell structure and themitigation of the cooling conditions of the decay heat in the same time without the increase of total floor spacerequired for the storage. The changes of the dose rate and the temperature distribution in the compartment withtime were investigated by means of gamma ray transport calculation and thermofluid calculation. The purpose of
the present study is to propose the effective scenario and design of the temporally storage area in the hot cell.
The used segments are stored dispersively in the plural compartments in the hot cell for thefollowing dismantle step. The total floor space required for the temporally storage of the segments are
estimated less than 1500 m2. The compartment, which is made of a heavy concrete, has high shielding
performance and cooling performance. Fig. 1 shows the schematic drawing of the compartment. The
temporally storage of the segments with the pentagonal arrangement in the compartments, which can makethe gentle distribution of the dose rate without the formation of the hot spot in the compartment, is proposedin the present study. The change of the dose rate in the compartment with time was evaluated by means of the
gamma ray transport calculation performed by PHITS Monte Calro analysis code. It was found that thedismantle process by the remote handling can be available after the temporally storage for 10 years. The
decay heat during the temporally storage of the segments can be removed by the flowing He. The segments
can be kept at the temperature lower than 550 ºC, which does not affect the mechanical properties of the backplate as the heat treatment for their reuse in the other operation of the reactor.
1m

A-A’

Coolant inlet
Heavy concrete

1m

1m

1m

Lining(Stainless steel)

D=1m

Point DI

F=1m

Segment(Blanket)A
Blanket

A’
5m

10mm

10m

22m

1m

C
Point D

0.5m

Back Plate

0.6m

0.4m

B (Floor surface)
T1
T1’
T2
T3

Lining ofstainless steel

Heat insulator
50mm
Cooling pipe

C (Side wall)
Cooling pipe

50mm

(Back plate)

1.5m

T4
Tb

B

Coolant outlet

Heavy concreteT1 T2 T3 T4 Tb(a)(b)
Fig. 1 Structure of (a) Simplified model of the segment having a rectangular parallelepiped shape and(b) Compartment for temporally storage of plural segments in hot cell- 120 -

P1-65

Effect of HIP temperature and cooling rate on the microstructure andhardness of dissimilar-metal joints between ODS-RAFM steels and JLF-1 steel
Haiying Fu, Takuya Nagasaka a, Takeo Muroga a, Akihiko Kimura b, Shigeharu Ukai c
SOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292, Japana
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanb
Institute of Advanced Energy, Kyoto University, Uji 611-0011, Japanc
Graduate School of Engineering, Hokkaido University, N13, W18, Kita-ku, Sapporo 060-8628, Japan

Because of excellent high-temperature mechanical properties, ODS-RAFM (oxidedispersion-strengthened reduced activation ferrtic/martensitic) steels can be used partly for thesurface of fusion blanket by bonding with conventional RAFM steels to enhance the acceptabletemperature of the blanket surface by 100-150oC. In this work, candidate ODS-RAFM steels,such as 9Cr-ODS and 12Cr-ODS, were diffusion bonded to a conventional RAFM steel JLF-1 by
hot isostatic pressing (HIP) for the purpose of the connection between ODS-RAFM steel andconventional RAFM steel components. A previous investigation revealed that JLF-1 base metal(BM) was hardened by quenching with martensite transformation after the HIP at 1000oC in acommercial furnace with 5oC/min in cooling rate, however the 9Cr-ODS BM was softened due tono quenching and carbides coarsening. Since nano-oxide particles impede the martensite
transformation, rapid cooling after HIP is required for 9Cr-ODS to keep their strengthening bymartensite. However, the cooling rate is generally decreased due to the increase in heat capacity
for the large mass of blanket itself and also the large mass of HIP furnace for it. On the other
hand, higher content of Cr in 12Cr-ODS can strengthen the steel instead of martensite andstabilize its ferrite phase, and thus can avoid the degradation of strength in the HIP process.
However, microstructural change with carbide coarsening is also an issue for strengthening of
12Cr-ODS. In the present study, therefore, the effects of HIP temperature and the subsequentcooling rate on the microstructure and hardness were investigated to obtain the optimum HIP
bonding conditions and proper strength at the whole part of the joints.
HIP bonding was examined at 1000oC, 1050 oC, and 1100 oC under a pressure of 191MPa for
3h with a cooling rate of 5oC/min. In addition to the conditions for a commercial HIP furnace,various isochronal annealing was carried out at 800-1100oC with controlled cooling rate of from
0.5 to 36oC/min by a laboratory-scale furnace.
Quench with hardening was observed for 9Cr-ODS after the annealing at 1000oC with
36oC/min in cooling rate, however was not induced with 0.5oC and 5oC/min. While, the annealing
at 1050oC induced quench hardening with 5 oC/min in cooling rate in addition to 36 oC/min. Basedon these results, HIP temperature can be low as 1000oC under rapid cooling condition, 36 oC/min
in cooling rate, to get quenched martensite and avoid softening of 9Cr-ODS BM. On the other
hand, the cooling rate 0.5 oC/min never induced quench hardening even up to 1100 oC. The quench
is enhanced not only by rapid cooling after the HIP, but also by carbide decomposition andcarbon dissolution at the HIP temperature. Therefore the critical cooling rate for quench, whichshould be between 5oC/min and 36oC/min for the 1000oC HIP, is increased with decreasing the
HIP temperature.
The proper HIP conditions for 12Cr-ODS will also be investigated, and will be comparedwith that for 9Cr-ODS.

- 121 -

P1-66
Investigation on the Magnetic Field Distortion by Ferromagnetism of theBlanket for the Helical Fusion Reactor
Xiang Jia, b, Nagato Yanagia, Takuya Gotoa, Hitoshi Tamuraa, Junichi Miyazawaa,
Akio Sagaraa, Songke Wangb, Minzhong Qib, Yuntao Songbab

National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
Institute of Plasma Physics, Chinese Academy of Sciences, P.O.BOX 1126, Hefei 230031, China,

In the future fusion reactors, ferritic steel is a realistic candidate to be used as a structural
material of the blanket, because of the low activation characteristics after neutron irradiation
which can reduce the issue of radioactive waste disposal. However, since the ferritic steel is aferromagnetic material, distortion of magnetic field produced by the magnet coils is a concern
on the plasma confinement due to island formation and partial changes of magnetic surfaces.
The conceptual design studies on the LHD-type helical fusion reactor (FFHR) areprogressing at National Institute for Fusion Science (NIFS) by employing the heliotron
magnetic configuration with continuous helical coils. In order to include the distortion ofmagnetic field produced by the ferritic steels in the blanket, the numerical calculation code
ANSYS based on the finite element method (FEM) is incorporated. First, the precise structureof the magnet coils, vacuum vessel, and blanket is input to establish the 3D modelling for thefinite element calculation. Second, the magnetic field by assuming nonmagnetic materials in theblanket is calculated as a reference. In this phase, the magnetic field-line tracing code is
combined to obtain the magnetic surfaces and their associated parameters, such as the rotationaltransform and magnetic well. Third, the nonmagnetic material is changed to ferritic materials
and the new magnetic fields are calculated. Comparison of magnetic field properties bychanging the permeability of the structural material is investigated. The calculation results will
be described in the presentation.

- 122 -

P1-67
Feasibility study of react-and-wind method for helical coils wound fromcable-in-conduit conductors.
S. Imagawa
National Institute for Fusion Science, 322-6 Oroshi, Toki, GIFU 509-5292, JapanSOKENDAI, Department of Fusion Science, 322-6 Oroshi, Toki GIFU 509-5292, Japan
imagawa@LHD.nifs.ac.jp

Cable-in-conduit (CIC) conductors made of a multi-strand cable of superconducting wires
within a tube-like conduit have been developed for large superconducting magnets. Major
features of the CIC conductors are a large current up to 100 kA, high mechanical strength with athick conduit, small AC losses, and high cryogenic stability by forced-flow cooling withsupercritical helium. Since the technology related to high-field CIC conductors with Nb3Sn
wires has been widely advanced through the construction of ITER, they should be the firstcandidates for the magnets of next fusion reactors. The major disadvantage is the reduction ofcritical current (Ic) caused by la4*( 6+(40$/ 564$,1 '7( 62 +($6 64($60(16 73 62  = )24
production of the A15 phase. In the case of SS316 conduit, the compressive thermal strain in
the superconductor such as Nb3Sn or Nb3Al exceeds 0.7%, and their critical current is
considerably degraded. In order to prevent the further increase of the strain, the heat treatment
is usually carried out after winding, which is called a "wind-and-react" method. A "react-andwind" method, however, is preferred for a helical coil to avoid a huge oven for the heat
treatment. In this concept, the heat treatment of a CIC conductor is carried out on aconventional bobbin, the circumference of which is the same as the length of one pitch of thehelical coil [1]. After that, the conductor is wound into the helical coil case with being pulled
aside, that is, being twisted. Since the superconducting wires in the conduit are multi-stage
twisted, the effect of twisting the conductor on the strain of the wires is considered to be small.
Furthermore, if the compressive thermal strain of the wires is reduced by being twisted, Ic of
CIC conductors can be increased with the react-and-wind method.
In order to examine the effect of twisting the conductor on the strain of the wires,
experimental study with a sub-size CIC conductor sample has been carried out. Three copper
wires are twisted together, and 6×4 sets of the triplets are wound around a SS304 pipe of thediameter of 6 mm that simulates a center channel for coolant. The whole assembly is inserted
into a SS304 conduit of the diameter of 17.3 mm. The wires are fixed to the conduit with epoxy
resin at both the ends. The longitudinal strains of the wires are measured with strain gauges on
the copper wires. The experimental results show that the tensile or compressive strain is inducedin the wire by twisting the conduit in the same or counter direction to the wire twisting direction.
In this experiment, the amounts of change in strain of wires are in the order of 1/10 of thetorsional strain of the conduit. The strain of the wires is considered to depend on the change of
their lengths. In order to examine the change of Ic of Nb3Sn wires by twisting the conductor,
small CIC conductor samples in spring-shape are under preparation. The torsion strain is
induced on the conduits by changing the length of the spring. In this paper the concept of reactand-wind method for a large helical coil and the expected effect on Ic are discussed.
[1] S. Imagawa et al., Concept of Magnet Systems for LHD-type Reactor, Nuclear Fusion,Vol. 49 (2009) 075017(7pp).

- 123 -

P1-68
Evaluation of Shear Strength in Soldered and Mechanical Lap joints of HighTemperature Superconducting tapes Intended for a Remountable Magnet
L. Aparicio, S. Ito, H. Hashizume
Tohoku University, 6-6-02-1 Aramaki Aza Aoba-Ku, Sendai 980-8579, Japan
lapa@karma.qse.tohoku.ac.jp

Remountable and joint-winding high-temperature superconducting (HTS) magnets have been
proposed for a helical fusion reactor FFHR-d1. Mechanical lap joint (MLJ) and soldered lap joint(SLJ) of HTS coated conductors (CC) are joint method candidates for the joint above magnets.
These joints must have sufficiently large mechanical strength to be tolerant to electromagnetic
forces. Tensile test of MLJ between two HTS CC having copper (Cu) stabilizer with a contact areaof 25 mm2 was performed in a previous study [1], whose results suggested a linear dependency
between joint conductivity and shear strength of the joint Also, shear strength with joint pressure>50 MPa was higher than the required value. On the other hand, in tensile test for a SLJ of HTS
CC having no Cu stabilizer with a contact area 200 mm2, fracture in the joint occurred due to stressconcentration at the edge of the joint [2].
Furthermore, evaluation of the mechanical behavior of SLJ of HTS CCs with Cu stabilizer has notbeen reported yet. Therefore this study performs tensile test on both MLJ and SLJ using HTS CCs
with Cu stabilizer, to analyze and compare their mechanical behavior and fracture mechanism
based on contact and friction theories.
Figure 1 shows the experimental setup for the tensile test performed in liquid nitrogen. Joint areafor MJL and SLJ used in this study were 50 and 25 mm2, respectively. In fabrication of MLJ,
pressure of 100 MPa is applied at joint section where a 100 m indium film is inserted between
joint surfaces, then the pressure is once released and after that the pressure was kept a certain value.
SLJ was fabricated using Tin-lead (Sn63Pb37) solder.
Normalized tensile test results plotted in figure 2 show
a good agreement with those given in ref. [1] for the
MLJ with contact area of 50 mm2. Also, notablediscrepancy between both joint methods is seen, that is,in the SLJ the maximum stress is almost constant due
to different failure mechanism reported in [2]. Further
results will be shown in full paper.

Fig. 2. Maximum tensile load vs Joint
conductivity of mechanical and soldered
lap joint with different joint pressure
values.

Fig. 1. Sample holder set-up.

[1] S. Ito et al., IEEE Trans, Appl. Supercond, Vol 25, 2015, 4201025.
[2] M, Sugano et al, IEEE Trans, Appl. Supercond, Vol. 17, 2007, 9606863- 124 -

P1-69NITA Coil ―Innovation for Enlarging the Blanket Space in the Helical
Fusion Reactor
Nagato Yanagi, Takuya Goto, Hitoshi Tamura, Junichi Miyazawa, Akio Sagara
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan

Conceptual design studies of the LHD-type helical fusion reactor FFHR-d1 are progressing
steadfastly. A multi-path strategy is being pursued for securing the design foundation, and it
presently specifies FFHR-d1A as the base option for promoting the 3D engineering design [1].
In d1A, a pair of helical coils (major radius: 15.6 m) has the helical pitch parameter c of 1.20
(the corresponding minor radius: 3.744 m), which was changed from the previous value of 1.25
for the original d1 (minor radius: 3.9 m) [2]. Due to the c decrease, the “blanket space” cp at
the inboard side of the torus is enlarged from 890 mm (for d1) to 950 mm (for d1A). Here cp isdefined as the distance between the innermost layer of the helical coil windings and the ergodic
layer outside the last closed flux surface (LCFS). In order to further improve the structural
integrity and to lower the neutron flux, it is preferable to further enlarge cp.
Four methods were considered for enlarging cp in the previous studies. The first is to choose asmaller c, such as 1.15 for FFHR-2m1 [3], but it gives a smaller minor radius of LCFS (plasma
volume). The second is to set high current density in the helical coil, which intensifies theengineering challenge. The third is to make an outward shift of LCFS (together with themagnetic axis) and cut the remnant ergodic layer by target plates on X-points [3, 4]. In theoutward shifted configuration, the alpha particle confinement and neoclassical transport are
deteriorated, and thus, an inward shifted configuration is employed for FFHR-d1. The fourth isto split the helical coils at the outboard side of the torus which improves the poloidal symmetryof magnetic surfaces [5], but the outward shifted configuration is again the problem.
In 2014, another innovative method was proposed to shed a new light on this problem. The idea
is to vary the current density within the helical coil by subdividing the coil windings into multi
blocks [6]. The proposed method was, however, later determined to be almost infeasible fromthe engineering viewpoint since the current density was unrealistically high. Then, a modified
version of the similar concept was proposed, in which a pair of sub-helical coils, presently
named NITA (Newly Installed Twist Adjustment) coils, is employed at about two times theminor radius of the main helical coils [7]. An oppositely directed current of 5-10% of the main
helical coil current is applied to the NITA coils and this combination effectively lowers c. As
the main helical coils have the original minor radius of FFHR-d1, cp can be resultantly
enlarged to 1100-1200 mm. We should emphasize that the minor radius of LCFS is stillcomparable to that of FFHR-d1A, which is a contrast to choosing a smaller c for the main
helical coils. Presently, the electromagnetic stress analysis is being carried out and the result
will be compared with that for FFHR-d1A to examine the effect of inclusion of NITA coils.
[1] A. Sagara, H. Tamura, T. Tanaka, et al., Fusion Eng. Des. 89 (2014) 2114.
[2] H. Tamura, T. Goto, T. Tanaka, et al., Fusion Eng. Des. 89 (2014) 2336.
[3] A. Sagara, O. Mitarai, S. Imagawa, et al., Fusion Eng. Des. 81 (2006) 2703.
[4] T. Morisaki, S. Imagawa, A. Sagara, O. Motojima, Fusion Eng. Des. 81 (2006) 2749.
[5] N. Yanagi, K. Nishimura, T. Goto, et al., Contrib. Plasma Phys. 50 (2010) 661.
[6] T. Watanabe, N. Yanagi, A. Sagara, Plasma Fusion Res. 9 (2014) 3403089.
[7] T. Watanabe, private communication.

- 125 -

P1-70
Summary of conductor and joint test results of JT-60SA CS and EF coils
using NIFS test facility
T. Obana, K. Takahata, S. Hamaguchi, K. Kizua, H. Murakamia, H. Chikaraishi, H. Noguchi,
T. Kobuchi, S. Moriuchi, S. Imagawa, T. Mito, K. Tsuchiyaa, K. Natsumea, K. Yoshidaa
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanaJapan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, JapanThe JT-60 tokamak is being upgraded to an advanced superconducting tokamak referred to as
the JT-60 super advanced (JT-60SA) at the Japan Atomic Energy Agency (JAEA) Naka site inJapan. [1,2]. The JT-60SA magnet system consists of a central solenoid (CS) coil, 18 toroidal
field (TF) coils and six plasma equilibrium (EF) field coils. Nb3Sn cable-in-conduit (CIC)
conductors are used for the CS coil, and NbTi CIC conductors are used for the TF and EF coils.
JAEA was taking charge of conductor production for the CS and EF coils, and the conductor
production was started from 2008 at Naka site [1,3].
In 2007, the collaborative project between JAEA and the National Institute for Fusion
Science (NIFS) was launched to evaluate the performance of the CIC conductors and conductor
joints using a conductor test facility in NIFS [3,4]. Conductor tests for four types of coilconductor and joint tests for seven types of conductor joints have been conducted for the past
eight years. In this paper, all of the conductor and joint test results are summarized.

[1] K. Yoshida, K. Tsuchiya, K. Kizu, H. Murakami, K. Kamiya, T. Obana, et al., Physica C,
470 (2010) 1727-1733.
[2] H. Murakami, K. Kizu, K. Tsuchiya, Y. Koide, K. Yoshida, T. Obana, et al., IEEE
Transactions on Applied Superconductivity, 24 (3) (2014) 4200205.
[3] K. Kizu, K. Tsuchiya, T. Obana, K. Takahata, R. Hoshi, S. Hamaguchi, et al., Fusion
Engineering and Design, 84 (2009) 1058-1062.
[4] T. Obana, K. Takahata, S. Hamaguch, N. Yanagi, T. Mito, S. Imagawa, et al., Fusion
Engineering and Design, 84 (2009) 1442-1445.

- 126 -

P1-71
Effect of Laval Nozzle in the GAMMA 10 SMBI Experiments
M. M. Islam, Y. Nakashima, S. Kobayashia, Y. Nakanoa, N. Nishinob, K. Hosoi, K. Ichimura,
M. S. Islam, K. Shimizu, K. Fukui, M. Ohuchi, A. Terakado, M. Yoshikawa, J. Kohagura,R. Ikezoe, X. Wang, M. Ichimura and T. Imai
Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577, Japana
Institute of Advanced Energy, Kyoto University, Gokasho, Uji 611-0011, Japanb
Graduate School of Engineering, Hiroshima University, Higashi-Hiroshima, JapanIn GAMMA 10, supersonic molecular beam injection (SMBI) has been carried out at thecentral-cell where the main plasma is confined as shown in Fig.1 [1]. The proper control of gas
fueling is very important to obtain good performance plasma. The results of SMBI experiment withplenum pressure from 0.4 MPa to 2.0 MPa were obtained. The electron line density in thecentral-cell (NLcc) of the plasma was increased during SMBI as well as with the increase ofplenum pressure (Fig.2). In Fig.2, NLcc was also increased before injecting SMBI because ofgas-puffing. Hα intensity was measured by Hα detector at different position along z-axis. The Hαline intensity was increased by SMBI (Fig.3). The neutral particle behavior was investigated basedon 2-dimensional image captured by a high speed camera which is installed at the central-cell(Fig.4). The distribution of the emission intensity were also investigated by the image captured by
the camera as an index of the neutral transport. The full width at half maximum (FWHM) value
with laval nozzle was lower than that of straight nozzle. This means that the effect of laval nozzle
reduces the diffusion of injected hydrogen molecules. Detailed results of SMBI experiments will
be presented in the poster session.

Fig. 1 Overview of GAMMA 10 experimental set-up

Fig. 2 Time responses of line density in central cellhorizontal

verticalY

X

Z

Fig. 3 Variation of Hα with time in central-cell

Z

Fig. 4 2-D image by the high-speed camera[1] K. Hosoi, et al., J. Plasma Fusion Res. 7 (2012) 2402126- 127 -

P1-72
Speedup of shielding current analysis in high-temperature
superconducting film: implementation of H-matrix method
A. Kamitani, T. Takayama, A. Saitoh, H. Nakamuraa
Yamagata University, 4-3-16, Johnan, Yonezawa, Yamagata 992-8510, Japana National Institute of Fusion Science, 322-6, Oroshi, Toki, Gifu 509-5292, Japan

Two types of numerical methods have been investigated for analyzing the shielding
current density in a High-Temperature Superconducting (HTS) film: diﬀerential-algebraicequation method [1] and virtual-voltage method [2]. If an initial-boundary-value problemof the shielding current density is discretized with n nodes, both methods require O(n2 )operations at each time step. In order to further accelerate the virtual-voltage method, theH-matrix method [3–5] is applied to matrix-vector multiplications in its linear-system solver,
GMRES.
A numerical code for calculating the shielding current density has been developed on thebasis of both proposed methods and, by means of the code, the influence of the H-methodon the performance of the shielding current analysis has been numerically investigated.
Conclusions obtained in the present study are summarized as follows.
• When GMRES is implemented as a linear-system solver, the virtual-voltage method
can realize a high-speed shielding current analysis in an HTS film with cracks.
• Acceleration by the H-matrix method is eﬀective especially for a large-sized shielding
current analysis in a high-aspect-ratio HTS film.

[1] A. Kamitani, T. Takayama, A. Saitoh, H. Nakamura, Plasma Fusion Res. 10 (2015)
3401059.
[2] A. Kamitani, T. Takayama, S. Ikuno, IEEE Trans. Magn. 49 (2013) 1877.
[3] S. Kurz, O. Rain, S. Rjasanow, IEEE Trans. Magn. 38 (2002) 412.
[4] Y. Takahashi, C. Matsumoto, S. Wakao, IEEE Trans. Magn. 43 (2007) 1277.
[5] V. Le-Van, B. Bannwarth, A. Carpentier, O. Chadebec, J.M. Guichon, G. Meunier, IEEE
Trans. Magn. 50 (2014) 7010904.

- 128 -

P1-73

Fuelling effect by axial injection of torus plasma into ST plasma

S. Koike, T. Takahashi, N. Mizuguchi1, and O. Mitarai2

Division of Electronics and Informatics, Gunma University, Kiryu, Gunma 376-8515, Japan
1
2

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

Tokai University, 9-1-1 Toroku Higashi-ku, Kumamoto 862-8652, Japan

Development of fuelling technique for supplying fuel particles in the core region of a burning plasma isof great importance to realize nuclear fusion power generation. Experiments of compact torus (CT)injection into a relatively large-scale torus plasma [1] have explored an innovative deep-fuelling method.
However, the experiments have found retard of an accelerated CT plasma by the vertical field of a target
torus plasma is a technological problem. Recently, a new fuelling technique using the merging of two
plasmas with the same plasma current directions (Merging Fueling) has been proposed for a D-3He
spherical tokamak (ST) reactor [2]. In this scheme, a smaller-scale torus plasma produced by the auxiliary
poloidal coils is vertically accelerated toward a main ST plasma by controlling the external field. In a fact,
the C-2 experiment shows two field-reversed configurations (FRCs) are successfully merged after thehead-on collision, and the merged FRC is found to exhibit a better confinement property [3]. If themerging process completes, steady state operation of a fusion reactor is available by intermittent injection.
We study here feasibility of fuelling by a compact-size torus plasma merging with a ST plasma by themagnetohydrodynamic (MHD) simulation code MHD Infrastructure for Plasma Simulation (MIPS) [4].
Detailed results will be shown in our presentation.

[1] L. J. Perkins, S. K. Ho, and J. H. Hammer, Nucl. Fusion 28, 1365 (1988).
[2] O. Mitarai, "Impacts of STOR tokamaks on fusion research", W2-6 (June 17)，Invited talk, CAP
Congress 2015 (University of Alberta, Edmonton, Canada June 15-19, 2015).
[3] M. W. Binderbauer et al., Phys. Rev. Lett. 105, 045003 (2010).
[4] Y. Todo et al., Plasma Fusion Res. 5, S2062 (2010).

- 129 -

P1-74Progress in development of TE31,11 mode gyrotron for ITER EC system
Ryosuke Ikeda, Yasuhisa Oda, Takayuki Kobayashi, Masayuki Terakado,
Ken Kajiwara, Koji Takahashi, Shinichi Moriyama and Keishi Sakamoto
Japan Atomic Energy Agency (JAEA), Naka, Ibaraki 311-0193, Japan
ikeda.ryosuke@jaea.go.jp

A 20 MW electron cyclotron resonance heating and current drive (EC H&CD) system isplanned in ITER. ITER EC system is composed of 24 gyrotrons including superconducting
magnets and matching optics units (MOU), high voltage power supplies, 24 set oftransmission lines, 4 upper port launchers and equatorial port launcher. The required
specifications of a gyrotron are 170 ± 0.3 GHz oscillation frequency, > 1 MW continuous
wave operation and > 50 % electrical efficiency. To suppress a neo-classical tearing mode, 5
kHz high–speed power modulation is required. In addition, to achieve low loss transmission,
HE11 mode purity at the MOU exit demands > 95 %. In JAEA, the gyrotrons with a triode
magnetron injection gun has been developed. In TE31,8 mode gyrotron, the output power, theelectric efficiency and the pulse width achieved 1 MW / 55 % / 800 s and 0.8 MW / 57 % /3600s [1, 2]. The demand performance was almost filled by these results. Moreover, 5 kHz
full power modulation was demonstrated by an electron beam modulation using an anodevoltage switching [3]. To decrease the heat load on the cavity resonator for > 1 MW CW
operation, TE31,11 mode gyrotron is under development. The cavity radius at its straightsection was increased from 17.90 mm to 20.87 mm to excite TE31,11 mode and the cavity wallloading becomes considerably low. Therefore, high power operation of 1.5 MW is expected.
In last campaign, maximum output power and the total electric efficiency were 1.19 MW
and 43%. And, in steady-state operation, a 1000 s oscillation length and output power of 0.5
MW was achieved [4]. In last campaign, electron beam radius injected in a cavity resonator
was 9.1 mm. This value corresponds to peak radius of coupling coefficient between electron
beam and electric field. However, mode coupling with counter-rotating modes such as TE28,12
mode and TE29,12 mode was observed. Especially, TE28,12 mode which is excited in low fieldside interferes with high efficiency oscillation region in TE31,11 mode. To avoid excitation ofcounter-rotating TE28,12 mode, it has extended the electron beam radius to 9.3 mm. As theresults, TE31,11 mode was oscillated in a hard-self excitation region without mode coupling.
Maximum efficiency at 0.6 MW operation reached 50 %. In this campaign, 1.24 MW/45 %/2
sec and 1.13 MW/49 %/2 sec were obtained in high power experiments. In steady-operation,
0.6 MW/1000 sec and 1.0 MW/200 sec were achieved.

[1] K. Sakamoto et al., Nat. phy. 3 (2007) 411.
[2] K. Sakamoto et al., Nucl. Fusion 49 (2009) 095019.
[3] K. Kajiwara et al., Nucl. Fusion 53 (2013) 043013.
[4] R. Ikeda et al., Fusion Eng. Des., High-power and long-pulse operation of TE31,11 mode
gyrotron, ARTICLE in PRESS (2015).

- 130 -

P1-75Characteristics of a Large diameter RF negative hydrogen ion source
Yuko Sasaki1, Sho Takayama1, Haruhisa Nakano2, Atsushi Komuro1, Kazunori Takahashi1and Akira Ando1
1

Department of Electrical Engineering, Tohoku University, Sendai 980-8579, Japan
2
National Institute fir Fusion Science, 322-6, Orishi-cho, Toki, 509-5292, Japane-mail address of submit author: yukosasaki@ecei.tohoku.ac.jp
Negative Neutral Beam Injection system (NNBI) is a powerful tool for fusion plasma heating
and is planned to be installed in ITER. A negative ion source is a key technology to develop thehigh performance NBI. A long-pulsed negative-ion-based NBI is required to be operated with
the D- beam energy of ~1 MeV and beam current of 200 A/m2. From the view point of thestable and long-pulsed operation, development of an radio frequency (RF) negative ion sourceis one of the vigorous research issues [1].
In our previous researches a high density plasma production (~1019 m-3) and the resultant
increase in a negative beam current have been performed in the RF negative ion source driven
with the RF frequency of 100-400 kHz and the RF power of 10-30 kW [2,3,4]. The innerdiameter of the driver region of the source is 70 mm and the axial magnetic field is applied to
improve the source performance.
In this research we have investigated a larger RF source with the inner diameter of the driver
region is 230 mm, which is the same diameter as the ITER NBI. Two solenoids to create theaxial magnetic field are installed in the driver region. Permanent magnets are also installed at
the source back plate and around a diffusion chamber, eliminating plasma loss to the wall.
Efficient plasma production was obtained and effects of Cs vapor evaporation on the ion source
performance will be shown.

[1] U. Frantz, et al. AIP Conference Proc. 1655, 040001 (2015)[2] A.Ando, et al. Rev. Sci. Instrum. 83, 02B122 (2012).
[3] K.Oikawa, et al. Rev. Sci. Instrum. 85, 02B124 (2014).
[4] A.Ando, et al. AIP Conf. Proc. 1390, 322 (2011).

- 131 -

P1-76
Temperature dependence of tritium enrichment by electrolysis with solid
polymer electrolyte
N. Ikemotoa, N. Akataa, b, M. Tanakaa, b, N. Shimac, H. Kakiuchida

SOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292,
Japanb
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanc
Fukushima University, 1 Kanayagawa, Fukushima 960-1296, Japand
Institute for Environmental Science, 1-7 Obuchi-ienomae, Rokkasho 039-3212, Japan
ikemoto.norihiro@LHD.nifs.ac.jp

In the future fusion facilities, tritium will be released from the facility. It is important tounderstand the impact of tritium in the environment. Tritium exists as water in the environment.
However, recently the tritium concentration in environment is almost lower than the detection
limit of a liquid scintillation counter (LSC). Therefore, sample water must be enriched byelectrolysis for tritium measurement. It is known that tritium enrichment is affected by thetemperature of the sample water [1]. The purpose of this study is obtaining a relation between
tritium enrichment and the temperature of the sample water.
In this study, the temperature of sample water is cooled from outside of water tank withcooling chiller. An electrolytic apparatus (TRIPURE XZ001, Permelec Electrode Ltd.) withsolid polymer electrolyte (SPE) is utilized for tritium enrichment. The temperature of sample
water is measured by thermocouple attached near the SPE. The sample water is electrolyzed
from 800 mL to 75 mL. The electrolytic current is 55 A for 35 hours, then it is reduced to 20 A.
The total operational time is approximately 48 hours. Distilled sample water is mixed with
scintillator (ULTIMA GOLD LLT, PerkinElmer) in a polyethylene vial. The tritium
concentration of the sample is measured by low background LSC (LSC-LB5, Aloka).
Fig. 1 shows the relation between the average temperature of sample water and theenrichment factor. The enrichment factor is defined as:
T
Enrichment factor = f ,(1)
Ti
where Ti and Tf are the initial and final
tritium concentration, respectively. The
enrichment factor is increased when the
average temperature is low. In addition, themost stable enrichment is confirmed bycooling to 15 °C. It is suggested that tritium
enrichment by SPE method with coolingwater circulator is effective for high andstable enrichment.

Fig. 1. Relation between the averagetemperature and the enrichment factor.
[1] Y. Ogata et al. J. Radioanal. Nucl. Chem. 255 (2003) 539-541- 132 -

P1-77
Study of tritium release behavior from soil particle
Hiroyuki Datea, Kazunari Katayamaa, Kazuya Furuichia, Tatsuro Hyugab, Toshiharu Takeishib,
Satoshi FukadaaaDepartment of Advanced Engineering, Kyushu University
6-1, Kasuga-koen, Kasuga, Fukuoka, Japan, 816-8580bFaculty of Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka, Japan, 819-0395

Recently, the demand on requirements to assess the risk for severe accidents in nuclear
facilities is increasing and countermeasures for various accidents are being reconsidered. The
spill of tritiated water to the ground is one of accidents which should be assumed because alarge amount water containing tritium is handled in a fusion facility. The transport model oftritiated water under the ground is required from a viewpoint of safety of a fusion reactor andnuclear facilities but studies on tritium behavior in natural soil are limited. In this study, thepresent authors discuss how tritium is trapped in soil particles.
Sample soil of 39.2g was immersed in tritiated water of 2.5×105 Bq/cc for 62 days. 3.0g of wet
tritiated soil was packed in a quartz tube and Ar gas was passed in it for 18 hours at room
temperature in order to dry. After that, soil bed was heated to 1000°C at 5°C/min and heated at
1000°C constant for 30 minute while Ar gas passing. The released tritium was collected in a
double-water bubbler and tritium concentration was measured by LSC. In addition, non-tritiatedsoil was heated on the same condition and released water was measured by moisture meter.
The amount of tritium in water decreased and 12% of them are resorbed in soil. Sample soil
resorbed tritium of 1.2×106 Bq and tritium concentration of tritiated water became 2.2×105
Bq/cc. It is because isotope exchange reaction. After 18 hours Ar gas purging, the mass of soil
decreased 0.318g. The amount of tritium in released water was calculated with 7×104 Bq and it
conformed to amount of tritium collected in bubbler. Fig.1 shows the amount of released tritium
and water content every 100°C. Total amount of released tritium was 6×103Bq. This is 6% of alltritium trapped in soil sample. Generally, it is said that the soil particles have 3 kinds of water:
surface water, interlayer water, and structural water. These water have different release peaks.
Among released tritium, it is thought that surface water and interlayer water had most. However,
it is difficult to recover all tritium resorbed in soil by heating.

2x10

1000
Water Content
Tritium

5

800
600
400

1x105

200
0

~100~200~300~400~500~600~700~800~900~10001000

Tritium[Bq/g]

Water Content[ppm/g]

3x105

0

Temperature[ºC]
Fig. 1 Released tritium and water concent every 100°C- 133 -

P1-78
Evaluation of tritium sorption rate in soil packed bed
by numerical analysis
Kazuya Furuichi, Kazunari Katayama, Hiroyuki Date, Taturo Hyuga, Satoshi Fukada
Department of Advanced Energy Engineering Science, Kyushu University,
6-1, Kasugakoen, Kasuga-shi, Fukuoka 816-8580, Japan

Tritium concentrate in effluent water [kBq/cc]

From a viewpoint of safety, understanding of tritium behavior in the environment is an
important issue for the development of a fusion reactor. Because a large amount of tritiatedwater will be handled in the fusion plant, the accidental spill of tritiated water to the natural soil
around the plant should be assumed. When the tritiated water is penetrating into the ground, apart of tritium is considered to be retained in the soil particles by absorption, adsorption andisotope exchange reaction. In our previous study, the percolation experiment of tritiated water
into the soil packed bed were performed and it was found that a part of tritium is sorbed by theisotope exchange reaction [1]. In this study, a simplified transfer model for tritium in a soil
packed bed was proposed and experimental results were numerically analyzed.
Sample soils ware collected in the Hakozaki campus of Kyushu University. The soils wereseparately packed in the fluoroplastic tube and tritiated water was poured to the top of thepacked bed. The percolated water was drained into a tray which was put on an electric balance.
Weight change of the tray was continually monitored to measure the weight of the effluent
water. The effluent water was periodically sampled and tritium concentration was quantified bya liquid scintillation counter. The detail of experimental system and procedure was reported in[1]
The present authors have been proposed a transfer model for tritiated water vapor in concretematerials [2]. This model was simplified for analysis of tritium behavior in the soil packed bed.
In our previous study [1], it was found that650
the Darcy’s law is valid for the water600
percolation in the soil packed bed.
Input concentration
Additionally, the saturated hydraulic550
conductivity was obtained. This means that500
the percolation behavior of water can be450
calculated by using the obtained hydraulic400
conductivity assuming Darcy’s law. In thisExperimental
350
work, total isotope exchange capacity and the
Calculated
overall sorption rate by isotope exchange
300
0
10
20
30
40
50
reaction in soil particles were assumed andTime [min]tritium concentration in the effluent water inthe tray was calculated. The comparison of Fig.1 Comparison of experiment andexperiment and calculation are shown in calculation for the change of tritium
Fig.1. The overall sorption rate was evaluated concentration in the effluent water in theto be 6×10-4 s-1. By using obtained sorption percolation of tritiated water into the soil
rate and water percolation rate, tritium packed bed.
penetration rate into the ground can be
estimated approximately.
This work is supported by KAKENHI Grant-in-Aid for Young Scientists (A) 25709087.
[1] T. Honda, K. Katayama et al., Fusion Sci. Technol., 67 (2015) 382-385.
[2] K. Furuichi, H. Takata et al., J. Nucl. Mater., 367-370 (2007) 1243-1247.

- 134 -

P1-79Effects of gas-liquid ratios on separative performance of dual-temperaturewater hydrogen chemical exchange column
T. Sugiyamaa, K. Kotohb, K. Munakatac, A. Taguchid, T. Kawanoe, M. Tanakae, N. AkataeaNagoya University, Fro-cho 1, Chikusa-ku, Nagoya 464-8603, Japan
Kyushu University, Moto-oka 744, Nishi-ku, Fukuoka 819-0395, Japanc
Akita University, Tegata-gakuen-machi 1-1, Akita 010-8502, Japand
University of Toyama, Gofuku 3190, Toyama 930-8555, Japane
National Institute for Fusion Science, Oroshi-cho 322-6, Toki, Gifu 509-5292, Japanb

One of the primary components for Water Detritiation System (WDS) of DEMO fuelcycle is Combined Electrolysis and Catalytic Exchange (CECE). A simple CECE process is,however, not feasible to treat the huge amount of tritiated water, because very large electrolyzer
will be required. In the case of WDS for ITER, some combination processes, such as CECE
combined with water distillation or vapor phase chemical exchange, have been proposed andinvestigated [1-3]. Dual temperature system is a typical method to produce a reflux without any
chemical conversion. A dual temperature water-hydrogen chemical exchange (DT-CE) can be,then, a candidate method to combine with CECE, because the process does not require anelectrolyzer. A DT-CE was investigated as the method of deuterium recovery [4]. The optimalpressure of the method explained in the literature was very high as 69 atm and the method
seems to lose the feasibility.
In the present study we propose a system which consists of two liquid phase chemical
exchange (LPCE) columns operated under different pressures each other as well as different
temperatures. The system is named ‘DTDP-CE’ hereinafter. By changing the operatingpressures of the columns each other, we can examine the possibility to reduce the optimalpressures and to accomplish a feasible combination of operating conditions. Although there aresome reports on the system like DTDP-CE for the limited cases of operating conditions [5, 6],variety effects of operating parameters on separative performance are not sufficiently
investigated. The purpose of the present study is to investigate effects of gas-liquid ratios on
separative performance of a DTDP-CE process. Detail results and discussions about effects ofhydrogen-vapor flow ratio as well as hydrogen-water flow ratio on concentration profiles in thecolumns were presented using McCabe-Thiele diagram.
[1] J. M. Miller, S. L. Celovsky et al., Fusion Sci. Technol., 41 (2002) 1077.
[2] Y. Iwai, Y. Misaki, T. Hayashi et al., Fusion Sci. Technol., 41 (2002) 1126.
[3] R. Michling, I. Cristescu, L. Doerr et al., Fusion Eng. Des., 84 (2009) 338.
[4] M. Benedict et al., Nuclear Chemical Engineering 2nd Edition, McGraw-Hill, Inc. (1987).
[5] I.A. Alekseev, S.D. Bondarenko et al., Fusion Eng. Des., 69 (2003) 33.
[6] T.V. Vasyanina, I.A. Alekseev et al., Fusion Eng. Des., 83 (2008) 1451.

- 135 -

P1-80Preparation of hydrophobic platinum catalyst with porous silica supports
for water hydrogen chemical exchange
Y. Morita, T. Sugiyama, K. Sawada, Y. Enokida
Nagoya University, Fro-cho 1, Chikusa-ku, Nagoya 464-8603, Japan

Hydrogen isotope separation by water-hydrogen chemical exchange is one of thecandidate methods for Water Detritiation System. In order to accelerate the isotopic exchange
reaction between hydrogen and water vapour molecules, a hydrophobic platinum catalyst is
indispensable. Many kinds of catalysts have been developed using hydrophobic porous
supports such as polytetrafluoroethylene, polyethylene, polystyrene and styrene-divinylbenzen[1]. In the present study, a porous silica support was employed in order to improve heat
resistance and hydrophobicity under pressurized condition.
Platinum catalysts were prepared using porous silica beads (Fuji silysia chemical Ltd.,CARiACT Q-50) as a support. A number of silane coupli ng agents was tested for hydrophobic
treatment of the beads. Among them a catalyst treated with hexadecyltrimethoxysilane
provided superior performance in hydrophobicity and isotope exchange. The catalyst
performances for isotopic exchange between hydrogen and water vapour molecules were
measured with tritiated water of 1 MBq/kg, and the overall mass transfer coefficient was
obtained as 84±5 s−1. Figure 1 shows a relationship between the overall mass transfer
coefficient k and the alkyl chain length of the silane coupling agents. The mass transfer process
in the catalyst bed was also examined theoretically.
[1] J. P. Butler, J. H. Rolston, W. H. Stevens, A.C.S. Symposium Series, 68 (1978) 93.
100

k [s−1]

80

60
40
20
0
0

3

6
9
12 15
Alkyl chain length

Fig. 1- 136 -

18

P1-81A new pretreatment technique for environmental tritium analysis withmicrowave heating method
N. Akataa,b, M. Tanakaa,b, S. Takayamaa , H. Kakiuchic, T. Tamarid, S. SanoeaNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292,
Japan
c
Institute for Environmental Sciences, 1-7 Ienomae, Obuchi, Rokkasho, Aomori 039-3212, Japand
Kyushu Environmental Evaluation Association, 1-10-1 Matsukadai, Higashi-ku, Fukuoka 813-0004,
Japane
Advanced Industrial Science and Technology, 2266-98 Anagahora, Shimoshidami, Moriyama,
Nagoya, Aichi 463-8560, Japanbakata.naofumi@LHD.nifs.ac.jp
The deuterium plasma experiment is being planned to produce higher performance
plasma using the Large Helical Device (LHD) of National Institute for Fusion Science (NIFS)[1]. In the experiments, a small amount of tritium will be produced by the D(d, p)T reaction indeuterium plasma. Therefore, it is important to understand the tritium concentration inenvironmental samples for radiation safety and environmental impact assessment. Tritium inenvironmental organic samples usually consists of free water tritium (FWT) and organicallybound tritium (OBT). The FWT exists as HTO form, and its behavior is similar to natural water.
On the other hand, OBT is tritium bounded to organic molecules (or carbon). The OBT has alonger biological half-life and higher incorporation to organic materials than the FWT [2]. The
conventional method for FWT and OBT analysis is a liquid scintillation counting method after
freeze-drying and combustion of the sample as summarized in MEXT report [3]. However,pretreatment for FWT and OBT analysis are complicated and time consuming processes over a
week. Thus, we propose the application of microwave heating technique to save time and effortof the pretreatment of plant samples for FWT analysis.
To understand the behavior of the combustion and drying of organic samples, we
conducted TG/DTA analysis of pine needle samples. It is found that the sample has to be heated
up to 100-110 oC for complete drying under N2 gas flow condition. Then, we tried the water
recovery from fresh pine needle samples by the multi-mode microwave heating system. The
recovery rate was achieved more than 97% under these experimental conditions (microwave
output: 1.0kw, time: 30 min, N2 flow: 2L/min). This result shows more preferable for the
pretreatment of organic samples than the conventional method.
[1] A. Komori et al. Fusion Science Technol, 58, 1 (2010)[2] C. Boyer et al., Environmental and Experimental Botany, 67, 34-51 (2009)[3] MEXT, Radioactivity Measurement, Series No. 9 (2002)- 137 -

P1-82
Tritium decontamination from carbon deposited layer by baking
N. Ashikawaa, Y. Torikaib, K. Nishimura and the LHD experiment group
National Institute for Fusion Science, Toki, Gifu, 509-5292, JapanSOKENDAI (The Graduate University for Advanced Studies), Toki, Gifu, 509-5292, Japanb
Toyama University, Toyama, Toyama, 930-8555, Japana

Investigations on hydrogen isotope inventories in plasma facing walls are important withthe view of controls of fuel recycling and in-vessel tritium inventories in fusion devices. But,
removal processes of hydrogen isotopes have not been optimized yet and still serious problems
in ITER and DEMO. In particular, retained hydrogen isotopes in deposition layers are higher
than that in bulk materials. Depth profiles in target materials are different between the retainedhydrogen isotopes originating from energetic hydrogen isotopes during plasma discharges andthe molecular hydrogen isotopes [1]. In tokamaks, high-level tritium retentions were observed
under the dome regions and this phenomenon is the trappings of molecular hydrogen isotopes
into deposition layers. Thus, investigations of the trapped molecular tritium are also required.
In this study, tritium retentions by the molecular gasses into the deposition layers are analyzed,
and the removal methods for these tritium from the deposition layers are investigated.
A target deposition layer, namely sample S2, was produced on stainless steel (SS) 316
target samples which was set near the graphite divertor targets during one experimental
campaign in LHD. On the sample S2, the thickness of deposition layer was 382 nm and theatomic concentrations were 80 %C, and 15 % Fe. The retained hydrogen from the deposition
layers were 1.03x1022 mol./m2 on sample S2．The hydrogen was mainly trapped into thedeposition layers through the main plasma discharges and the glow discharges in LHD. The
hydrogen-carbon (H/C) ratio for the sample S2 was 0.23. The ratio of sp2/sp3 bonded carbon
is 0.1-0.2 measured by X-ray photoelectron spectroscopy (XPS). Based on the ellipsometry
analysis, single C-H bonds in the carbon deposited layers were observed. Four kinds ofanalytical parameters, H/C, sp2/sp3 ratio, Raman spectroscopy, and the ellipsometry show thatthese carbon deposited layers have characterizations of hydrogenated amorphous carbon withsingle C-H bonds [2].
The tritium imaging plate technique is one of the useful tools for tritium measurement. The
LHD deposition layer on sample S2 was exposed to 7% of tritium gasses at 423K for 3 hours.
After the tritium gas exposure, the amount of retained tritium into the deposition layer was
evaluated by -ray-induced X-ray spectrometry (BIXS) and the detected tritium amount was
5.25x1017 molecular/m2. Molecular tritium was confirmed to be trapped near the surface of
deposition layers.
Long-term tritium desorption from the deposition layer after the tritium gas exposure
measured by the tritium imaging plate technique. The target was kept at the room temperaturein the air. The released tritium amount from the deposition layer was about 30% for 3 months.
After that time to one year, the released tritium is negligible. After the observation of long-term
tritium desorption, the baking experiments were done at 368K, 423K, 573K and 773K. The
effective temperatures for the tritium removal are from 425 K to 573 K, and the tritium
desorption reached a background level at 773K.
This study reveals the fundamental processes for tritium decontaminations from carbon
deposited layers by the baking. Molecular tritium trapped near the surface regions of the LHD
deposition layers has a limitation to long-term tritium desorption, and the baking using an
isochronal annealing is effective for the tritium decontamination from the deposition layer.
[1] T. Tanabe, et al., Journal of Nuclear Materials 313 (2003) 478.
[2] N. Ashikawa, et al., JPFR-SERIES 11 (2015) 015.

- 138 -

P1-83
Microstructural and Microchemical Evolution of V-4Cr-4Ti under
Heavy Ion Irradiation
T. Muroga1, H. Watanabe2, T. Miyazawa3, T. Nagasaka1
1

2

National Institute for Fusion Science, 509-5292 Gifu, Japan
Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
3
Japan Atomic Energy Agency, Rokkasho, Aomori, 039-3212, Japan

V-4Cr-4Ti alloy is a promising candidate low activation structural material for fusion blankets.
For this material, it is pointed out that the factor determining its low temperature operation limit ishardening and loss of elongation by irradiation. Since it is known that initial defects formed byirradiation are interstitial dislocation loops at the lower boundary temperature region (<673K), thekinetics of dislocation loop evolution needs to be investigated. It is also pointed out that precipitateshighly enriched with Ti and O are formed in the matrix and on the dislocation sites, which can influence
the dislocation evolution. The precipitation is known to be highly dependent on irradiation temperature.
This study investigates the dislocation and precipitate evolution and their mutual interactions in
V-4Cr-4Ti during heavy ion irradiations.
The material used was NIFS-HEAT-2, a reference V-4Cr-4Ti alloy. The ion irradiations werecarried out with 2.4 MeV Cu2+ ions at 408K, 473K, 573K, 673 and 773K to 0.15 to 10 dpa and at
4.5x10-3 dpa/s at the damage peak depth (800nm from the surface) using Tandem accelerator of
RIAM, Kyushu University. Microstructural and microchemical evolution was characterized by a TEM
and a high resolution STEM-EDS of Kyushu University, respectively.
Dislocation loops were observed at 573K and below. The loop density decreased withtemperature. At 473K the loop growth rate was low. It was not until 10 dpa that tangled dislocation
was observed. Enriched region with Ti or O was not observed in the matrix at 573K and 1 dpa. At
673K a complex dislocation network structure was observed. The contribution of the precipitation to
the complete difference in dislocation structure between 573k and 673K is investigated.

- 139 -

P1-84
Transport properties of high-beta plasmas in the
inward shifted conﬁgurations on LHD
H. Funaba, K.Y. Watanabe, S. Sakakibara, S. Murakami a , R. Seki, S. Ohdachi,
Y. Suzuki, I. Yamada, R. Yasuhara, K. Tanaka, T. Tokuzawa, M. Osakabe, H. Yamada,Y. Takeiri and LHD Experiment Group
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japana Kyoto University, Department of Engineering, Kyoto 606-8501, Japan
funaba@lhd.nifs.ac.jp

The local transport property of high beta plasmas on the Large Helical Device (LHD) is
studied by referring the ISS04 scaling [1], since the plasmas in the low beta range of LHD
have the similar local transport property as χISS04 , which is a modeled transport coeﬃcientwith the same non-dimensional parameter dependence of ISS04. In the previous analysis [2],high beta plasmas were mainly produced in the conditions of the magnetic ﬁeld strength,
B ≤ 0.75 T, the pitch parameter of the helical coils, 1.197 ≤ γc ≤ 1.254 and the magnetic
vac
axis position in vacuum, Rax= 3.50 ∼ 3.90 m. Recently, the high beta experiments ofvac
LHD are mainly carried out with B = 1 T and Rax= 3.56 m in order to produce high betavacplasmas in the lower collisionality region. The conﬁgurations with Rax< 3.60 m are calledvac’inward shifted conﬁgurations’ since the standard Rax = 3.60 m. In this study, it is intended
to investigate how the high beta is achieved by comparing the thermal transport coeﬃcientsin the plasmas with diﬀerent β, γc , heating schemes, fueling methods and so on.
The thermal transport coeﬃcients χexp is evaluated with the TASK4LHD system [3], where
the power balance analysis in the steady state is made by TR-snap which is a modiﬁcation of
one module in TASK code [4], TR by including the 3D equilibrium. Then, χexp is comparedint
int
with grenχχISS04 , where grenχis a renormalization factor for transport coeﬃcients, which corresponds to the renormalization factor, fren in ISS04 and represents the eﬀects of devices
int
or conﬁgurations. The coeﬃcient grenχ
includes the eﬀect of the change of the magnetic
conﬁguration due to the increment in beta.
int
From the ﬁrst results, the ratio χexp /(grenχχISS04 ) is compared with unity, where the transport
property is similar to the ISS04 scaling, at the normalized average minor radius ρ = 0.5, 0.7
and 0.9. In a high beta plasma of the volume-averaged beta value, hβi = 3.4 % which were
vac
produced in the magnetic conﬁguration of B = 1 T, γc = 1.254 and Rax= 3.56 m and with
ISS04
exp
int
) becomes smaller than unity at ρ = 0.5 and 0.7. This
the gas-puﬀ fueling, χ /(grenχ χ
indicates reduction in thermal transport at main plasma region in the beta range around
hβi ' 3 %.
[1] H. Yamada et al., Nucl. Fusion 45, 1684 (2005).
[2] H. Funaba, K.Y. Watanabe, et al., Plasma Fusion Res., 3, 022 (2008).
[3] R. Seki, H. Funaba, et al., Plasma Fusion Res., 6, 2402081 (2011).
[4] A. Fukuyama, et al., Proc. of 20th IAEA Fusion Energy Conf.
(Villamoura, Portugal, 2004) IAEA-CSP-25/CD/TH/P2-3.

- 140 -

P1-85Heat Transport Characteristics of High Temperature Discharges in LHD- Dependence on Hydrogen/Helium Ratio K. Nagaoka1,2, H. Takahashi1,2, K. Tanaka1, M. Osakabe1,2, S. Murakami3, M. Yokoyama1, K.
Fujii4, H. Nakano1,2, H. Yamada1,2, Y. Takeiri1,2, K. Ida1,2, M. Yoshinuma1,2 and the LHD
experiment group
1National Institute for Fusion Science, Toki, 509-5292, Japan

2SOKENDAI(The Graduate University for Advanced Studies), Toki, 509-5292, Japan
3Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan

4Department of Mechanical Engineering and Science, Kyoto University, Kyoto 615-8540, Japan

Deuterium experiment is planned in the Large Helical Device (LHD) project, and the isotope
effect is one of the most important subjects in the deuterium experiment. The ion heat transport
is expected to be improved by the isotope effect due to the reduction of damping rate of zonal
flow [1].
In order to investigate the physics mechanism of the isotope effect, the comparison of the ionheat transports of high temperature discharges (ion ITB plasmas) [2-3] between the helium
dominated and hydrogen dominated plasmas in LHD. The central ion temperature is higher inhelium dominated plasma than that in hydrogen dominated plasmas, while electron
temperatures are almost identical between them, which are shown in Fig. 1. The transport
analysis indicates that the ion heat transport increases with the ratio of hydrogen ions
(nH/nH+nHe), while the turbulent characteristics are almost unchanged. The heating efficiency,
impurity content, neutral density profile will be presented in detail. The comparison of heattransport in the electron ITB plasma will be also discussed in the paper.
[1] T. H. Watanabe, et al., Nucl. Fusion 51 123003 (2011).
[2] K. Nagaoka, et al., Nuclear Fusion 51 083022 (2011).
[3] H. Takahashi, et al., Nuclear Fusion 53 073034 (2013).

6

T

i

[keV]
4

Te

[keV]
2

0

0

0.2

0.4
0.6
H/(H+He)

0.8

1

Figure 1. The central ion-temperature and electron-temperature as a function ofhydrogen ration (nH/nH+nHe). The plasma density is ne =1.3x1019m-3.

- 141 -

P1-86
Analysis of LHD NBI-plasma interaction with upgraded FIT3D module of
TASK3D-a analysis transport suite
P. Vincenzi1,2, S. Murakami3, M. Osakabe4, R. Seki4, M. Yokoyama4, T. Bolzonella1
1Consorzio RFX, Corso Stati Uniti 4 - 35127 Padova, Italy
2
Università degli Studi di Padova, 35122 Padova, Italy
3
Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan
4
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

LHD [1] is a superconducting helical device running at NIFS. The experiment is equipped withdifferent auxiliary heating systems such as electron cyclotron resonance heating (ECRH), ion
cyclotron resonance heating (ICRH) and neutral beam injection (NBI). 5 NBI systems currently
working with H gas and carrying a maximum total power of 28MW are installed: 2
perpendicular positive injectors (40-50 keV) and 3 tangential negative injectors (180-190 keV).
At present it is the only machine worldwide running negative NBI systems.
TASK3D-a analysis suite [2] is a tool for interpretative transport analysis of LHD experiments,
built for the analysis of hydrogen and multi-ion species plasma [3, 4]. FIT3D is the module of
TASK3D-a code which performs the analysis of LHD NBI-plasma interaction, calculatingradial profiles of NBI absorbed power, beam pressure, beam source and induced momentum.
LHD will be run with deuterium (D) gas in the next future instead of the current hydrogen (H)experiments, and at least 3 of the NBI systems will be upgraded to D neutral particle injection.
For this reason an upgrade of TASK3D-a code has started to enable the analysis of D
experiments. In particular FIT3D NBI code has been updated [5] with a more recent andcomprehensive ionization cross section formula for treating NBI neutral particle ionization
process by plasma and impurities [6], suitable for the case of D injection in D plasma. A routine
which estimates the neutron production from background plasma and beam-plasma fusion
reactions has been implemented too.
The upgrade of FIT3D with the improved ionization cross section allows also a detailed analysis
of H discharges with relevant presence of He impurity. In recent similarity H experiments withdifferent concentration of He, better confinement properties have been observed with He
majority. The upgraded FIT3D code is here used to investigate how the different backgroundgas composition influences the ionization of the NBI particles, in order to understand thecontribution of NBI power deposition on the increased general confinement properties. First
results indicate that the ionization of the injected neutral particles is not dramatically affected by
the gas composition change and the ionization of the NBI does not seem to play a direct role in
the observed confinement properties.
[1] O. Motojima et al, Nucl. Fusion 40 (2000) 599[2] M. Yokoyama et al., Plasma Fusion Res. 8 (2013) 2403016[3] A. Sakai et al., Plasma Fusion Res. 9, (2014) 3403124[4] H. Yamaguchi and S. Murakami, Plasma Fusion Res. 9 (2014) 3403127[5] P. Vincenzi et al., 42nd EPS Conf. on Plasma Physics (Lisbon, Portugal, 2015), P1.150,
http://ocs.ciemat.es/EPS2015PAP/pdf/P1.150.pdf[6] S. Suzuki et al, Plasma Phys. Control. Fusion 40 (1998) 2097

- 142 -

P1-87
Design study for high spatial resolution of beam emission spectroscopy
viewing from 6-O port on large helical device
H. Nakano, M. Yoshinuma, K. Ida
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Beam emission spectroscopy (BES) is one of the powerful diagnostics to measure local densitywith high time resolution for turbulence and transport studies in toroidal plasmas. BES systems
has been installed several toroidal devices (PBX-M, DIII-D, CHS, LHD, etc.). Sample volume(SV) is defined by Doppler-shifted H-alpha emission (beam emission) region where line ofsight (LOS) and neutral beam cross in the plasma. High spatial resolution of minor radius canbe performed by aligning the SV along magnetic flux surface. To detect small scaled fluctuation,
the SV has to be also aligned with magnetic field line. In BES on large helical device (LHD),
although low frequency mainly less than 10 kHz and large scale fluctuations have been
observed with some BES systems, high frequency and small scale fluctuations have not been
detected. Candidates of reasons for this are followings: the magnetic filed line curvaturechanges along toroidal direction, and accessibility of LOS to the plasma is not high due tocomplicate vacuum chamber and large distance between the plasma and ports. Recently, we
designed a new BES system where the emission from the tangential neutral beam of beam line
3 (BL3) on 7-T port in the plasma was viewing from the 6-O port through mirrors. The LOS ofthis BES was set mostly along the magnetic field in periphery plasma. Measure region is limited
until outer area of 0.8 in normalized minor radius due to prevent the final mirror from touchingwith main plasma. The calculated spatial resolution was around 0.03 in normalized minor radius.
An angle between the LOS and BL3 is 32 degree. When the beam is injected with energy of
188 keV which is maximum beam energy, the Doppler shift is 11 nm which is enough far toseparate the beam emission from background H-alpha.

- 143 -

P1-88Preliminary design of an imaging bolometer for W7X
Byron J. Petersona,b, Daihong Zhangc, Marcin Jakubowskic, Yuhe Fengc, Ryuichi Sanoa,
Kiyofumi Mukaia,b and the W7X experiment group.
ab

National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advance Studies), 322-6 Oroshi-cho, Toki 509-5292, Japanc
Max Planck Institue for Plasma Physics, Greifswald, Germany

The InfraRed imaging Video Bolometer (IRVB) measures plasma radiation images, which
can be directly compared with the results of impurity radiation models to provide aquantitative understanding of the spatial distribution of the plasma radiation [1]. In athree-dimensional (3D) device such as W7X, the two-dimensional (2D) images provided by
the IRVB are necessary to understand the complicated divertor radiation pattern. The
ten-fold W7X divertor is divided into two sections, designated as the ‘low iota’ divertor andthe ‘high iota’ divertor, referring to the magnetic configuration under which the divertor flux
to those sections is dominant. In this paper we give information on the design of the imaging
bolometer for the ‘high iota’ divertor from the AEM30 port of W7X. This design assumes a
640 x 480 pixel, 30 fps, 100 mK, IR camera imaging the IRVB foil. The bolometer camera
consists of a 9 cm x 7 cm x 2 micron thick Pt foil blackened with graphite which is located
119 mm from a 5.2 mm x 5.2 mm aperture. This provides a view of almost the entire ‘highiota’ divertor as shown in Figure 1. The 8 cm x 6 cm central section of the foil is divided into
20 x 15 IRVB channels each consisting of 726 IR camera pixels. Using a previously derived
expression [1] the noise equivalent power of the IRVB is given as 90 microwatts/cm2.
Assuming 5 MW of radiated power uniformly emanating from the 30 m3 plasma, the signal
level is estimate as 3 mW/cm2 giving a signal to noise ratio of 33. In the paper more details
and ongoing work on the design of an IRVB for the ‘low iota’ divertor region andcorresponding synthetic images derived from impurity transport models will be presented.
This work was supported by NIFS budget code ULHH026 and the NIFS International
Collaboration budget.
[1] B. J. Peterson et al., Rev.
Sci. Instrum. 74 (2003) 2040.

FIG. 1. Designed field of view of IRVB at Port AEM30 in
W7X. The inner blue rectangle corresponds to the 6 cm x 8
cm central portion of the foil.
l- 144 -

P1-89Development of in-situ calibration system for foil detector of infrared
imaging video bolometer in Large Helical Device
K. Mukaia,b, B. J. Petersona,b, R. SanoaabNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advance Studies), 322-6 Oroshi-cho, Toki 509-5292, JapanThe InfraRed imaging Video Bolometer (IRVB) is a useful diagnostic for the measurement ofplasma radiation profiles [1] to investigate phenomena such as plasma detachment [2]. The
fundamental schematic of an IRVB is a combination of a pinhole camera and an IR camera. The
profiles of the plasma radiation are collimated by an aperture (e.g. 8 mm × 8 mm) in the pinhole
camera and are projected onto a thin platinum foil (e.g. 130 mm × 100 mm, t = 2.5 µm) as a
two-dimensional temperature distribution. The IRVB has many channels by dividing the large
foil into many bolometer pixels (~ 1000). The foil detector is coated by carbon on both sides toincrease the emissivity, then, the distribution can be observed from the opposite side of the foilusing an IR camera as an IR image. Considering the heat-diffusion effect of the foil, the heatcharacteristics of each bolometer pixel should be calibrated to estimate the radiation profile
using the two-dimensional heat diffusion equation. For the application of the IRVB
measurement to the neutron environment of deuterium plasma experiments in the Large Helical
Device (LHD), the effect of the neutron irradiation on the heat characteristics should be
investigated. In this study, an in-situ calibration system was developed.
The schematic of the in-situ calibration system [3] is shown in Figure 1. A periscope system
must be applied to protect the IR camera detector from the damage by the direct irradiation of
X-rays, neutrons, and gammas from the plasma. A He-Ne laser as the known radiation powersource is injected to the foil. Laser
Power meter,
Mirror with
irradiation points can be scanned using a
Beam profiler
biaxial goniometer
Movablemirror with two motorized goniometers,(SIGMAKOKI/
Laser
mirror
GOHTM-40A60,
which correspond to the center of the
GOHTM-40A75)
each bolometer pixel. The visible laser<Removable unit>
can be transmitted to the foil and the IRIR
TS hot mirror
radiation from the foil can be reflectedcamera(Edmund Optics/
to the IR camera using a hot mirror since
45° 101 × 127)
it has the transmittance of > 85% for
Neutron shield
Window
visible light and the reflectance of >95% for IR signal. The results of the
Vacuum vessel
characteristic tests of importantcomponents and the effect of uniform
and reproducible graphite coatings byvacuum evaporation on the heatcharacteristics will be presented.
[1] B. J. Peterson et al., Rev. Sci.
Instrum. 74 (2003) 2040.
[2] K. Mukai et al., Nucl. Fusion 55(2015) 083016.
[3] K. Mukai et al., Rev. Sci. Instrum.
85 (2014) 11E435.

Pt foil

Aperture

FIG. 1. Schematic of in-situ calibration system for IRVB
foil detector. Thin arrows and thick arrows indicate pathsof visible light and IR, respectively.

- 145 -

P1-90
Improvement of 3D radiation imaging using Euclidean distance from model
phantom in LHD
Ryuichi Sano1,2, Masaru Teranishi3, Naofumi Iwama1, Masahiro Kobayashi1,4,
Byron J. Peterson1,4, Kiyofumi Mukai1,41National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan
2
JSPS Research Fellow
3
Hiroshima Institute of Technology, 2-1-1, Miyake, Saeki-ku, Hiroshima Japan
4
SOKENDAI (The Graduate University for Advanced Studies), Hayama, Kanagawa 240-0193, Japan

Radiation which is emitted from impurities is one of the major paths of power loss fromfusion plasmas. Heat load reduction on the divertor in future devices requires the detachedplasma produced by the enhancement of the radiation loss from the divertor plasma. Because of
a 3D complicate structure of radiation from the outside of the last closed flux surface (LCFS), a
3D understanding of radiation is necessary for radiation control. Therefore, the study of 3D
dynamic behavior of the plasma radiation is important for the control of fusion plasmas. In
LHD, 3D radiation imaging using four InfraRed imaging Video Bolometers (IRVBs) andstandard solvers for inverse problems had been developed. Current reconstructed images by 3D
radiation imaging still have artifacts caused by the ill-conditions of Field of View (FoV) for the
IRVBs. To avoid this kind of artifact, mechanical and numerical approaches for improvementof the 3D radiation imaging are in advance. The mechanical approach is an addition of IRVB.
Numerical approach is the use of alternative solvers for image reconstruction. In this study, analternative solver using model phantoms is applied for the 3D radiation imaging andnumerically examined.
The alternative solver employs model phantoms in the constraint for image
reconstruction. In the solver, the Euclidean distance between the model phantom and thereconstructed image is employed as a penalty function. Image reconstructions of a simulated
radiation phantom by EMC3-EIRENE have been carried out using various model phantoms.
The image reconstruction using an appropriate model phantom, such as the simulated phantom
itself, provides a sufficiently smaller amount of artifacts. Even with use of inaccurate model
phantoms such as a simple elliptic cylinder, the solver also provide smaller amounts of artifacts
than the standard solver. These results indicate that the alternative solver works well in the 3D
radiation imaging. The method of model selection for image reconstruction from experimental
data will also be presented at the meeting.
[1] B. J. Peterson et al., Rev. Sci. Instrum. 74, 2040 (2003).
[2] R. Sano et al., Plasma Fusion Res. 7, 2402041(2013).
[3] R. Sano et al., Plasma Science, IEEE Transactions on, 42, 10, 2860 (2014).

- 146 -

P1-91
Design and evaluation of broad-band high pass filter
using metal perforated plates with circular holes
A. Tani, M. Koga, H. Tsuchiyaa, Y. Nagayamaa
Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji 671-2280, Japana
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japane-mail: er15e007@steng.u-hyogo.ac.jp

ECE (Electron Cyclotron Emission) measurement is useful method in observation ofelectron temperature profile and its local fluctuations. Generally ECE is detected by heterodyne
type filter bank radiometer. The frequency band of radiometer is not broad. It is assumed that
two radiometer systems with different local frequency are used for detection of the broad-band
ECE in the high-magnetic devices. It is necessary to divide the RF with high precision before
radiometer detection to avoid the gain loss. In the case of LHD (Large Helical Device),
frequency range of ECE is 50GHz-150GHz. The broad-band high pass filter for such frequency(E-band, W-band) is not commercially available. Therefore, we have developed the broad-band
high pass filter originally. The schematic of the filter is shown in Fig.1.
The purpose of this study is to design the porous metal plane type broad-band filters (cutoff frequency is 44GHz), and to evaluate their characteristics by 3D-numerical electromagnetic
field analysis (E-field). The validity of the numerical analysis is guaranteed by the good
agreement with calculated result and measured results shown as fig.2. It is found that thefrequency characteristics depend on the filter thickness (see fig.3), and the appropriate thickness
is in the range of 0-0.5mm.

Fig.1 Schematic of the metal perforated plate with circular holes.
1

Reflectance

Measured value

0.5

Calculated value
0
20

40

60

Frequency[GHz]Fig.2 Frequency characteristic of calculated
and measured reflectance.
(Filter thickness is 1mm.)

Fig.3 Filter thickness dependence of reflectance.

- 147 -

P1-92Development of Electron Bernstein Emission Diagnostic for Heliotron J andLHD
K. Nagasaki, H. Igamia, G. M. Weir, Y. Nakamurab, S. Kamiokab, K. Sakamoto, H. Okada, T.
Minami, S. Kado, S. Kobayashi, S. Yamamoto, S. Ohshima, S. Konoshima, N. Kenmochib, Y.
Ohtanib, F. Volpec, N. Marushchenkod, S. Kuboa,e, Y. Gotoe, T. Mizuuchi
Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011, Japan
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
Graduate School of Energy Science, Kyoto University, Uji, Kyoto 611-0011, Japanc
Department of Applied Physics and Applied Mathematics, Columbia University, New York, USAdMax-Planck-Institut für Plasmaphysik, EURATOM Association, D-17491, Greifswald, GermanyeDepartment of Energy Engineering and Science, Nagoya University, Nagoya, 464-8603, Japanae-mail address: nagasaki.kazunobu.4x@kyoto-u.ac.jp

Use of electron Bernstein Waves (EBWs) is attractive for heating/current drive, electron
temperature diagnostic [1] and magnetic pitch angle profile measurement [2] at high densities
where electron cyclotron waves are not accessible. The EBWs have advantage of no density
limit and high optical thickness even at low electron temperature. We have been studying
Electron Bernstein Emission (EBE) in the Heliotron J and LHD devices in order to measureelectron temperature at high density plasmas. A ray tracing code, KRAY, has been developed,
which evaluates the power absorption and emission profiles via O-X-B process. The codesolves the radiative transfer equations under geometrical optics approximation in thethree-dimensional magnetic field structure [3]. The calculation results in the Heliotron J
configuration show that the existing EC launcher antenna has an access window for the O-X
conversion for Ka-band (24-42 GHz), and the radiation is located at the core region, r/a0.5,
for B=1.25 T and ne=21019 m-3. In LHD, a ray tracing calculation also shows that a current
horizontal ECRH launcher has an access window for the O-X conversion for the same
frequency band. These indicate that a radiometer of Ka-band frequency range is applicable toboth the Heliotron J and LHD devices. The designed EBE measurement system consists of a
Gaussian optics antenna, a waveguide transmission line and a heterodyne radiometer. As theradiometer is connected to the transmission line and steerable antenna of current ECRH/ECCD
system, angular scan is possible to find the best conversion efficiency to measure the EBW
radiation. Accessing the EBE by means of B-X-O conversion requires the plasma densities overaround 11019 m-3 which are routinely obtained in the Heliotron J and LHD devices. The status
of EBE measurement setup will be presented.
[1] F. Volpe, H. P. Laqua and the W7-AS Team, Rev. Sci. Instrum. 74 (2003) 1409
[2] F. Volpe, Rev. Sci. Instrum. 81 (2010) 10D905[3] K. Nagasaki, N. Yanagi, Plasma Phys. Control. Fusion 44 (2002) 409- 148 -

P1-93
Validation of the Digital Correlation ECE measurement techniqueusing low frequency fluctuation in LHD
H. Tsuchiya, S. Inagakia, T. Tokuzawa, N. Tamura, Y. NagayamaaNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
Research Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan

We have proposed and developed the Digital Correlation Electron Cyclotron Emission(DCECE) Measurement Technique [1] with a Giga Hertz sampling digitizer [2] in LHD. In theconventional radiometer type ECE measurement [3], the power detector with filter bank forintermediate frequency (IF) is used. In case of the ECE measurement in the magnetized plasma,
generally, the frequency band is up to tens of Giga Hertz. Instead of filter bank system, the IF
digitizing technique is applied for DCECE. The conceptual diagram of DCECE is shown in
Fig.1. One of the advantages of this technique is to enable to choose the spatial and temporal
resolutions after the plasma experiment. This characteristic is considered to be useful for the
study of the multiscale dynamics which should determine the whole confinement in plasmas.
The IF digitizing technique was applied to the ECE of LHD plasmas. We can reconstruct
the electron temperature flucutuaion profile and compare the analysis results of conventional
radiometer type ECE and DCECE about 5 kHz flucuation as shown in Fig.2. The data length of
DCECE is 51 msec. The sampling rate is 10GHz. The IF resolution is 166MHz whichcorrespods to 3mm resolution in radial direction. It is confirmed that several kHz fluctuation
with strong correlation is detected more
sharply than conventional radiometer ECE. It
is also found that the aspect of fourie analysis
result depends not only on the sampling rate ofdigitizer but also the parameter of DCECE
analysis such as IF resolotion and time
resolution. In this presentation, the parameter
dependence will be shown as example of theanalysis results of actual LHD plasma data.

Fig.2 The coherence profile of Te fluctuationby conventional radiometer type ECE and
DCECE. The profile by DCECE (IF= 0.5 to
4.5GHz) is drawn on the profile byradiometer ECE (IF=-25GHz to 17.5GHz).
The IF resolution (left axis) is 166MHz in
DCECE, and 1GHz in radiometer ECE. The
radial position R[m] (right axis) is derived
from ECE frequency.

Fig.1 Conceptual diagram of DCECE
analysis.

[1] H. Tsuchiya, et.al., Plasma Fusion Res. 9, (2014) 021.
[2] http://teledynelecroy.com/[3] Y. Nagayama, et al., Rev. Sci. Instrum. 70, (1999) 1021

- 149 -

P1-94Development of OXB mode conversion window searching method using
inverse emission process in the LHD
Y. Gotoa , S. Kuboa,b , H. Igamib , M. Nishiurac , T. Shimozumab , Y. Yoshimurab , H. Takahashib , T.
Tsujimurab , R. Makinob , T. Mutohb and the LHD experiment groupa Dept. of Energy Eng. and Sci., Nagoya Univ., Furo-cho. Chikusa-ku. Nagoya, 464-8603, Japanb National Institute for Fusion Science,322-6 Oroshi-cho, Toki, 509-5292, Japanc Graduate School of Frontier Sciences, The University of Tokyo, Kashiwanoha, Kashiwa, 277-8561, Japangoto-yuuki14@ees.nagoya-u.ac.jp

Intensity [a.u.]

Intensity [a.u.]

Tf [m]- 150 -

Steering position of the Rf direction [m]Electron cyclotron resonance heating (ECRH) using an Electron Bernstein Wave (EBW) is aneﬀective method for the RF heating of the over-dense plasma. The EBW can be excited via theordinary(O)-extraordinary(X)-EBW(B) (OXB) mode conversion process from low field side. The
OX mode conversion window is known to be narrow because the conversion rate depends on the
propagating angle of the EC wave. In the Large Helical Device (LHD), EBW heating experiments
using the OXB mode conversion process have been carried out, but the achieved heating eﬃciency
had been stayed at almost 10 percent. For improving the heating performance and the experimental
eﬃciency, it is essential to tune the injection direction precisely to aim the OXB mode conversion
window experimentally. The contour lines in Fig. 1(a) indicate the calculated OX mode conversion
eﬃciency projected on a lower port (1.5L port) antenna target plane. Since a thermally excited
EBW in the plasma core region can be emitted as the O-mode via the inverse OXB mode conversion mechanism, a method to find out the window by measuring the O-mode emission intensity
near the heating frequency using a fast steerable antenna is developed. This method can be used
as a reference to execute a real time optimization of the injection angle in the future. As a firststep, a change of the emission intensity near 77 GHz O-mode was measured by scanning the receiving angle. Figure 1(b) shows the time variation of the radiation intensity when the antenna isscanned along the path indicated by a straight line in Figure 1(a). The increase in the radiation
intensity near this region is confirmed experimentally when an over-dence plasma is sustained.
Antenna scan time is shifted by 0.4 sec for
1.2
50
4both shots, and the increase of the radi#127651#127653 3.6
1
40
0.4s
ation intensity is shifted correspondingly.
127651 vs 127653
50
Since plasma parameters for both shots are
0.8
30
3.2
40
kept identical and constant from t=4.7 to
0.6
20
2.8
30
6.3 sec, we can conclude this radiation is
0.4
10
2.4
20
distributed spatially. A ray-tracing code0.2
02
Over-dense10
is extended to simulate the electron cy0-10
1.6
2.8 3 3.2 3.4 3.6
2
3
4
5
6
7
8
clotron emission (ECE) including the emis0
Rf [m]
Time [s]sion originated from the EBW (EBE) to un(a)
(b)
-10
2
3
4
5
6
7
8
derstand the observed change in the radiaTime [s]tion using experimentally obtained plasma Figure 1: (a) Calculated mode conversion eﬃciency
profiles for each time slice during the an- and steering path. (b) Time variation of the radiation
for the shots 127651 and 127653.
tenna steering.

P1-95Magnetic flux surface measurements on TOKASTAR-2
Y.Shimooka, H. Arimoto, T. Fujita, A. Okamoto, T. Sakito, H. Itou,
K. Muraoka, R. Sugioka, K. Yasuda, R. Yokoyama
Department of Engineering and Science, Nagoya University,
Furo-cho, Chikusa-ku, Nagoya 464-8603, Japanshimooka-yuusuke14@ees.nagoya-u.ac.jp

TOKASTAR-2 device was constructed to study the influence of helical magnetic field
application to the tokamak plasma and the influence of plasma current to helical plasma [1]. It
has two coil systems, one for tokamak and the other for stellarator.
The stellarator coil system is designed to make closed magnetic flux surfaces without any
plasma current. However, due to the influence of manufactural accuracy and installational
accuracy of coils, the location and shape of the helical magnetic flux surface can be deviatedfrom the analytical one. The object of this study is to measure the magnetic flux surfaces
experimentally.
There are mainly two methods to measure magnetic flux surface, one using a fluorescentscreen [2] and the other using a probe [3] for detecting electrons emitted by an electron gun.
The method using a fluorescent screen can trace a magnetic flux surface faster than the otherone and does not need a mechanical driving system. So, we examined whether the method
using a fluorescent screen can be applied to TOKASTAR-2.
Figure 1 shows the conditions for detecting fluorescence in the preliminary experiment donein a test chamber without magnetic field. P15 and P24 (ZnO:Zn) are used for the fluorescenceagent. We detected fluorescence by eye. It was found that the electron energy needed to belarger than about 80 eV to detect the fluorescence of P15 and P24.
Figure 2 shows the orbit of electrons with different energies in the typical magnetic
configuration condition. It was found that the orbit of the electron with 80 eV differs largely
from the magnetic flux surface. From these results, we concluded that the method using afluorescent screen cannot be applied to TOKASTAR-2.
At the present, we are preparing the method using a probe. It is expected that the electronenergy needed for this method is lower than for the method using a fluorescent screen. The
required electron energy will be examined in a test chamber while a mechanical driving systemof the probe will be developed.

Fig. 2. Orbit of electrons in
the typical magnetic configuration condition.
condition.
[1] K. Yamazaki et al., J. Plasma Fusion Res. SERIES 8 1044 (2009).
[2] R.J. Colchin et al., Review of Scientific Instruments 60 2680 (1989).
[3] R.M. Sinclair et al., Review of Scientific Instruments 41 1552 (1970).

Fig. 1. Conditions for detecting fluorescencein the preliminary experiment.

- 151 -

P1-96The development of 2D potential profile measurementwith heavy ion beam probe on the Large Helical Device
A. Shimizua, T. Idoa,b, M. Nishiurac, H. Takahashia, H. Igamia,
Y. Yoshimuraa, S. Kuboa,b,d, T. Shimozumaa, S. Katoa, M. Yokoyamaa,dand the LHD Experiment GroupaNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
Graduate School of Engineering, Nogoya University, Chikusa, Nagoya 464-8603, Japanc
Graduate School of Frontier Science, The University of Tokyo, Chiba 277-8561, Japand
Graduate University for Adcanced Studies (Sokendai), 322-6 Oroshi-cho, Toki 509-5292, Japanbe-mail: akihiro@nifs.ac.jp

Heavy Ion Beam Probe (HIBP) is a powerful diagnostic tool to study physics related to theelectric field in magnetically confined plasmas. By using this tool, we can measure equilibrium
and fluctuation of electrostatic potential, and density fluctuation simultaneously in the coreregion of high temperature plasma. For this unique advantage, many HIBP diagnostic systems
have been installed on various devices, such as tokamaks, reversed field pinches, mirror
devices, and helical devices.
In the Large Helical Device (LHD), HIBP system was installed and has been developed to
study Er formation physics in the plasma [1]. Although the large size (R ~ 3.9 m, a ~ 0.6 m) andstrong magnetic field strength (~3 T) of LHD increase the difficulty of HIBP measurements, wehave successfully measured equilibrium potential, and fluctuations of potential and density, in
LHD by using tandem accelerator and sensitive micro channel plate detector [2,3].
For HIBP diagnostic, by changing the incident angle of the probe beam, observation point canbe changed to obtain one-dimensional profile of plasma potential. If the probe beam energy isadditionally changed, two-dimensional potential profile can be measured. However, in
LHD-HIBP system, the beam transport line is long (~ 20 m), therefore, beam orbit in the beamtransport line has to be adjusted, when the beam energy is changed. This work requires
adjusting many deflector voltages and consumes a lot of time. In order to reduce the time forbeam adjustment, a PC-based automatic beam adjustment system was developed. By using thissystem, the alignment of the beam in the beam line can be completed during a discharge cycleof LHD (three minutes), and two-dimensional equilibrium electrostatic potential profile was
successfully measured by changing the beam energy shot-to-shot [4]. To measure thetwo-dimensional potential profile in magnetic islands, trial experiments were conducted byapplying rmp field. Although clear island structure was not obtained, a partial flat area wasseen in the potential profile. Details of the automatic beam adjustment system and recent
analysis of 2-D potential measurement experiments are reported.
[1] T. Ido, A. Shimizu, M. Nishiura, A. Nishizawa, S. Katoh, K. Tsukada, M. Yokota, H.
Ogawa, T. Inoue, Y. Hamada et al., Rev. of Sci. Instrum. 77 (2006) 10F523.
[2] T. Ido, A. Shimizu, M. Nishiura, S. Nakamura, S. Katoh, H. Nakano, Y. Yoshimura, K. Toi,
K. Ida, M. Yoshinuma et al., Nucl. Fusion 51 (2011) 73046.
[3] A. Shimizu, T. Ido, M. Nishiura, R. Makino, M. Yokoyama, H. Takahashi, H. Igami, Y.
Yoshimura, S. Kubo et al., Plasma and Fusion Res. 8 (2013) 2402122.
[4] A. Shimizu, T. Ido, M. Kurachi, R. Makino, M. Nishiura et al., Rev. Sci. Instrum. 85 (2014)
11D853.

- 152 -

P1-97
X-Y-t Edge Plasma Reconstruction with Digital Imaging Technique
M. Irie, M. Kubo-Irieaa

Waseda University,3-4-1Ohkubo,Shinjyuku Tokyo 169-8555, JapanThe Open University of Japan, 2-11, Wakaba, Mihamaku, Chiba 261-8586, JapanThe striking high speed CCD camera technique developed under Dr. A. SYKES in late ‘80s, in
START device (Small Tight Aspect Ratio Tokamak : CULHAM laboratory) established the
new era in the history of plasma boundary diagnostics.
The refined system has been applied to the spherical tokamaks in early last decade by Prof. N.
Nishino and explains the “Actual Plasma Behaviour” in critical issues of the Fusion
Engineering, e.g. “L to H transition”, “the Edge Local Mode”.
In 2013, Dr. L. Zang [1] monitored the movement of edge filamentary structure during
Supersonic Molecular-Beam Injection in Heliotron–J device run under Prof. T. Mizuuchi.
In this conference, we would like to mine further quantitative information from these movie
data with the state of art “Digital Imaging Technique”. This is the third approach to the Fusion
Research based on the “Experiment-Simulation” in good old days established by Institute of
Fusion Theory: Hiroshima Simulation Group, Prof. Tetsuya SATO and his colleagues.
Reference.
[1] L. ZANG, N. NISHINO, T. MIZUUCHI et al., Plasma Fusion Res. 8 (2013) 1402066- 153 -

P1-98Estimation of total carbon radiation at each ionization stage of C2+ to C5+ ions in Large
Helical Device
H. M. Zhang1, S. Morita1, 2, T. Ohishi1, 2, I. Murakami1, 2, X. L. Huang1 and M. Goto1, 2
1Graduate University for Advanced Studies, Toki 509-5292, Gifu, Japan
2

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

Abstract
A detached plasma has been successfully maintained by externally supplying m/n=1/1 magnetic island in
Large Helical Device (LHD) without introduction of additional impurity gas puffing. The study on the carbon
radiation during plasma detachment with edge magnetic island formation is very important to understand thephysical mechanism of effective edge plasma cooling because the carbon radiation is believed to be a mainpart of the total radiation loss in high-density LHD discharges. The total radiation power of carbon ions at
each ionization stage of C2+ to C5+ ions can be analyzed from line spectra of CIII to CVI measured withextreme ultraviolet (EUV: 10-500Å) and vacuum ultraviolet (VUV: 300-2400Å) spectrometers.
2

The

2

resonance lines of CIII (977.02Å, 2s2p – 2s ), CIV (1548.2Å, 2p – 2s), CV (40.27Å, 1s2p – 1s ) and CVI(33.73Å, 2p – 1s) are applicable to estimate the total radiation power at each ionization stage because theresonance transition occupies a considerably large percentage in the total radiation power at each ionization
stage. The C2+ and C3+ ions exists outer side of ergodic layer and the C4+ and C5+ ions are located near LCFS.
It is also very interesting to examine which carbon ions effectively work to create the plasma detachment. The
spectral intensity from EUV spectrometers has been absolutely calibrated based on the bremsstrahlung
measurement in previous studies, while the VUV spectrometer system has not been yet calibrated. As the firststep of the present study, therefore, the spectral intensity from VUV spectrometers has to be calibrated bycomparing the spectral intensities between EUV and VUV spectrometers. The ratio of resonance line intensity
to total radiation power is calculated at each ionization stage of carbon ions using ADAS atomic code. The
total radiation power from C2+ to C5+ ions can be then estimated by taking account of an effect of electron
temperature.

- 154 -

P1-99Measurement of tokamak plasma with the external helical fieldusing a high-speed camera in TOKASTAR-2
T. Sakito, H. Arimoto, T. Fujita, A. Okamoto, Y. Shimooka,
H. Itou, R. Sugioka, K. Muraoka, K. Yasuda, R. Yokoyama
Department of Energy Engineering and Science, Nagoya University,
Furo-cho, Chikusa-ku, Nagoya 464-8603, Japan
sakito-tatsuya14@ees.nagoya-u.ac.jp

1turn voltage
R=0.18m, Z=+0.13m
#3667
#3691

1.2

#3667 w/o Helical field#3691 w/ Helical field

0.8
0.4
0.0
2.50

2.60

2.70 2.80 2.90
Time [ms]

3.00

3.10

Fig.1 (Top) PVF coil and OH coil current(Middle) loop voltages at the outer legs of TF coil
(Bottom)Plasma current[1] K. Yamazaki et al. J. Plasma Fusion Res. SERIES 8 (2009)
1044.
1.2

0.5

1.0

0.4

0.8
0.6
0.4
0.2
0.0

w/o Helical field (two or more peak)w/o Helical field (one peak)w/ Helical field (one peak)

0.26

0.28
0.30
VPVF[KV]

0.32

Duration of IP [ms]

10
8
6
4
2
0

Maximum value of IP [kA]

Plasma current (IP) [kA]

Vloop [V]

IPVF [A]IOH [kA]TOKASTAR-2 device is designed and constructed to study tokamak-helical hybrid magnetic
configuration (TOKASTAR configuration) [1]. The main purpose of TOKASTAR-2 experiment is to evaluate the effect of helical field application on tokamak plasmas and of the plasma
current on compact stellarator configurations. The TOKASTAR-2 coil system consists of six
kind of coils. The magnetic field strength in the plasma center is Bt ~ 0.1 T. To induce plasma
current, electron cyclotron resonance (ECR) heating (2.45GHz) is used for the pre-ionizationand the ECR injection power is ~ 1.4 kW.
To improve tokamak equilibrium control, we adjusted the charging voltage of the capacitor(VPVF) after determining the optimal parameters of the capacitor for the Pulsed Vertical Field(PVF) coil circuit. In this study, the effect of helical field application on tokamak plasma isevaluated. The plasma position and shape are evaluated by light intensity distribution using ahigh-speed camera in the tangential direction.
Figure 1 shows the time evolution of the PVF coil current, Ohmic Heating (OH) coil current,
one-turn voltage, plasma current of a tokamak discharge and that with the external helical field.
Figure 2 shows the peak value and the duration of plasma current as a function of VPVF. From
Fig. 1 and Fig. 2, it is found that under the condition of vertical field (VPVF < 0.30 kV), theduration and the plasma current of tokamak discharge were improved by the external helicalfield.
The plasma position was measured by a high-speed camera in the tangential direction, but thefield of view was limited. So we are now preparing300
to expand the field of view using a mirror. In the
1.5
200
conference,
the result of helical field application on
1.0#3667(VPVF=0.29[kV])tokamakplasmain other conditions and
100
IPVF
0.5
measurement of plasma behavior by a high-speed
IOH
0.0
0
camera will be presented.

0.3
0.2
w/o Helical field (two or more peak)w/o Helical field (one peak)w/ Helical field (one peak)

0.1
0.0

0.26

0.28
0.30
VPVF[KV]

0.32

Fig.2 (Left) The peak value of plasma current and (Right) the duration ofplasma current as function of the charging voltage of the capacitor
for the PVF coil circuit.

- 155 -

P1-100
Observation of Electron Density Fluctuations by Using the O-mode
Microwave Imaging Reflectometry (O-MIR) in LHD
Y. Nagayama, H. Tsuchiya, Y. Soichiroa, S. Hashimoto,N. Itob, D. Kuwaharac, S. Sugitod, Z. B. Shie, M. Jiange
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
Faculty of Engineering Science, Kansai Univ., 3-3-35 Yamate-cho, Suita 564-8680, Japanb
Ube National College of Technology, 2-14-1 Tokiwadai, Ube 755-8555, Japanc
Tokyo University of Agriculture and Technology, 3-8-1 Harumi-cho, Fuchu 183-0057, Japand
Institute for Molecular Science, 38 Nishigo-Naka, Myodaiji, Okazaki 444-0867, JapaneSouthwestern Institute of Physics, P.O.Box 432, Chengdu 610041, ChinaaMicrowave Imaging Reflectometry (MIR) has been intensively developed in Large Helical
Device (LHD) [1]. So far the 60 – 65 GHz X-mode MIR (X-MIR) had been developed in orderto observe core plasm region in the case of hollow density profile, but the ion ITB plasma was
not observable due to high field. In ion ITB plasmas in LHD, the electron density profile ispeaked with the central density of 1 – 2 × 1019 m-3. Therefore the O-mode MIR (O-MIR)
system has been newly developed in order to observe the ion ITB plasma. In O-MIR, themicrowave with four frequencies are simultaneously illuminated the LHD plasma. The
illumination wave frequency can be controlled between 26 – 35 GHz. So, the observable
density is 0.84 – 1.5 × 1019 m-3. The imaging detector named Horn-antenna Mixer Array(HMA) is newly developed for the Ka-band microwave. In HMA, each printed circuit board(PCB) is installed between aluminum plates, each of which have half cut horn antenna array.
The signal wave entering into the horn antenna is transduced from the waveguide to themicro-strip line by the fine-line transducer [2]. The signal wave is amplified by 12 dB before
mixing in order to compensate the conversion loss (12dB) of the double balanced mixer (DBM).
The half frequency of local wave is divided by Wilkinson power divider on PCB and is doubled
by active doublers just before entering into DBM. The optical imaging system of O-MIR ismuch simpler than that of X-MIR since the local wave is delivered by cables. The sensitivity of
O-MIR is significantly improved as the observed noise of O-MIR is 4 mV, while the noise of
X-MIR is 800 mV. Intensity of detected power is calibrated by comparing reflections by aplane aluminum plate and by the form absorber (Eccosorb CV3F). By using O-MIR, we haveobserved electron density fluctuations in the ion ITB plasma and in the H-mode plasma. Thek-ω spectra is obtained in the radial direction, the poloidal direction and in the toroidal
directions by the use of two point cross power spectrum. Preliminary results are as follows:
The density fluctuation near the plasma edge, which is incoherent in the L-mode phase,
becomes coherent after L-H transition.
[1] Y. Nagayama, et al: Rev. Sci. Instrum. 83 (2012) 10E305.
[2] D. Kuwahara, et al: Rev. Sci. Instrum. 85 (2014) 11D805.

- 156 -

P1-101
Examination of efficient electron Bernstein wave heating and measurement
scenario with ray-tracing and full-wave calculations in LHD
H. Igami, A. Fukuyamaa, H. Ideib, K. Nagasakic, Y. Gotod, S. Kubo, T. Shimozuma, Y.
Yoshimura, H. Takahashi, T. I. Tsujimura and R. Makino
National Institute for Fusion Science, 322-6 Oroshi-cho Toki 509-5292, Japan
Department of Nuclear Engineering, Kyoto University, Nishikyo-ku, Kyoto 615-8530, Japanb
Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasuga-koen, Kasuga 816-8580,
JapancInstitute of Advanced Energy, Kyoto University, Uji, 611-0011, Japand
Department of Energy and Technology, Nagoya University, Furo-cho Chikusa-ku Nagoya 464-8603,
JapanaElectron cyclotron resonance heating and current drive (ECRH and ECCD) in over dense
plasmas where the electron density is more than the cutoff density are available with use of
the electron Bernstein waves (EBWs) excited via the linear mode conversion process inside
the plasma by launching the electromagnetic waves from the vacuum region [1]. When theelectromagnetic waves are injected from the lower density side, they encounter the evanescentregion. A part of the waves’ energy can tunnel through the evanescent region and couple withthe extraordinary (X-) mode in the higher density side that propagates toward the upperhybrid resonance (UHR) and couples with the EBWs there. It is important to tune the wave
incident angle and polarization to obtain high coupling efficiency [2].
To describe the wave propagation around the evanescent region precisely the full wave
analysis that solves the Maxwell’s equation directly is required with taking into account the
non-uniform plasma profiles and distribution of the wave electric field. We are developing a
numerical code including TASK/WF2D code that solves the Maxwell’s equation by finite
elements method (FEM) in two-dimensional space. This numerical code can describe thewave propagation across the evanescent region with taking into account the non-uniform
plasma profiles and arbitrary distribution of the electric field of the incident wave. The
ray-tracing calculation cannot describe the wave propagation precisely including thereflection and the tunneling around the evanescent region. An expedient manner [3] to restart
the ray-tracing calculation at the edge of the higher density side of the evanescent region aftertunneling has been used to analyze the wave propagation in the over-dense region.
Comparison between the full wave calculation around the evanescent region and the
ray-tracing calculation with use of the expedient manner to check the validity of this manner
will be reported.
Measurement of the electromagnetic waves originated from thermally emitted EBWs that areradiated via the reversal process of the heating can provide the optimum injection condition ofelectromagnetic for ECRH/ECCD by EBWs. Examination of the measurement with use ofabove method will be also reported. With comparing the results of the measurement to theresults of the examination, validation of the method can be expected.
[1] H. P. Laqua, Plasma Phys. Cntrol. Fusion 49 (2007) R1[2] H. Igami, H. Tanaka and T. Maekawa, Plasma Phys. Cntrol. Fusion 48 (2006) 573[3] H. Igami, H. Idei, S. Kubo, et. al, Plasma Science and Technology, 13 (2011) 405- 157 -

P1-102
Improvement in flexibility of ECCD by use of upgraded ECH antenna
system on LHD
Y. Yoshimura, Y. Narushima, S. Kubo, T. Shimozuma, H. Igami, H. Takahashi,
T. I. Tsujimura, R. Makino, S. Ito, K. Okada, S. kobayashi, Y. Mizuno, Y. Gotoa,
A. Ejirib, R. Sakamoto, T. Mutoh, H. Yamada, A. Komori, Y. Takeiri
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
Graduate School of Engineering, Nagoya University, Nagoya 464-8601, Japanb
Graduate School of Frontier Science, University of Tokyo, Kashiwa 277-8561, Japanaplasma current (kA)

By 2009, three high-power 77 GHz gyrotrons, over 1 MW each, have been installed andapplied to large helical device (LHD) experiment [1, 2]. In addition, two 154 GHz gyrotronsof 1 MW each were installed in 2012 and 2014. Corresponding to the increase in the numberof the high-power gyrotrons, in 2014, power injection mirror antenna system for electron
cyclotron heating (ECH) and electron cyclotron current drive (ECCD) was modified andupgraded. Formerly, an outside port 2-O is furnished with two antenna sets, 2-OLR for power
injection from the first 77 GHz gyrotron, and 2-OLL for the first 154 GHz gyrotron. In
addition to them, two new antenna systems 2-OUR and 2-OUL were installed in the 2-O port.
The second 77 GHz gyrotron changed its power injection port from an upside port 9.5-U to
the 2-O port, using the 2-OUR antenna. The second 154 GHz gyrotron uses the 2-OUL
antenna. An upside port 5.5-U has been used for power injection from the third 77 GHz
gyrotron.
Each antenna in the 2-O port has wide range of EC-wave beam direction control so thatthese are suitable for ECCD which requires toroidally oblique EC-wave beam injection. In the
LHD 18th experimental campaign in 2014-2015, an ECCD experiment with second harmonicresonance condition: on-axis magnetic field of 1.375 T for 77 GHz waves with the magnetic
axis position of 3.75 m, was performed in which some combination patterns of two 77 GHz
ECCDs were applied. Figure 1 shows waveforms of plasma currents in the discharges of dual
co- and counter-ECCDs. 77 GHz EC-wave
powers are applied from 2-OUR and 2-OLR for50
ECCDs, and from 5.5-U for ECH. The injected#128317
ECH + co-ECCDx2
powers from 2-OUR, 2-OLR and 5.5-U antennas
are 365, 366 and 271 kW, respectively. In the
discharge #128316, counter-ECCDs and ECH0
are applied from 1.0 s to 7.6 s, and in #128317
NBI
co-ECCDs and ECH from 1.5 s to 7.6 s.
Tangential NBIs are applied from 7 s so that
#128316
plasma currents driven by NBIs are superposed.
ECH + ctr-ECCDx2-50
The EC-driven currents by the two ECCD
0
2
4
6
8
10
systems are evaluated referring to a former
time (s)ECCD experiment [3], showing that the ECCD
abilities of the two antenna systems are nearly Fig. 1 Waveforms of plasma currents in theidentical.
discharges of dual co- and counter-ECCDs.
References[1] H. Takahashi et al., Fusion Sci. Technol. 57 (2010) 19.
[2] T. Shimozuma et al., Fusion Sci. Technol. 58 (2010) 530.
[3] Y. Yoshimura et al., Plasma Fusion Res. 7 (2012) 2402020.

- 158 -

P1-103
Optimization of ECH injection polarization in real-time polarization scan
experiments
T. Ii Tsujimuraa , Y. Mizunoa , S. Kuboa , R. Makinoa , T. Shimozumaa , Y. Yoshimuraa , H. Igamia ,H. Takahashia,b , T. Mutoha,b , the LHD Experiment Groupaa National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanb SOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292, Japantsujimura.tohru@nifs.ac.jp

A high central electron temperature with an electron internal transport barrier by use of a highly
eﬃcient electron cyclotron resonance heating (ECH) system are required for steady plasma discharge. The operation scenario of the ECH injection system optimally preset before a discharge
may result in a high performance plasma. In addition to that, feedback control of the ECH injection
system set to an optimum target in real time can create a desired power deposition and reduce thestray radiation level in the vessel.
Experiments were recently performed in the Large Helical Device (LHD) in order to optimize the
ECH injection polarization during a single discharge by using the developed fast polarization scansystem with the fast rotatable polarizers installed in a transmission line. These fast polarization
scan experiments showed clear plasma response to change of the injection polarization in real timeand revealed the maximum absorbed power at a certain polarization setting.
The important point is that the polarization was estimated by the upgraded ray-tracing calculation code LHDGauss, which includes the mode purity analyses. The calculation indicates that the
injected millimeter wave with the polarization setting couples to the plasma wave only with thefundamental ordinary mode. The calculated absorbed power is in good agreement with the experimental value. This suggests that prediction of optimum polarization with the help of the LHDGauss
can be used for a control model in a real-time feedback control system if information on an electron
density profile is available as a sensor for the feedback system.

- 159 -

P1-104
Ion heating by ECRH through the equipartition process in the LHD
H. Takahashi, K. Nagaoka, R. Seki, M. Osakabe, M. Yokoyama, Y. Yoshimura, R. Sakamoto,
S. Murakamia, A. Ejirib and the LHD experiment group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japana
Department of Nuclear Engineering, Kyoto Univ., Kyoto 606-8540, Japanb
Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa 277-8561 Japan

Burning fusion plasmas are mainly heated by alpha particle as electron heating. Thus we mustclarify how much ion temperature Ti is realized through the energy equipartition process underthe electron heating dominant situation in order to construct an operation scenario and to predict
plasma parameters in prototype fusion reactors.
The ion heating effect by the energy equipartition has been investigated in the Large Helical
Device using high power electron cyclotron resonance heating (ECRH). Target plasma wassustained by a perpendicular neutral beam injection NBI (3.8 MW) and on-axis ECRH wassuperposed in stepwise up to 4.8 MW. Figure 1 shows the radial profiles of (a) the electron
temperature Te and the electron density ne, and (b) Ti for the moderately high ne plasma. The
solid and the open symbols represent the data with 4.8 MW ECRH and those without ECRH,
respectively. The profiles of ne were approximately same on these two cases. Not only Te but
also Ti clearly increased in whole plasma region during ECRH beam injection. The parabolic
shape of Ti profile was not changed by ECRH but the central Ti value increased from 1.6 keV to
2.0 keV.
Increase of Ti by ECRH was observed in wide ne range (1.3x1019 m-3 < ne). The increment of Ti
increased up to ne ~2.5x1019 m-3 and was saturated in higher ne plasmas. The central ionpressure was found to be expressed as a function of the 0.5 power of the equipartition heat flux.

763(a) #126572
Te_NBI + ECRH
Te_Only NBI

5

Ti_NBI + ECRH(b) #126572

ne_NBI + ECRH
ne_Only NBI

Ti_Only NBI

2
T [keV]

4i

Te [keV], ne [x1019 m-3]

8

3

1

2
1
0

-0.6 -0.4 -0.2

0

0.2

0.4

0

0.6

reff [m]-0.6 -0.4 -0.2

0

0.2

0.4

0.6

reff [m]Fig. 1 The radial profiles of (a) Te and ne, and (b) Ti for the moderate high ne plasma. Not only Te but
also Ti clearly increased in whole plasma region during ECRH beam injection.

- 160 -

P1-105
Experimental Evaluation of Fast-ion Velocity Distribution Produced by
Neutral Beam with Si-FNA in LHD
S. Kamio, M. Osakabe, R. Seki, M. Yokoyama, S. Murakamia, T. Mutoh, and the LHD
experiment group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japana
Kyoto University, Yoshida-honmachi, Sakyo-ku, Kyoto 606-8501, Japan

In order to evaluate confinement property of fast ions, a fast ion measurement system using a
Silicon-diode-based Fast Neutral Analyzer (Si-FNA) was developed in the Large Helical
Device (LHD) [1]. In evaluating the energy of the fast neutrals, the Pulse Height Analysis(PHA) technique is adopted. This system can measure the fast neutrals one by one and its
detection efficiency is almost 100%. A Si-FNA array was newly installed at a tangential port of
the LHD in order to evaluate the confinement of fast ions which was produced by tangentiallyinjected Neutral Beam (NB). Since the sight lines of the newly installed Si-FNAs haveintersections with a radially injected NB and this NB can be utilized as an active source of
charge exchange neutrals, an absolute measurement of fast ion density becomes possible.
Combining the obtained local fast ion information by the Si-FNA with that by the Fast Ion
Charge eXchange Spectroscopy (FICXS), verification and validation of fast-ion simulation
code such as TASK3D[2], FIFPC, GNET[3], and MORH[4] become possible. The comparison
between these experimental data and simulation will reveal the confinement properties of fastions in helical devices and provide deep understandings of them.
In the NB short pulse injection (NB-blip) experiment, the slowing-down process ofmono-energetic fast ions were clearly observed. In addition to the measurement of theslowing-down process, the local energetic particles were measured in three kind of toroidal
magnetic field (2.75 T, 1.375T, 0.75 T) and various electron density. By using theseexperimental results, the orbit effect and the effect of the slowing-down to the velocity
distribution are investigated by comparison with the simulation codes.

[1] M. Osakabe et al., Rev. Sci. Instrum. 72 (2001) 788-791.
[2] S. Sato et al., Plasma Fusion Res. 3, S1063 (2008).
[3] S. Murakami et al., Nucl. Fusion 46, S425 (2006).
[4] R. Seki et al., Plasma Fusion Res. 5, 027 (2010).

- 161 -

P1-106
Developments of millimeter-wave interferometer for electron density
measurements with beam extraction in the negative ion source
T. Tokuzawa, M. Kisaki, K. Nagaoka, K. Tsumori, K. Ikeda, H. Nakano,S. Genga, M. Osakabe, Y. Takeiri, and O. kanekoaNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanThe Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki 509-5292, Japan

In the NIFS 1/3-scaled negative ion source for LHD [1], the Cavity-Rign-Down (CRD)
method and Langmuir probes have been utilized for studying the dynamics of the negative ion
and electron during beam extraction, respectively. The Langmuir probe can give theinformation on the electron density qualitatively, but it has difficulty in obtaining the electron
density precisely due to the strong magnetic field in the plasma chamber and the contribution ofnegative ions to the electron saturation current.
The millimeter-wave interferometer is one of the presumable tools for measuring theelectron density under such condition. At first, we had installed the millimeter-wave
interferometer to the negative ion source and successfully obtained the electron density in thepure hydrogen and Cs-seeded plasmas [2]. In order to measure the electron density with
applying the high voltage on the plasma chamber, the millimeter-wave interferometer has been
upgraded. Figure 1 shows the time evolution of the electron density obtained with improved
millimeter-wave interferometer in the pure hydrogen plasma. The beam was extracted from
10.8 s to 11.8 s. The electron density changes with the arc power and rapidly rises when theextraction voltage is applied. Amount of
the electron density change before andafter the beam extraction is 2.6×1017 m-3.
It has been observed that the negative ion
density shows the different response to theextraction voltage and it rapidly drops
with beam extraction. In the conference,
the improvements on the diagnostic
system will be described in detail and the
response of the electron to the extraction
voltage in other conditions (e.g. Cs-seeded
plasma and different bias voltages) will be Fig. 1 Time evolution of electron density measured
with millimeter-wave interferometer.
presented.

[1] K. Tsumori, et al., Rev. Sci. Instrum. 83 (2012) 02B116.
[2] K. Nagaoka, et al., AIP Conf. Proc. 1390 (2011) 374.

- 162 -

P1-107
Simulation study of NBI beam ion distributions in the presence ofmagnetic islands in helical plasmas
H. Yamaguchi and S. Murakami
Department of Nuclear Engineering, Kyoto University, Nishikyo, Kyoto 615-8540, Japan

E-mail: yamaguchi@p-grp.nucleng.kyoto-u.ac.jp

Existence of nested closed magnetic flux surfaces is an essential requirement for a good
plasma confinement in toroidal configurations. Magnetic perturbations (MHD modes, error
fields and etc.) generate magnetic islands and, finally, the nested flux surfaces are broken
by the magnetic island overlapping. Thus, these magnetic islands deteriorate the plasma
confinement. On the other hand, it is suggested that optimized and controlled magnetic
island can have benefit for plasma performance in tokamak plasmas. Therefore a detailstudy should be be done to evaluate the eﬀect of magnetic island on the plasma confinement.
The eﬀect of magnetic islands on energetic ion confinement has been studied using several
guiding-center codes in tokamak plasmas. They found that the magnetic island enhances theenergetic ion loss due to the enhanced radial ripple diﬀusion. However, the eﬀect of magnetic
islands on energetic ion confinement has been studied in helical plasmas.
In the Large Helical Device (LHD) experiments, magnetic island eﬀect and its dynamics in
LHD plasmas have been studied applying the resonant magnetic perturbation (RMP) coils [1],and the behaviors such as temperature flattening in the magnetic islands have been observed
as in tokamaks. In these experiments three tangential and two perpendicular neutral beam
injection (NBI) heating systems are applied. These neutral beams produce energetic ions,which draw variety of passing and trapped orbits more complicated than that in tokamaks.
Especially, the confinement of ripple-trapped particles are important for energetic particle
confinement in LHD.
In this paper, we study the eﬀect of magnetic islands on the NBI beam ion confinementin LHD plasma. We model the resonant magnetic perturbation as that with a single helicity.
Taking into account those magnetic perturbations, we solve the drift kinetic equation for NBI
beam ions in 5-D phase space using the GNET-TD code [2], which is based on GNET [3].
The complex guiding-center motion and the pitch-angle and energy scatterings during energy slowing down process are considered in the simulation. We discuss the eﬀect of RMP
producing magnetic islands on confinement and velocity space distribution of NBI beam ionsin LHD plasmas.
[1] Y. Narushima et al., Nucl. Fusion 55 (2015) 073004[2] H. Yamaguchi et al., Plasma Fusion Res. 8 (2013) 2403099[3] S. Murakami et al., Nucl. Fusion 46 (2006) S425

- 163 -

P1-108
Optimization of helical coil structure
for the improved magnetic configuration of LHD
Shoichi Okamura
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

An optimized magnetic configuration was found for the LHD-type confinement device [1, 2].
The new configuration has a reduced Shafranov shift for high beta equilibria which gives
better neoclassical confinement than present LHD configurations. This result was obtained
based on the physical understandings of the Fourier components of the boundary shape of
LHD plasmas. It was found that two fundamental Fourier components representing the helicalexcursion of the magnetic axis and the rotating triangularity are important to determine thebasic confinement characteristics of magnetic configuration, that are the neoclassical transport
and the MHD stability. A new optimized configuration was produced by modifying these two
essential components which can not be produced by the present LHD magnetic coils. The
magnetic coil design for this new configuration was made using the coil optimization
technique and a combination of simple advanced helical coil and the axisymmetric poloidal
field was found [3]. The obtained helical coil structure has an additional parameter of theellipticity of the poloidal cross section of a coil winding surface. In this paper, the coil design
study is extended to survey the physical properties of the magnetic configurations created by
the advanced helical coil structure with an additional geometric parameters to the existing
helical devices.
[1] S. Okamura, Plasma Phys. Control. Fusion 55 (2013) 032002.
[2] S. Okamura, Plasma Fusion Res. 8 (2013) 2402029.
[3] S. Okamura, 41th EPS Conference on Plasma Physics, Berlin (2014) P4.011- 164 -

P1-109
Spatial structures of energetic-particle driven geodesic acoustic modein the LHD

T. Idoa, A. Shimizua, M. Osakabea,b, T. Wataria, K. Ogawaa,b, K. Toia, M. Nishiurac, K. Itoha,
M. Lesurd, and the LHD Experiment GroupaNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi, Toki, Gifu 509-5292,
Japanc
Graduate School of Frontier Sciences, The University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chiba
277-8561, Japand
Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasuga-koen, Kasuga, Fukuoka,
816-8580, Japanb

Measurement of spatial structures of fluctuations is important for identification of thefluctuations and for investigation of their dynamics and influence on plasmas. In thisexperiment, spatial structures of energetic-particle driven geodesic acoustic mode (EGAM)
have been measured using a heavy ion beam probe (HIBP) and Mirnov probes in the LHD
plasmas.
EGAM in the LHD plasmas with normal magnetic shear profiles, where the rotational transformmonotonically increases from the center to the edge of the plasmas, is excited in the central region of theplasmas, and is usually accompanied by the frequency up-chirping. Measured spatial structures of theelectric potential fluctuation and the density fluctuation associated with the EGAM agree the theoreticalprediction: m = 0 for the electric potential fluctuation and m = 1 for the density fluctuation, where m is
the poloidal mode number. The structures are almost same during the frequency chirping[1].

When the frequency of the chirping EGAM reach twice the GAM frequency determined withparameters of bulk plasmas, a mode with the GAM frequency is excited abruptly. The mode isexcited in the same region as the chirping EGAM. Because the spatial structures are also sameas the EGAM, the abruptly excited mode is identified as a GAM[2] and it is considered to be
excited through driven by a cooperative collaboration of fluid nonlinearity and kinetic
nonlinearity[3]. The amplitude of the electric potential of the abruptly excited GAM canbecome roughly twice larger than that of the EGAM, and the GAM may affect transport ofbackground plasmas.
Measurement by both HIBP and Mirnov coils reveals that the chirping EGAM are affected byelectron cyclotron heating (ECH) and by resonant magnetic perturbation (RMP). By applying
ECH, the density fluctuation of the chirping mode become less up-down antisymmetry, while
the structure of the electric potential does not changes. Thus, m ≠ 1 for the density fluctuation,
though m = 0 for the electric potential fluctuation. By applying RMP with m/n=2/1, where n is thetoroidal mode number, n of the chirping mode becomes one. Both results may reflect structuresof the magnetic field.

[1] T. Ido, et al., Nucl. Fusion, 55 (2015) 083024[2] T. Ido, et al., submitted to Phys. Rev. Lett. (2015)[2] M. Lesur, et al., submitted to Phys. Rev. Lett. (2015)- 165 -

P1-110Slowing down mechanism of instability leading to minor collapse in LHD
Y. Takemura, K.Y. Watanabe, S. Sakakibara, Y. Suzuki, S. Ohdachi, Y. Narushima, K. Ida, I.
Yamada and LHD Experiment Group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanThe MHD instability causing rapid drop in core electron pressure (Minor collapse) is observedin the LHD discharge with the magnetic configuration where the Mercier stability condition is
not satisfied [1]. The activity of the instability is changed through three phases divided bybehavior of its rotation. In constant rotation phase, rotational frequency and amplitude of theinstability’s fluctuation is constant and clear flattening of the pressure profile is not observed.
In slowing rotation phase, the amplitude increases with the decrease of the rotational frequency
and flattening of the pressure profile occurs. In stopping rotation phase, the rotationalfrequency is zero and the amplitude rapidly grows, the minor collapse takes place. The slowing
down of the instability’s rotation is believed to lead to occurrence of the minor collapse.
However, the reason for the slowing down is not clear. If the LHD experiment with themagnetic configuration where the minor collapse occurs is performed, the rotational frequency
therotationalof the instability’s fluctuation in slowing rotation phase coincides withfrequency of the plasma flow at the resonant surface within the error bar, as shown in Fig.1.
This is consistent with the result of the research [2] on the rotational frequency of the typicalresistive interchange instability which does not cause the minor collapse. Here the electron
diamagnetic drift flow is zero because
flattening of the pressure profile appears in thisexperiment. According to time evolution of theplasma flow profile in Fig.2, the plasma flow islarge in edge region but small in core region.
The region of the small plasma flow is slowly
extended to edge region with the decrease of
the rotational frequency. On the other hand, theresonant surface is rapidly moved to the coreregion. Especially at R~4.2 m at an early timein slowing rotation phase, the situation of thefast plasma rotation is kept even with thedecrease of the rotational frequency of theinstability’s fluctuation. These results suggest Fig.1 Typical discharge where the minorthat the resonant surface moving to the core collapse occurs due to the instability
region is the reason for the slowing down. The
move to the core region is believed to be causedby the increase of the Co- plasma current which
increase the rotational transform.

[1] Y. Takemura, Nucl. Fusion 52 (2012)
102001.
[2] Y. Takemura, Plasma and Fusion Research
8 (2013) 1402123.
Fig.2 Time evolution of the plasma flow
profile and the resonant surface- 166 -

P1-111
Stabilization of Helically Trapped Energetic Ions driven Resistive
Interchange Mode by High Power Injection of Electron Cyclotron
Resonance Heating
X.D. Du1 , S. Ohdachi1,2 , K. Nagaoka1,2 , M. Osakabe1,2 , Y. Suzuki1,2 , K. Toi2 and LHD
Experimental Group1 Department of Fusion Science, Sokendai(The Graduate University for Advanced Studies), 322-6

Oroshi-cho, Toki 509-5292, Japan
2 National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japandu.xiaodi@lhd.nifs.ac.jp

b!θ [T s][1] X.D. Du, K. Toi et al., Phys. Rev. Lett. 114 , 155003 (2015)- 167 -

PECH [MW ]The resistive interchange mode destabilized through resonant interaction with a characteristic
motion of helically trapped energetic ions (EPs) are observed in Large Helical Device (LHD),
called the EIC, exhibiting bursting character and rapid frequency chirping down. The EIC
strongly impacts the confinement of helically trapped EPs and induces noticeable losses. The
loss fraction from the plasma confinement region is crudely estimated to be ∼ 40% over the
impact region of the EIC [1]. Therefore, the study on the control of the EIC in LHD is crucial
to reduce the losses of helically trapped EPs.
In recent LHD experiment, the
EIC
EIC
clear suppression of the EIC is foundduring the injection of the high power(a)suppression
20
8electron cyclotron resonance heating(ECH).One typical example is shown in
Fig. 1. In the so-called high Ti dis0
4
charge of central electron density ne ∼19
1 × 10 and temperature Te ∼ 3keV
, the EIC bursts exhibits the fishbone−200shape oscillation in magnetic probes.
4.4
4.6
4.8
5.2
5.4
However, during the time window fromtime[s]
4.65s to 4.85s when the high ECH
power of 5.2M W is injected without Figure 1: Time evolutions of the magnetic fluctuations
changing the power of neutral beam in- measured by a magnetic probe signal and injected ECH
jection perpendicular to magnetic field power.
line (PERP-NBI), the strong bursting
oscillations in magnetic probes disappear. Moreover, the EIC suppression phase even persists ∼ 100ms after switching off the ECH, as seen from Fig. 1. This might be due to thedecrease of the deposition power of the PERP-NBI by the strong electron density pump-out
induced by the ECH, i.e., ∼ 20% decrease of the electron density in this case, since thedestabilization of the EIC strongly correlates with the contents of helically trapped EPs inplasma. The other possibilities of the physics mechanism of the suppression of the EIC by
the ECH will also be discussed.

P1-112Development of fast-neutron directional detector
for fusion neutron profile monitor at LHD
E. Takadaa, N. Nakadaa , M. Isobeb, c, K. Ogawab, c, T. Nishitanib, H. TomitadaNational Institute of Technology, Toyama College, 13 Hongo-machi, Toyama 939-8630, Japanb
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanc
SOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292,
Japand
Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8601, Japan

takada@nc-toyama.ac.jp

In order to measure the incident directions of 14 MeV neutrons, a direction sensitive
detector composed of scintillating fibers (Sci. Fi.) has been used[1], [2]. In that system, thesignals from the Sci. Fi. were measured with a photomultiplier tube (PMT) where thepenetration of recoil proton from one fiber to the other was prevented with the shielding region.
The system was applied to TFTR and JT-60U to evaluate the time dependence of 14 MeV
neutron emission for the triton burn-up study in deuterium discharges. In the present study, weare trying to optimize the design of the detector to use at LHD. In TFTR and JT-60U, the Sci. Fi.
with the length of 10 cm and the diameter of 1 mm has been used. With PHITS code, we
calculated the proton flux distribution for 14 MeV neutrons incident parallel to the Sci. Fi.
Figure 1 depicts the recoil proton distribution when neutrons are incident parallel to the Sci. Fi.
It shows that recoil protons are generated mainly in the region near the inlet, which is caused by
the self-shielding by the Sci. Fi. This result suggests the necessity of design optimization.
We calculated the angular dependence of counts over a threshold energy for the case of
1 mm-diameter Sci. Fi. The lengths were set to 10 cm and 5 cm. Also, experiments werecarried out at Fusion Neutron Source (FNS) facility of Japan Atomic Energy Agency. It wasobserved that the measured angular distribution is broader than calculated results. As Sci. Fi. issensitive to gamma rays, the difference should be due to the background gammas. To reduce
the gamma sensitivity, it is effective to adopt smaller core Sci. Fi. In Fig. 2, calculated results
are shown for different core size. We can conclude that, by using smaller core Sci. Fi., higher
angular resolution can also be achieved. However, smaller diameter causes lower measuring
efficiency also for neutrons. For optimizing the design of Sci. Fi. detector to measure the triton
burn-up neutron profile in the LHD plasma, additional simulation and experiments will be
carried out.
[1] Wurden, G. A. et al., Rev. Sci. Instrum. 66 (1) (1995) 901.
[2] Harano, H. : JAERI-Research 97-060 (1997) (in Japanese).

Fig. 1. Distribution of proton flux in themeasurement system.

Fig. 2. Comparison of calculated angular
distribution.

- 168 -

P1-113
Observation of termination process of long pulse plasma discharges using
stereoscopic fast framing cameras in the Large Helical Device
M. Shoji, H. Kasahara, H. Tanaka, T. Murase, M. Tokitani, S. Morita, M. Goto, T. Oishi,
T. Mutoh, E. de la Cala, C. Hidalgoa and the LHD experiment groupa

National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, 40 Avenida Complutense,
Madrid 28040, Spain

The Large Helical Device (LHD) has superconducting helical and poloidal coils by whichplasma confinement magnetic configurations can be formed without toroidal plasma current. It
is advantageous for sustaining steady state plasma discharges without instabilities induced by
the plasma current. In LHD, long pulse plasma discharge experiments have been performed formore than the last ten years using ICH and ECH. As a result, a long pulse discharge with aplasma density of ~1.2×1019 m-3, a central electron temperature of ~2 keV and a plasma heating
power of ~1.2 MW for about 48 minutes was successfully achieved in the experimental
campaign in FY2013. The plasma discharge was terminated by abrupt increase in carbon ion
emission triggered by the release of large amounts of carbon dusts from a lower divertor region
in the inboard side of the torus. Observation with a fast framing camera proved that the
carbon-rich mixed material deposition layers accumulated in front of divertor plates in thedivertor region were exfoliated just before the termination of the plasma discharge. It was also
found that the exfoliated mixed material layers were broken into small fragments (dusts) which
penetrated into the main plasma to terminate the long pulse plasma discharge [1].
Before the last experimental campaign in FY2014, the lower divertor configuration was
modified so as to suppress the deposition layers in front of the divertor plates, which
successfully controlled the termination of long pulse discharges induced by the large amountsof carbon dusts in the last campaign. Nevertheless, the long pulse discharges was often
terminated with abrupt increase in iron ion emission. One of the terminations of the plasma
discharges was directly observed with a stereoscopic fast framing camera installed in an outer
port, which clearly shows the following plasma termination process:1. in the beginning, a luminescent spot appeared on a gap between the surface of stainlesssteel tiles for protecting a side wall of a helical coil can in the inboard side of the torus,
2. the spot widely expanded to the entire region of the helical coil can in the inboard sidewith intensive sparks,
3. after the sparks vanished, large amounts of incandescent dusts were released from near
the top of the helical coil can in the inboard side,
4. the released dusts penetrated into the main plasma and induced the radiation collapse toterminate the long pulse plasma discharge.
The stereoscopic analysis of the three-dimensional trajectories of the observed dusts revealed
that the dusts locate in the inboard side of the torus. It also proved that the dusts penetrated into
the main plasma confinement region to the positions where the normalized minor radius ρ isabout 0.8. The analysis clearly indicates that the investigation of the physical mechanism
triggering the luminescent spot at the gap between the tiles is essential for controlling theintensive sparks and the release of the dusts. Appropriate measures for suppressing the
expansion of the spot can lead to extension of the duration time of the long pulse discharges.
[1] M. Shoji, H. Kasahara, M. Tokitani et al., Nucl. Fusion 55 (2015) 053014.

- 169 -

P1-114
Identi cation of helical structure of visible light in Low-A RFP
with concave lens array
A. Sanpei, S. Masamune, Y. Aoki, T. Nagano, M. Higuchi, K. Jikuhara, S. Nankanobo,
R. Tsuboi, R. Isomoto, H. Himura, S. Ohdachia , N. Mizuguchia , T. Akiyamaa
Kyoto Institute of Technology, Kyoto 606-8585, Japana National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japansanpei@kit.ac.jp

Three-dimensional (3-D) eﬀects on MHD phenomena in axisymmetric systems such as reversed ﬁeld pinches (RFPs) have attracted much attention. In the RFP, for example, recent
progress has shown the importance of helically deformed RFP conﬁguration where the single
helical magnetic axis state is self-organized [1].
In a low-aspect-ratio (low-A) RFP machine RELAX (R = 0.51 m/a = 0.25 m (A = 2)) [2], aquasi-periodic transition to quasi-single helicity (QSH) state has been observed. During the
QSH state, the ﬂuctuation power concentrates in the dominant m = 1/n = 4 mode, and a(toroidally rotating) 3-D helical structure has been observed with radial array of magnetic
probes [3]. We succeeded in identifying the characteristic helical SXR structures whichsuggest hot or dense helical core in QSH state, using the SXR imaging diagnostic system [4].
A high-speed camera diagnostic has revealed simple helix structure in the visible light image[5,6]. If there is no limitation on observation positions, a 3-D tomographic reconstruction is a
powerful method for obtaining 3-D information; however, the observation positions and theiravailable numbers are considerably restricted in plasma experiment devices. It is requiredthat a method for identiﬁcation of 3-D structure from a single exposure image obtained fromone viewing port.
In the present study, we are developing a 3-D recording system where multiple concave
lenses and a high-speed camera for the study of time evolution of 3-D structure. As an
initial experiment, we have taken vertical visible light images with time resolution of 10 µsec,
to identify time evolution of ﬁlament-like single helix structure in RELAX plasmas. We
will discuss these results to make clear experimental and theoretical requirements for 3-D
structural studies.
This work is partly supported by JSPS KAKENHI Grant Number 15K05364.

[1] R. Lorenzini et al., Nature Physics 5 (2009) 570[2] S. Masamune et al., 24th IAEA Fusion Energy Conf., EX/P4-24 (2012)[3] K. Oki et al., Plasma Fusion Res. 7 (2012) 1402028[4] A. Sanpei et al., IEEE Transaction Plasma Science, 39 (2011) 2410
[5] T. Onchi et al., Plasma and Fusion Res. 3 (2008) 005[6] A. Sanpei et al., Plasma and Fusion Res. 5 (2010) S2306- 170 -

P1-115
Radiation field estimation for the diagnostic and control components byMonte Carlo neutronics calculations with LHD 3-dimensional modeling
T. Nishitania, K. Ogawaa,b, M. Isobea,babNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI (The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki- 509-5292,
Japan

LHD has a plan of the deuterium experiment in 2017. Precise estimation of the radiation(neutron and gamma) field should be performed in order to confirm the soundness of thediagnostic and control components for the LHD deuterium experiment. Previously, radiation
filed in the LHD torus hall was calculated by 2-Dimensional neutron transport DORT. The
radiation field around the cryostat must be affected by the structure of the LHD, such as
horizontally located outside port (O-port) with large diameter, which could not be estimatedby 2-Dimensional neutrons calculations. Here we carried out Monte Carlo neutrons
calculation with 3-Dimensional modeling of LHD by using MCNP-6 code with thecross-section library of ENDF B-VI. In MCNP code, all surfaces of cells must be defined byquadratic (or less) equations or torus, however, helical surface could not be treated. In the
MACP calculation geometry, the LHD components within the support structure are divided
by small toroidal angle, and the components are assumed to be toroidally symmetric in a
toroidal pitch angle. The geometry in one toroidal pitch angle is modeled by the CAD
drawing with some simplification. As the first step, the toroidal pitch angle is selected to be 6°.
Figure 1 shows the poliodal cross sections of the LHD calculation model, where the toroidal
angle of φ=0 is defined by the center of the U/L port. A poloidal angle θ of a center of one
helical coil is presented by θ = φ (m/l) + α sin{φ (m/l)}, where m=10, l=2 and α =7.8°. The
LHD plasma is assumed to be a simple torus. Surface equations in each toroidal pitch angle
are generated automatically by EXCEL based on this formula. By this calculation, neutron
flux is estimated to be 7.50 × 109 n/cm2•s at R=8 m and φ = 18° on the mid plane (just in frontof the 0-port outlet) and 6.50 × 109 n/cm2•s at R=8 m and φ = 0° for the maximum neutron
yield shot with 1.9 × 1016 neutrons/s. In the region of R>10m, the neutron flux is almost
symmetric toroidally, and agrees with the result of the 2-Dimensional neutron transport
calculation by the DORT code.

Figure 1 Poliodal cross sections of the LHD calculation model at φ = 0°, 12° and 18°, where
the toroidal angle φ=0 is defined by the center of the U/L port.

- 171 -

P1-116
Integrated heat transport simulationof the helium-hydrogen plasma of LHD by TASK3D
S. Maeta, S. Murakami, H. Yamaguchi, A. Fukuyama, K. Nagaokaa , H. Takahashia ,H. Nakanoa , M. Osakabea , K. Tanakaa , M. Yokoyamaa , and LHD Exp Groupa
Department of Nuclear Engineering, Kyoto University, Kyoto 615-8540, Japana National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Various physical phenomena with wide ranges of time and space scales are connected in thefusion plasma. Thus, in order to predict the performance of the fusion plasma, an integrated
simulation combining various physics models is required. Many integrated simulation code
has been developed for tokamak and helical plasmas. TASK3D is an integrated simulation
code for helical plasma and has been applied to the analyses of the plasma transport of LHD
plasmas[1, 2].
On the other hand, the deuterium experiment project from 2017 is planned in LHD, where thedeuterium NBI heating beams with the power more than 30MW is injected to the deuterium
plasma. Main objects of this project are to make clear the isotope effect on the heat andparticle transport in the helical plasma, and to study energetic particle confinement in ahelical magnetic configuration measuring triton burn-up neutrons.
In the 2014 campaign of LHD we have performed the helium-hydrogen plasma experiment tomake clear the ion species dependence on the heat transport, which would be related to theisotope effect on the plasma confinement. We have found that the central ion temperaturegoes up as the helium fraction increases in the helium-hydrogen plasma.
In this paper, we study the heat transport of the helium-hydrogen experiment plasma to
make clear the ion species dependence on the heat transport. We evaluate the NBI heat
deposition of multi-ion plasma by GNET code[3, 4] and study the heat transport by TASK3D.
It is found that the turbulent transport improvement is necessary to explain the increase of
ion temperature in the high fraction helium plasma. We apply the turbulent transport
improvement models and obtain relatively good agreements.

[1] S. Murakami, et al., Plasma Phys. Control. Fusion 57 (2015) 054009.
[2] A. Sakai, et al., Plasma Fusion Res. 9 (2014) 3403124.
[3] H. Yamaguchi and S. Murakami, Plasma Fusion Res. 9 (2014) 3403127.
[4] S. Murakami, et al., Nuclear Fusion 46 (2006) S425.

- 172 -

P1-117
Full-orbit simulation of fast α-particle in heliac configurations
J. Tachibana, S. Kitajima, H. Takahashi, A. Okamotoa
Dept. of Quantum Science and Energy Engineering, Tohoku Univ., Sendai 980-8579, Japana Dept. of Energy Engineering and Science, Nagoya Univ., Nagoya, 464-8603, Japan

Heliac is one of stellarator configurations, which consists of only simple circular coils. In a
heliac a rotational transform of the magnetic field line is produced by helically placed toroidal
field coils. The magnetic field geometry of heliac configuration has a deep magnetic well in
vacuum configuration, and it is expected that the heliac configuration has good confinement
property in a high beta plasma.
In self-heating mechanism of a thermonuclear fusion reacter, it is an important process thatfast α-particles produced by D-T reaction are well confined with magnetic field, and are slowing down by collision with particles in a bulk plasma. Therefore it is necessary to examineand evaluate the performance of the fast α-particle confinement in a heliac type reactor. Recently, hybrid guiding-center/full-orbit simulations of fast ion in non-axisymmetric magnetic
geometry was reported [1]. It is diﬃcult to calculate accurately the orbit of a particle by theguiding center simulation in the case where the scale length of the gradient of magnetic field
is larger than the Larmor radius. Therefore, the full-orbit simulation becomes significant in
the orbit calculation of the fast α-particle born in D-T reaction.
In this presentation, we investigate the performance of the α-particle confinement in heliac
configuration by computing the full-orbit simulation instead of the guiding center simulation.
We also investigate the dependence of the performance of the α-particle confinement on
toroidal and helical ripples by changing the ratio of coil currents. And we will discuss
candidates for a suitable heliac configuration to confine a fast α-particle. In this simulation,
we use the magnetic configuration of Tohoku University Heliac device (TU-Heliac) [2] as oneof heliac configuration models.
This work is supported by Grant-in-Aid for Scientific Research (KAKENHI), 24246152.

[1] D. Pfeﬀerlé, et al., Plasma Phys. Control. Fusion 36 (2015) 054017.
[2] S. Kitajima, et al., Nucl. Fusion 46 (2006) 200.

- 173 -

P1-118
Recent progress of fast-ion loss detector project in magnetic confinement
fusion devices in Japan, Korea and China
M. Isobea,b, Junghee Kimc,d, Yipo Zhange, Jiafeng Changf,
K. Ogawaa, Jun-Young Kimd, and Yi Liueab

National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanSOKENDAI(The Graduate University for Advanced Studies), 322-6 Oroshi-cho, Toki 509-5292,
JapancNational Fusion Research Institute, Daejeon 305-806, Korea
dKorea University of Science and Technology, Daejeon 305-350, Korea
eSouthwestern Institute of Physics, P.O. Box 432, Chengdu 610041, ChinafInstitute of Plasma Physics, Chinese Academy of Science, Hefei 230031, China
isobe@nifs.ac.jp

Good confinement of fast ions in fusion plasmas is crucial to the success of the fusion
reactor since fusion-born α particles play an essential role in steady state sustainment of
future burning plasma as a primary heating source. Although unfavorable effect on fast-ion
orbit caused by axisymmetry breaking of the system is one of the oldest issues in fusion,
attention is now being focused on it again because the non-axisymmetric three-dimensional(3D) perturbed field produced by resonant magnetic perturbation (RMP) is often superposed
to mitigate ELMs in recent experiments. Fast-ion-driven MHD instabilities such as toroidal
Alfvén eigenmodes (TAE) and energetic-particle modes (EPM) are also of great concern
since those can potentially lead to fast-ion loss. For reasons above mentioned, tightcollaborative activities on fast-ion physics in fusion were initiated in 2012 in Asian three
countries in the support of the A3 foresight program on critical physics issues specific tosteady state sustainment of high-performance plasmas. Primary purpose of our joint work is toobtain comprehensive understanding of fast-ion transport and/or loss caused by 3D field andfast-ion-driven MHD instabilities in toroidal fusion plasmas. Our immediate goal in the earlystage of this program was to set up scintillator-based fast-ion loss detectors (FILDs) onto four
major fusion devices in East Asia, i.e., LHD heliotron [1], KSTAR [2], HL-2A [3], and EAST
tokamaks, and this task was complete in 2014. In LHD, TAE/EPM-induced fast-ion losses
have been intensively studied by using the FILD [4]. The KSTAR FILD is also successfully
working. A study on behavior of fast ions due to RMP field is in the center of attention in
KSTAR [5]. In HL-2A, fast-ion losses induced by tearing modes, long-lived mode, sawtoothcrash and disruption have been observed [6]. A recent big step in our work is that theoperation of the FILD on EAST has begun lately. The localized bright spot appears on thescreen while neutral beam (NB) is injected and disappears after NB is turned off as expected.
In the latter stage of the A3 program, joint experiments will be conducted to reveal common
physics issues related to fast ions in tokamak and heliotron/stellarator. In this paper, current
status on fast-ion loss detector projects in Japan, Korea and China, and representative resultsfrom ongoing activities will be presented.
[1] K. Ogawa et al., J. Plasma Fusion Res. SERIES 8 (2009) 655.
[2] Junghee Kim et al., Rev. Sci. Instrum. 83 (2012) 10D305.
[3] Yipo Zhang et al., Rev. Sci. Instrum. 85 (2014) 053502.
[4] M. Isobe et al., Contrib. Plasma Phys. 50 (2010) 540. ; K. Ogawa et al., Nucl. Fusion 53(2013) 053012.
[5] Jun-Young Kim et al., 41st EPS Conference on Plasma Physics. Berlin, 23-27 June 2014.
P4.067.
[6] Yipo Zhang et al., 25th IAEA FEC 13-18 Oct. 2014, St. Petersburg, EX/P7-24.

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Poster  Session  2  Tuesday,  4th  November,  2015  
  
  
  

  

  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  
  

  

P2-1
Silicon Nitride Film Deposition on Glass Substrates via Plasma
Sputter-Type Negative Ion Source
U. Jenica Rozettea, C. Aren Renza, R. HenryaaPlasma Physics Laboratory, National Institute of Physics, University of the Philippines Diliman,
Quezon City, 1101

Silicon nitride films were successfully deposited on glass substrates using the plasma
sputter-type negative ion source. The silicon metal target used was biased with -300V and was
sputtered with Ar-N2 plasma for 2 hours. Total flow rate of the Ar-N2 gas mixture was fixed to
5.0 sccm while varying the N2 concentration from 10% to 50%. Filling pressure of the system
during deposition is maintained at 2.6 x 10-5 Torr. Visual inspection of the deposited films
shows variation in color from grayish black (low N2) to yellowish gold (high N2) whichsuggests variation on film thickness. SEM images actually reveals that films deposited at lower
N2 concentration tends to have higher film thickness. It is because increasing the N2
concentration decreases the number of available Ar ions for sputtering which then decreases thenumber of sputtered Si that would react with N2 plasma. This observation is consistent with theprevious studies [1, 2]. However for this study, N2 concentration seems to have no significant
effect on the crystal structure and chemical composition of the deposited silicon nitride films.
XRD analysis of all the samples reveals only the same α-Si3N4 peaks present in all the samples
with increasing intensities as N2 concentration decreases.

References[1] S. K. Habib, A. Rizk, I. A. Mousa, Vacuum 49 (1998) 153–160.
[2] B. S. Yau, J. L. Huang, Surface and Coatings Technology 176 (2004) 290–295.

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P2-2
Anti-fog on PMMA by Nitrogen-Argon Atmospheric Plasma Jet Treatment
S.J. Park, K. Penado, M. Jallorina, C. Mahinay, I. Culabaa

Department of Physics, School of Science and Engineering, Ateneo de Manila University, Katipunan
Ave, Quezon City, 1108 MetroManila, Philippines

In this study,
polymethyl-methacrylate (PMMA) samples were treatedusing Nitrogen-Argon Atmospheric Plasma Jet (APJ). N2/Ar flow rate ratios of 30:5, 30:10,
30:15, 30:20 to 30:25 SCFH were used. Measurements of the contact angle of different
samples were used to determine the best treatment parameters for anti-fog application.
(a)

(b)

Figure 1. (a) A schematic diagram of the APJ device and (b) an image of nitrogen-argon APJ.

As shown in Figure 1(a), the APJ device was constructed using a borosilicate glass cross tube,
pair of tungsten electrodes was secured on each end of the tube and connected to a 15,000V
30mA Neon sign transformer. Rotameters attached to gas tanks were used to regulate the gas
flow.
(a)

(b)

(c)

Figure 2. Contact angles of (a) untreated and (b) treated PMMA with nitrogen-argon flow rate ratio of
30:10 SCFH and a (c) photo of the visor after the condensation test.

AFM images revealed an increase in the roughness of the PMMA sample after treatment due
to formation of nanograins [1]. Optical emission spectra showed the presence of argon,
nitrogen and oxygen ions in the plasma. Both the rough physical, sputtering, and formation offunctional groups [2,3] may have contributed to the increase in hydrophilicity of the samples.
The lowest average contact angle of 19.51° was achieved at nitrogen-argon flow rate ratio of
30:10 SCFH. Figure 2(c), condensation test using a helmet visor showed that visibility
remained clear on the left side, treated with APJ, because there was no formation of tiny
droplets. The aging test showed that the contact angle of the treated samples increased about
8~10° after 3 hours and it remained fairly constant at least up to 7 hours.
1. Masood, S.H., S.A. Mohamed, J Bagh Coll Dentistry 24, (2012)
2. T. Yuji, Y. Suzaki, T. Yamawaki, H. Sakaue, H. Akatsuka, Japanese Journal of Applied
Physics 46, pp 795-798, 2007
3. M. Nie, P. Patel, K. Sun, D.D. Meng, 4th IEEE International Conference on Nano/MicroEngineered and Molecular Systems (Hong Kong, China), pp 1017-1020, 2009- 176 -

P2-3
Ion species separated atmospheric pressure plasma source
M. Wada, Y. Fujita, T. Kasuya, M.C.C. Lacdan, J. Matsunaga
Graduate School of Science and Engineering, Doshisha University, Kyotanabe, Kyoto 610-0321, Japan

Atmospheric pressure plasmas can deliver activated molecules and atoms on a surface in an
atmospheric environment. They modify surface properties of materials without the usage of
wet reactive chemical solution. Highly collisional characteristics of these plasmas allow one topredict their behaviour based on two transport constants: diffusion coefficient and ion mobility.
The mobility of an air ion in atmospheric gas depends upon its mass, and several types of ion
mobility spectrometers have been developed for trace chemical analyses [1], while a Gerdien
condenser has been utilized to measure the ion concentration in the high altitude atmosphere [2].
A Gerdien condenser may realize high throughput of ions with reasonable resolution, and an
atmospheric pressure ion source development has been started to realize transport of species
separated ion flow.
The mobility separated
atmospheric pressure
ion source consists ofthree components: theplasma production part,ion mobility separation
part, and the ion flow
control part. Figure 1
shows the structure of
the current design. The
plasma is excited by thehigh voltage induced
Fig. 1. Atmospheric pressure source with ion mobility selector unit.
needle electrodes. The
flow field of ions are
created by a couple of DC electric fans attached to a 10 cm wide, 10 cm high, and 25 cm long
rectangular flow channel. Inside of the channel, two concentric cylinders form an electric fieldin the volume between the electrodes to guide ions of proper mobility into an opening. The
produced flow of ions are guided to a cylinder to be directed to work piece surface.
The current source employs a modular design, and the plasma production part can be changed
to a RF plasma source or a microwave plasma source. The front end that guides the ion flow
can be replaced with the component of different dimensions, or different material. Space
charge can cause a serious problem to reduce the throughput of the source. The source cancontain another tiny inert gas positive ion source at the center of the ion production part tomerge reactant negative ions to inert gas positive ions. The effectiveness of merged ion flow tothe enhancement of the flux of target ion is being investigated by using a Gerdien condenser forion species measurement.
[1] G.A. Eiceman, Z. Karpas, H. H. Hills, Jr. “Ion Mobility Spectrometry,” CRC Press, (2013).
[2] P.E. Raj, P.C.S. A.M. Selvam, A.S.R. Murty, Advances in Space Research, 12, 321 (1992).

- 177 -

P2-4
Degradation of acetic acid in water using gas-liquid plasma
with SPG membrane
Guanyang Tang, Atsushi Komuro, Kazunori Takahashi and Akira Ando
Department of Electrical Engineering, Tohoku University, Sendai 980-8579, Japan
E-mail: tang@ecei.tohoku.ac.jp

Atmospheric-pressure and gas-liquid plasmas have been widely investigated in association
with water purification because the plasmas can produce a large amount of active species suchas ozone (O3) and hydroxyl radical (∙OH), especially the hydroxyl radical which has a very highoxidation potential (2.81V) is well known to play important roles in degradation of persistentorganic pollutants (POPs) [1]. Since the hydroxyl radicals are produced at the gas-liquidinterface via the discharge process, the energy consumption at the interface rather than that at
the gas phase would yield the efficient production of the hydroxyl radicals. Therefore the better
production efficiency of the hydroxyl radicals can be obtained for the large surface-to-volume
ratio (surface area of the gas liquid interface / volume of the gas phase) [2].
Here a Shirasu porous glass (SPG) membranes tube which has dense micro-pores (φ50μm) isused to introduce a large number of micro gas bubbles into our previous gas-liquid plasma
reactor as shown in Figure 1 and to increase the surface-to-volume ratio. Comparing with theprevious reactor which has larger holes (φ1mm) [1], the generation of hydroxyl radical isobserved to be enhanced as shown in Figure 2 and the degradation of acetic acid is also
predictable to be enhanced. The dependences of acetic acid degradation on the shapes ofelectrodes and gas flow rate are also discussed with the electric field distribution analyzed bysolving two-dimensional Poisson’s equation. More details about plasma generation anddegradation process of acetic acid will be shown in the presentation.

.
Figure 1. Schematic of discharge
reactor with SPG membrane tube.

Figure 2. Time variation of H2O2
concentration with SPG membraneand previous reactor (φ1mm).

[1] G. Tang, et al. J. Plasma Fusion Res. 10, 3406022 (2015).
[2] Malik M A, J. Plasma Chemistry and Plasma Processing, 30. 1 (2010)- 178 -

P2-5
Fourier Transform Infrared Spectroscopy of Plasma Polymerized
Polyaniline
S. M. Palardonioa, M. R. Vasquez Jr.b, H. J.RamosaaPlasma Physics Laboratory, National Institute of Physics, University of the Philippines, Diliman 1101
Quezon City, Philippinesb
Plasma-Material Interactions Laboratory, Department of Mining, Metallurgical, and Materials
Engineering, University of the Philippines, Diliman 1101, Quezon City, Philippines.

This research studied the effect of polymerization time and discharge current to the molecular
structure of plasma polymerized polyaniline (PAni). Plasma polymerization of PAni on 1x1 cm
glass substrates was performed in a DC plasma-enhanced chemical vapor deposition facility.
Prior to polymerization, soluble oils and dusts on the glass substrates were removed byultra-sonification in 95% ethanol and then air-dried. The glass substrates were lowered in to a
6-cm diameter electrode that serves as a cathode. The anode is 1.5 cm above and has a shower
head configuration where precursor gasses can be introduced. The glass substrates were plasma
cleaned to further remove adsorbed compounds. Plasma cleaning was done with 5 minutes ofargon discharge at <0.4 Pa base pressure, 15 Pa filling pressure, 25 mA current, and 0.40 kV
discharge potential. The polymerization was then carried out at a base pressure <0.4 Pa, fillingpressure of 80 Pa, discharge potential at 0.48 to 1.00 kV while the discharge current was varied
at 1 mA, 2 mA, and 3 mA, and polymerization times of 5 mins, 15 mins, and 30 mins. The
plasma polymerized PAni was then analyzed by Fourier Transform Infrared Spectroscopy inAttenuated Total Reflectance mode (ATR-FTIR). ATR-FTIR spectra revealed that all plasma
polymerized PAni contains a benzenoid peak at 1600 cm-1 and a secondary amine peak at 3400
cm-1. In addition, peaks at 760 cm-1 and 870 cm-1 suggest a cross-linked polymer in ortho andpara position, respectively. Increasing the discharge current was found to introduce quinoid
units appearing at 1500 cm-1 and induce aromatic ring opening as suggested by the aliphatic
C-H bond peaks at 2900 cm-1. The effect of duration of polymerization was also found to
introduce quinoid units but aromatic ring opening was greatly enhanced as polymerization timeis increased. It is concluded that introduction of quinoid units and aromatic ring opening isdependent on discharge current and polymerization time.

- 179 -

P2-6
Effect of electrode shapes and the working gases on production amount ofhydroxyl radicals in atmospheric-pressure plasma jets
Y.Okunoa, H.Matsuurab, S.Okudab, T.Fujiyamaa,M.Furutab, Y.Takemurac, S.Kadod
aGrad. Sch. Eng, Osaka Pref. Univ., 1-1 Gakuen-cho, Naka-ku, Sakai City, Osaka, 599-8531, Japan.
bOsaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai City, Osaka, 599-8531, Japanc
Kinki University, 3-4-1 Kowakae, Higashiosaka City, Osaka 577-8502, Japand
Institute of Advanced Energy, Kyoto University, Gokasho, Uji, 611-0011 Japan

Atmospheric-pressure plasma jets have attracted attentions as a sterilization method in the fieldof medical applications and biotechnology because of little heat damage and rich reactive
oxygen/nitrogen species production. Hydroxyl radicals (OH) have a high sterilization effects in
the reactive species due to the strong oxidizing power and the light emission in the vicinity of
309 nm.
Our group has proposed to use plasma degradation of toxic phorbol esters and disinfection
procedure of Escherichia coli [1]. OH are directly not generated from a working gas of theplasma jets but by involving ambient air and/or water [2]. Thus it is important to improvegeneration efficiency of OH by optimizing the shape in vicinity of the outlet of the plasma jets.
The degradation and the disindection require a huge amount of expensive helium (He) gas forplasma production. Its running cost is not economical. Argon (Ar) gas, which is cheaper than He,
is one possible alternative. Both He and Ar plasma jets are able to eject by one plasma source. In
this study, the effect of the electrode shapes and the working gases on the production amount ofOH.
Emission spectral analysis are able to measure the production amount of OH. Deferent of lightintensity about 309 nm by the each electrode shapes and the working gases is analysed byspectrometer (Maya2000pro, Ocean Optics). The optimal conditions of OH generation will be
discussed in the conference.
This work was supported by the "ZE Research Program, IAE (ZE27B-4).

[1] H. Matsuura, Y. Onishi, S. Kongmany, M. Furuta, K. Imamura, Y. Maeda, S. Okuda,
Plasma Medicine, 4(1-4), 29 (2014).
[2] Y. Takemura, S. Umeji, K. Ito, S. Furuya, M. Furuta, Plasma Medicine, 4(1-4), 89 (2014).

- 180 -

P2-7
Effects of a magnetic Laval nozzle to the thrust performanceon an MPD thruster
Hiroaki Nabuchi, Kiyotaka Suzuki, Yohei Kobayashi, Atsushi Komuro, Kazunori Takahashi
and Akira Ando
Department of Electrical Engineering, Tohoku University, Sendai 980-8579, Japane-mail address of submitting author : nabuchi@ecei.tohoku.ac.jp

Recently, various missions of space explorations are proposed and intensive development of
space propulsion engines with high thrust and high specific impulse are urgently required. A
Magneto Plasma Dynamic (MPD) thruster is one of the candidates for main engines in manned
deep space missions because of its high operating electric power above a hundred of kW and theresultant large thrust.
To improve the thrust performance, an applied field MPD thruster has been investigated withvarious magnetic field configurations and characteristics of the fast-flowing plasma generated
in the MPDs have been investigated [1,2]. It has been reported that both the thrust and theplasma flow velocity can be increased by applying a diverging magnetic field [3]. This
acceleration seems to be due to the interaction between the discharge current and the appliedmagnetic field. However, it has been also observed that the fast-flowing plasma is thermalized
downstream of the divergent magnetic field.
In this study, the magnetic Laval nozzle shown in Fig. 1 is additionally applied to reaccelerate
the plasma via a momentum conversion process in the cross-section-varying plasma flow. The
uniform field B0 = 0.10 T, the diverging magnetic nozzle Bzc = 0.28 T, and the magnetic Laval
nozzle BzL corresponding to the magnetic nozzle throat set at Z = 100 mm downstream of the
MPD thruster are superimposed and helium is chosen as a working gas. Five configurations
being tested are labeled as BzL = 0, 0.10, 0.15, 0.20 and 0.25 T, respectively.
The maximum thrust of 6.9 N and the maximum
plasma flow velocity of 64 km/s at Z = 200 mm are
obtained for the case of BzL = 0.15 T. Assessment of
the Lorentz force suggests that the Lorentz force
exerted to the plasma flow is mainly in radial
direction and does not explain the thrust increase.
The detailed results of the thrust, the axial plasma
flow velocity, the ion temperature and the ion densityof the MPD thruster will be shown and the physics
underlying in the magnetic Laval nozzle attached to
the MPD thruster will be discussed in thepresentation.
Fig.1 The magnetic field configurations[1] A.Sasoh, et al., Journal of Propulsion and Power 11 , 351 (1995)[2] M. Inutake, et al., Plasma Phys. Control. Fusion 49 A121 (2007).
[3] Y. Izawa, et al., JPS Conf. Proc. 1, 015046 (2014)- 181 -

P2-8
Design of electron supplier in a new concept of neutron generator for BNCT
D. Adachi, T. Matsui, M. Yanagi, T. Takahashi, and H. MomotaaaGunma University,1-5-1 Tenjin-cho, Kiryu 376-8515, Japan
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Compare to conventional methods of cancer therapy, boron neutron capture therapy (BNCT)
is more effective to destroy cancer cells and to lessen damage of healthy cell. Since BNCT
needs thermal or epithermal neutrons with the yield of 109 particles/sec, an atomic reactor wasformerly used as a neutron source [1]. It is not available in Japan after the nuclear power plant
accident in 2011. Nowadays, a neutron source with an accelerator has been developed [2,3].
The accelerator neutron sources, however, need large space to install in an urban hospital andits construction cost will be high. Therefore in this study we propose the new concept neutron
source which can be installed in hospital room.
The new concept neutron source is electrostatic confinement fusion equipment, which
generates neutrons by the D-D reaction [4]. Its size is estimated to be smaller than traditional
neutron sources. Deuterium ions sources are installed at both ends of linear neutron generator.
Beam ions are accelerated axially in the electrostatic potential gap of 30 kV, and counter beam
particles react in the reaction chamber in the central region of the machine. In order to suppress
the deleterious two stream instability, the axial magnetic field is applied by the solenoid coil.
The electron supplier controls the electron density in the reaction chamber so that the plasma
in the chamber maintains to be in quasi-neutral condition. The electron supplier has two coils
that generate the cusp magnetic field, and it separates thermal electrons from beam ions[5]. The
electron generator installed in the line cusp region supplies electrons into the reaction chamber,and high-energy beam ions never go into this region. In the present study, we will show
calculation results of orbit of beam ions and thermal electrons in the cusp magnetic field, andwill design the efficient electron supplier.

[1] T. Yamamoto, A. Matsumura, K. Yamamoto et al., Radiation Research. 160 (2003) 70.
[2] H. Tanaka, Radioisotopes. 64 (2015) 29. (in Japanese)[3] H. Tanaka, Y. Sakurai, M. Suzuki et al., Appl Radiat Isot. 69 (2011) 1642.
[4] H. Momota and G. H. Miley, J Fusion Energ.. 28 (2009) 191.
[5] A. Taniguchi, N. Sotani, Y. Yasaka et al., J. Plasma Fusion Res. SERIES. 9 (2010) 237.

- 182 -

P2-9
Beam ion confinement in a new concept of neutron generator for
BNCT
T. Matsui, D. Adachi, M. Yanagi, T. Takahashi, and H. Momotaa
Gunma University, 1-5-1, Tenjin-cho, Kiryu 376-8515, Japana

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

Boron neutron capture therapy (BNCT) is a new cancer treatment that has been attracting
attention in recent years. It is possible to destroy cancer cells selectively and reduce risk ofdamaging healthy cells [1]. Although BNCT requires a neutron source with the yield of 109
particles/sec or more, no sufficient source is available in the present circumstances. Recently,
accelerator-based neutron source is further along in development [2].
In this report, we propose a new neutron source using nuclear fusion reaction [3]. It is expected
to be more compact than the accelerator. Therefore, this will be easily installed to the hospital.
We propose a neutron source consisting of ion source, electronic separator, and reaction
chamber. In the present study, we report on beam ion confinement in the reaction chamber. The
beam ion is accelerated by the electrostatic potential gap at both ends of linear confinement
system. Since counter-beam configuration will cause the deleterious two-stream instability, thesuppression effect by the external axial field is studied by 2-dimensional hybrid simulation.
Detailed results will be shown in our presentation.
[1] R. F. Barth, A. H. Soloway, R. G. Fairchild, and R. M. Brugger, Cancer 70 (1992) 2995.
[2] H. Tanaka, Y. Sakurai, M. Suzuki et al., Appl Radiat Isot. 69 (2011) 1642.
[3] H. Momota and G. H. Miley, J Fusion Energ. 28 (2009) 191.

- 183 -

P2-10Formation Mechanism of Periodic Nano-Grating Structure
by Surface Plasma Wave
A. M. Gouda, H. Sakagami,a M. Hashida,b and S. Sakabeb,
Department of Physics, Nagoya University, Nagoya, 464-8602, Japana
National Institute for Fusion Science, Toki, Gifu 509-5292, Japanb
Institute for Chemical Research, Kyoto University, Uji, Kyoto 611-0011, Japan
E-mail: amany.gouda@nifs.ac.jp

This abstract aims to introduce one of formation mechanisms for the periodic nano-grating
structure, which is an emerging multidisciplinary research area at the multifaceted edge at theforefront of the physics of plasmas and gas discharges, nano-science and nanotechnology,
astrophysics, materials science and engineering, and structural chemistry [1]. And show theimportance of the plasma environments in nano-scale processes spanning from astrophysics toplasma-aided nanofabrication in the laboratory.
The formation mechanism of the periodic nano-grating structure has been investigated andconfirmed by 2D PIC code [2] and with irradiating the laser pulses of 500 fs pulse, wavelength
800 nm, incidence angle 0º, p-polarized and intensity of 1016 W/cm2-µm2. The surface plasma
wave plays an expressive role with the oscillating two streams instability in explaining oursuggested mechanism. The surface plasma waves are exited by the laser beam and propagate toupper and lower direction composing the standing wave.
In laser plasma interaction, there is a force coupled to the particles in a somewhat subtle way
and is called the Ponderomotive force. The Ponderomotive force is a nonlinear force that acharged particle experiences in an inhomogeneous oscillating electromagnetic field. The
Ponderomotive force by the standing wave E- field forms perturbations at the boundary, which
can be grown by the oscillating two stream instability to form the periodic nano-grating
structure. Fig.1 shows the propagation of the electric field component in y-direction by thedashed line, which starts from time Tstart=268.47[fs] with time difference Δt=0.27[fs]. This is a
proof of existence of the surface plasma wave in the upper part of the electric field.

Fig.1. Time evolution of the upper part of the electric field component in y-direction.

[1] K. Ostrikov,  “Colloquium:  Reactive  plasma  as  a  versatile  nanofabrication  tool,”  Rev.  Mod.  
Phys., 77 (2005) 489–511.
[2] H. Sakagami and K. Mima: Proc. 2nd Int. Conf. Inertial Fusion Sci. Appl. (2001)
380-383.

- 184 -

P2-11Krylov Subspace Method with Communication Avoiding
Technique for Linear System Obtained from Electromagnetic
Analysis
S. Ikunoa , G. Chena , S. Yamamotoa , T. Itohb , K. Abec , and H. Nakamurad,ea School of Computer Science, Tokyo University of Technology, Hachioji, Japanb College of Industrial Technology, Nihon University, Narashino, Japanc Faculty of Economics and Information, Gifu Shotoku Gakuen University, Gifu, Japand Department of Energy Engineering and Science, Nagoya University, Toki, Japane National Institute for Fusion Science, Toki, Japan

In the present study, Krylov subspace method with a communication avoiding technique isapplied for the linear system obtained from electromagnetic analysis. It is well known thatthe algorithm of Krylov subspace method is very simple, and the most of all the procedure of
the method is constituted by addition of vectors, inner products and multiplication of matrices and vectors. These operations are very easy to get a high performance by parallelization.
However, the communication between processing units (PUs) must be necessary for parallelized inner products calculation, and the amount of the communication time increases as
the size of the system increase. That is to say, the communication time is bottle neck for theeﬀective parallelization.
The communication avoiding technique is one of a settlement for above issue [1, 2]. Although
the algorithm of Krylov subspace method with the communication avoiding technique is
diﬀerent from original method, the theoretical methodology is equivalent. In the technique,inner products is rewritten to the recurrence formular using the basis of Krylov subspace, andthe inner products are transcribed into scalar calculation. As the results, communications
between PUs can be gathered in only one time.
The purpose of the present study is to develop Krylov subspace method with the communication avoiding technique, and to applied for the linear system obtained from electromagnetic
analysis. In addition, the parallelized method is implemented on Graphics Processing Unit
(GPU) to obtain high performance calculation.

[1] Mark Hoemmen, “Communication-Avoiding Krylov Subspace Methods”, The doctoral
dissertation of Uni- versity of California at Berkeley, Berkeley, CA, USA, 2010.
[2] J. W. DEMMEL, M. H. and MOHIYUDDIN, M., ”Minimizing communication in sparse
matrix solvers”, in Proceedings of the 2009 ACM/IEEE Conference on Supercomputing(2009).

- 185 -

P2-12Magnetic property analysis for spherical shell magnetic substance using
Monte Carlo simulation
K.Terashima, K. Yamaguchi, T. Uchimotoa, T. Takagia, H. Nakamurab
Fukushima University, Kanayagawa 1, Fukushima, 960-1296, Japan
Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, 980-8577, Japanb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japana

In this study, magnetization property of minute spherical shell magnetic substance with hollow
inside was analyzed using Monte Carlo (MC) simulation. The spin behavior of spherical shellwith different thickness and size was verified. The spherical shell magnetic substance isincluded to many iron or nickel-based structural materials. For example, in spheroidal graphite
cast iron, magnetic layer was deposited around spheroidal graphite. It has been much studied for
relationship between mechanical strength of material and shape or composition of magnetic
layer [1]. In addition, Ni-base super alloy Alloy600 has ferromagnetism around grain boundarywith degradation progress [2]. Here, magnetic layer around grain boundary is consideredspherical shell magnetic substance. Thus minute spherical shell magnetic substance has
influence for the property of material and knowing the magnetic property is very important fromviewpoint of magnetic nondestructive evaluation (NDE).
In this simulation, some clusters were prepared with magnetic site on a lattice point in simple
cubic lattice which has different thickness and radius of spherical shell. Following Hamiltonianis set:H  HJ  HD  HB

 S S

3
 i j




   J ij S i  S j  D 

S

r
S

r
i
ij
j
ij   B  S i .
3
5

near
all  r
i
rij
 ij


(1)
Each term of Jij, D and B represents exchange interaction constant for i-th site and j-th site,
magnetic dipole interaction constant and external magnetic field, respectively. Here Si denotes
the magnetic moment of the magnetic site of i-th cell and rij represents the vector between i-thsite and j-th site.
In the magnetic reversal process, island-shape such as magnetic domain is formed and spin
behavior becomes complex. Additionally, reversal process has more complex due to size andthickness of spherical shell.
The behavior of spin in spherical shell magnetic substance was analyzed by MC simulation.
Therefore, the relationship between macroscopic and microscopic magnetic property was
cleared and shown the possible to estimate the state of microscopic region from macroscopic
magnetic measurement for iron or nickel-based structural materials with spherical shellmagnetic substance. As a result, the possibility of application was shown for magnetic NDE.
Furthermore, the high-precision analysis was expected by comparison for experimental result of
local magnetic property using μ-MOKE.
[1] N. Shinohara, N. Suura, N. Hattori and R. Sakuragi, Trans. Jpn. Soc. Mech. Eng. A, 79(797)(2013) 13-22.
[2] R. Oikawa, T. Uchimoto and T. Takagi, Trans. Jpn. Soc. Mech. Eng. A, 75(760) (2009)
1777-1783.

- 186 -

P2-13
Dissipative particle dynamics simulation of self-assembly inbolaamphiphilic solution
S. Fujiwara, Y. Takahashi, H. Ikebe, T. Mizuguchi, M. Hashimoto, Y. Tamuraa, H. Nakamurab,
R. Horiuchib
Kyoto Institute of Technology, Matsugasaki, Sakyo-ku, Kyoto 606-8585, Japana
Konan University, 8-9-1 Okamoto, Higashinada-ku, Kobe 658-8501, Japanb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanThe spontaneous formation of structures has recently become the focus of interest in
nonequilibrium and nonlinear systems such as plasma systems and soft matter systems. We aim
at improving plasma confinement by clarifying the self-organizing properties common to thesesystems. To this end, we investigate the structure formation in soft matter systems such as the
micelle formation and the mesophase formation in amphiphilic solutions. Amphiphilicmolecules such as lipids and surfactants are composed of both hydrophilic and hydrophobicgroups. In aqueous solvents, such amphiphilic molecules often spontaneously self-assemble
into various structures such as micelles, bicontinuous structures and vesicles [1,2]. Althoughnumerous computer simulation studies have so far been done on the structure formation ofamphiphilic molecules, each of which consists of a hydrophilic head group and a hydrophobic
tail group, there have been few theoretical and simulational studies on the structure formation of
bolaamphiphilic molecules, each of which consists of a hydrophobic stalk and two hydrophilicends. The purpose of this study is to clarify the effect of the interaction difference between thetwo different hydrophilic ends, ∆a , on the self-assembly in bolaamphiphilic solutions. With aview to investigating the self-assembly in bolaamphiphilic solutions at the molecular level, we
perform the dissipative particle dynamics simulations of bolaamphiphilic molecules with
explicit solvent molecules and analyze the formation processes of micelles and vesicles. Our
simulations show that various kinds of self-assembled structures such as worm-like micelles,
vesicles and nanotubes are obtained. It is also ascertained that the vesicles formed at small ∆achange to the worm-like micelles as the interaction difference, ∆a , increases.

[1] J. N. Israelachvili, Intermolecular and Surface Forces, 2nd ed. (Academic Press, London,
1992).
[2] I. W. Hamley, Introduction to Soft Matter, Rev. ed. (J. Wiley, Chichester, 2007).

- 187 -

P2-14
Improvement of divertor probe array for heat flux measurement of
Heliotron J divertor leg plasma
H. Matsuura, Y. Umeda, D. Odaa , T. Mizuuchia , and Y. Suzukib
Osaka Prefecture University, 1-2 Gakuen-chi, Sakai Naka-ku 599-8570,Japana Kyoto University, Gokasho, Uji 611-0011, Japanb National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanmatsu@me.osakafu-u.ac.jp

In the design of fusion reactors for high power and long pulse operation, vast heat ﬂux is
expected to ﬂow onto divertor target plates. In order to reduce this heat load, the divertor
design on stable detached plasma formation must be realized. Compared with Tokamak,helical systems have intrinsic diverted ﬁeld lines and large variety of divertor conﬁgurationsuch as island divertor. The asymmetry of divertor ﬂux is still an open question and few
report on heat ﬂux reduction with detachment has been published yet.
Heliotron J is a medium sized helical axis Heliotron device with a helical winding coil of
L = 1/M = 4. In the standard conﬁguration, so-called whisker ﬁeld line structure covers thelast closed ﬂux surface, and some of ﬁeld lines crosses the inner wall of Heliotron J, forming
four localized divertor legs in one toroidal period [1]. By adjusting coil current ratio, island
divertor conﬁguration is also available. From the beginning of Heliotron J experiment series,
multi channel langmuir probe array has been equipped to study particle ﬂux behavior andexistence of clear up-down asymmetry was reported [2]. As for heat ﬂux, however, suchobservation is lacking, since no calorimeters or no thermal probes is prepared in the divertor
probe array. According to sheath theory, heat ﬂux through the sheath is the function of thesheath potential drop and also depends on ion parameters such as its temperature or energy
reﬂection coeﬃcient.
In this work, new calorimeter chips is designed to combine with conventional divertor probearray, which is a SUS right-angled parallelepiped with many Molybdenum probe pins on one
plasma facing surface. Since magnetic ﬁeld lines in the divertor leg are almost parallel tothis surface. So the calorimeters on this surface may fail to detect the divertor heat ﬂux.
calorimeters on the side surface is also necessary. For choice of the calorimeter material, notonly thermal diﬀusion time but also estimated temperature increment must be considered.
Copper has fast thermal diﬀusion necessary for fast heat ﬂux measurement, but its temperature might accumulate during plasma shots and reach the melting point. In the conference,the size and shape of designed calorimeter head will also be presented. This work is partially performed with the support and under the auspices of the NIFS Collaborative Research
Program. (NIFS14KUHL061)[1] T. Mizuuchi et al., J. Plasma Fusion Res. 3 (2000) 192.
[2] W. Ang et al., J. Plasma Fusion Res. 5 (2002) 292.
[3] H. Matsuura et al., 22nd ITC (2002) P1-23.

- 188 -

P2-15
Study of Heat and Particle flux in the case of Gas Injection in the D-moduleof GAMMA 10/PDX
M. S. Islam, Y. Nakashima, H. Matsuuraa , K. Ichimura,
M. M. Islam, K. Shimizu, K. Fukui, M. Ohuchi, K. Nojiri, N. Ezumi,
M. Sakamoto and T. Imai
Plasma Research Center, University of Tsukuba, Ibaraki 305-8577, Japan
Radiation Research Center, Osaka Prefecture University, Osaka 599-8570,,Japana

Influence of gas injection in the D-module of GAMMA 10/PDX has been investigated based on
calorimeters and Langmuir probes data in this research. In GAMMA 10/PDX, divertor
simulation experiments have been started by using a divertor simulation experimental module(D-module) [1]. In the D-module of GAMMA10/PDX as shows in Fig.1, a V-shaped target
made of tungsten has been installed. A set of Langmuir probes and calorimeters has been
installed at the upper and lower plates respectively of the V-shaped target for simultaneous
measurement of particle and heat flux. Heat flux has been measured by the calorimeter. Electron
temperature (Te), electron density (ne) and ion saturation current (Ii-sat) have been measured by
the Langmuir probe. According to the increase amount of gas injection, reduction of Te, Ii-sat, neand heat flux have been observed. In the case of gas injection through either gas injection fromthe port at the front-side (F) or the rear-side (R) , the reduction of heat flux, ne, Te and Ii-sat
remained almost same. However, a remarkable reduction of the heat flux, Te and Ii-sat have been
observed when gas was injected through both gas injection ports (F) and (R). In the Ar injection
experiments, H2 gas was injected simultaneously to examine the effect of molecular process on
detached plasma formation. In this case, both the heat flux and ion saturation current are
drastically reduced. The detail of the experimental results will be reported in the presentation.
ASDEX
gauge

Only-F, H - 2000mbar
2Both F & R. H2- 2000mbar
W/O H

2

Only-R, H -2000 mbar
2

0.2

Langmuir Probe

0.15

Plasma0.1

0.05

Piezo-valve

Front, (F)Piezo-valve

Rear, (R)

0
1060

1065

1070

1075

1080

1085

1090

Z [cm]

Gas Source

Fig.2 Distribution of heat flux on Z axis in the D-module of
GAMMA10/PDX. (Pressure indicates the plenum
Fig.1 Schematic view of D-modulepressure of the gas source.)[1] Y. Nakashima et al. J. Nucl. Mater. 438 (2013) S738-S741.

- 189 -

P2-16
Study of the axial plasma confinement for high particle flux operations in
GAMMA 10/PDX
K. Ichimura, Y. Nakashima, M. M. Islam, M. S. Islam, K. Shimizu, K. Fukui,
M. Ohuchi, M. Hirata, R. Ikezoe, M. Yoshikawa, N. Ezumi, M. Sakamoto,M. Ichimura and T. Imai
Plasma Research Center, University of Tsukuba, Tsukuba, Japan

Enhancing the parameter of the end-loss ion flux is an important issue in the divertor simulation
studies in GAMMA 10/PDX [1]. In recent experiments, the end-loss ion flux relevant to the
ITER scrape off layer plasma (1.6×1023 particles/m2s) was obtained by superimposing theadditional ICRF wave heating into minimum-B regions of the tandem mirror. With the ICRF
heating, large increase in the electron line-density in the central-cell (NLCC) and the end-loss
particle flux were observed [2]. The behavior of the electron line density, plasma temperatureand the amount of end-loss ion flux during the experiments suggested that the mechanism of theaxial ion loss was changed due to the increase of ion collision frequency. In standard operationsof GAMMA 10/PDX, plasma has high ion temperature (Ti = 4000 eV, Ti// = 400 eV) and theratio of mirror length and the axial mean free path of ions is small (L/i// ≈ 0.01). In contrast
with the standard operations, plasma parameters in high-density operations stays in the range ofhigh collision frequency and the ratio L/i// is about 0.6. In such collisional plasma, the end-loss
ion particle flux show linear dependence on the plasma density instead of the parabolic
dependence expected in plasmas of low collision-frequency [3]. Figure 1 shows theexperimental and theoretical dependence of the end-loss ion particle flux on NLCC. The
dependence in such wide range of NLCC is experimentally observed for the first time. The
result is useful to discuss the future prospect of the particle flux enhancement in the tandem
mirror machine. In the high collision frequency plasma, it is notable that plasma potential largely changes the dependence of the particle flux on the plasma density. In the presentation,
detailed results of experiments and the discussion on the ion confinement time will be reported.

Fig. 1 Dependence of the end-loss ion particle
flux on the line density at the central-cell.

[1] Y. Nakashima et al., J. Nucl. Mater. 438 (2013) S738.
[2] K. Ichimura et al., Trans. Fusion Sci. Technol. 63 1T (2013) 209-212.
[3] R. Minai et al., J. Phys. Soc. Jpn. 66 7 (1997) 2051-2056.

- 190 -

P2-17
A study on hydrogen transport in liquid lithium
under steady state plasma bombardment
H. Bi 1,*, Y. Hirooka 1,2, J.Yagi 2, Y. Xu 1
1

The Graduate University for Advanced Studies, 322-6 Oroshi, Toki 509-5292, Gifu, Japan
2
National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292, Gifu, JapanThe application of liquid metals as plasma facing materials draws an increasing interest as apotential means to resolve the technical issues associated with exhaust power and particle
handing in magnetic fusion devices. Lithium is one of the promising candidates because of its
low atomic number and melting point, and high absorptivity of impinging species. It has beenused in fusion devices including NSTX[1] and EAST[2]. In these confinement devices, it isbelieved that hydrogen transport in liquid lithium plays an important role in determining edge
plasma characteristics and hence core confinement performance. Nonetheless, little information
is available on hydrogen transport in liquid lithium under steady state plasma bombardment,
presumably due to the difficulty of handling liquid metals.
In the present work, an innovative experimental technique holding liquid lithium on a mesh
substrate by surface tension has been put together for the plasma-driven permeation (PDP)
experiments in a plasma-wall interactions facility: VEHICLE-1[3]. Temperature effects on PDP
behaviour have been observed in the temperature range from 313oC to 402oC. It has been foundthat hydrogen permeation through liquid lithium is recombination limited judged from thebreakthrough curve of permeation flux. This shows a similar behaviour to prior PDP experiments
on titanium membranes, which are conducted because hydride is formed both in titanium andliquid lithium. As a result, H recombination coefficients are obtained and compared withliterature data in Fig. 1. [4]. No convection is considered in this work.
o400

Temperature( C)

350

300

H in Li in this work
Linear fit of data in this workD in Li by R.J. Baldwin

4

H recombination coefficient (cm /s)

450
1E-25

1E-26

1E-27

1E-28

1.4

1.5

1.6

1.7

-1

1000/T(K )Fig.1. Hydrogen recombination coefficient obtained from PDP experiments.
The thickness of liquid Li membrane is about 2mm. H+ incident flux is about 1x1016 H/cm2s with 100eV.

[1] H.W. Kugel et al, Fusion Eng. Des.87 (2012) 1724– 1731
[2] J.S. Hu et al, Fusion Eng. Des.89 (2014) 2875–2885[3] H. Zhou et al, Journal of Nuclear Materials 455 (2014) 470–474[4] M.J. Baldwin et al, Journal of Nuclear Materials 306 (2002) 15–20- 191 -

P2-18Characterization of detached recombining plasma in Pilot–PSI
Y. Hayashi, K. Ješkoa, H.J. van der Meidenb, J. Vernimmenb,
T.W. Morganb, N. Ohno, S. Kajitac
Graduate School of engineering, Nagoya University, Nagoya, Aichi 464-8603, Japana
CEA, IRFM, F-13108 Saint-Paul-Lez-Durance, Franceb
FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, partner in the Trilateral
Euregio Cluster, Eindhoven, the NetherlandscEco Topia Science Institute, Nagoya University, Nagoya, Aichi 464-8603, JapanTe [eV]19-3

ne [10 m ]Plasma detachment accompanied by volumetric recombination processes is part of the key
physics required for the success of magnetically confined fusion. Linear plasma devices,which simulate the magnetic field geometry from the upstream scrape–off layer (SOL) to thedivertor plate in magnetically confined fusion devices, have previously made great
contributions to understanding plasma detachment [1, 2]. However, it is necessary to conduct
plasma detachment experiments under the high particle flux conditions relevant to detacheddivertor operations in ITER. The linear plasma device Pilot–PSI is able to generate highdensity plasma (> 1020 m-3) with high ion flux (> 1024 ions m-2s-1) [3]. However the essential
plasma parameters such as electron temperature, Te, and electron density, ne, in detachedrecombining plasmas have not been sufficiently well characterised. In this study, in order tounderstand the fundamental properties of detached recombining plasma in Pilot–PSI, theradial profiles of plasma parameters such as Te and6● Doube probe
ne have been measured by reciprocating electrical■ Single probe
probes and Thomson scattering (TS) measurement◆ Thomson
4focusing on the neutral pressure, P, dependence.
Electrical probes support the TS especially in2plasma peripheral region, where the scattering lightbecomes weak while TS can give an unambiguous(a)determination of the range of probe validity when0
its signal/noise ratio is high.
1.5
The experiments were performed by using
the single probe and double probe for Te and ne
1
measurements scanned across the plasma column
with a reciprocating arm and compared to TS.
0.5
Figure 1(a) and (b) show, respectively, the P(b)0dependence of ne and Te at plasma centre. Single
510
15
and double probe values of ne monotonically
Vessel pressure [Pa]decrease with increasing P, while the ne from TS
Fig. 1 Neutral pressure dependence of (a)has a peak at P ~ 8 Pa. On the other hand, Te fromne and (b) Te at plasma center by double
the probes is in a good agreement with the TS. The probe,
single probe and Thomson
results indicate that a steep gradient of Te leads to scattering measurements.
the plasma recombination and the decrease ofdensity as neutral pressure increases.
This work was supported by JSPS KAKENHI Grant Number (25289337). This work
is partially supported by NIFS/NINS under the project of Formation of International Network
for Scientific Collaborations and NIFS Collaboration Research Program (NIFS14KUGM094).
[1]
[2]
[3]

D. Nishijima, et al., Plasma Phys. Control. Fusion 44 (2002) 597.
E.M. Hollmann, et al., Phys. Plasmas 20 (2013) 093303.
G.De Temmerman, et al., Nucl. Fusion 51 (2011) 073008.

- 192 -

P2-19Erosion of bubble formed tungsten materials
S. Saito, M. Tokitania, H. Nakamuraa
Institute of Technology, Kushiro College,2-32-1 Otanoshikenishi, Kushiro 084-0916, Japana
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

When tungsten materials are exposed by helium plasma, bubbles are formed [1] at the surfaceof tungsten material in wide range of incident energy. The existence of bubbles change thecharacteristics of tungsten materials as plasma facing material such as sputtering yield, retention,and so on. The depth where bubble forms depends on the incident energy. To reveal the
sputtering
yield
and
erosionon
bubbleformedtungstenmaterials,
binary-collision-approximation-based (BCA-based) simulation is performed in this paper.
Figure 1 shows the simulation results of the time evolution of the surface of a tungsten material
irradiated by helium plasma with incident energy of 25 keV. The red and black dots denote thetungsten and helium atoms, respectively. This figure shows the erosion of the tungsten materials.
We investigate incident energy dependence of the erosion speed and the sputtering yield byperforming the BCA-based simulation (AC∀T code) for wide range of incident energy.
80.11 Å

0

19.5

97.4
175.3
Fluence [×1020 He/m2]

253.2[1] M. Tokitani, N. Yoshida, K. Tokunaga, H. Sakakita, S. Tiyama, H. Koguchi, Y. Hirano,
and S. Masuzaki: Plasma Fusion Res. volume:5 (2010) 012.
[2] S. Saito, A. Takayama, A. M. Ito, A. Kenmotsu, and H. Nakamura: Progress in Nuclear
Science and Technology, volume: 2 (2011) 44.
[3] S. Saito, A. Takayama, A. M. Ito, and H. Nakamura: Proc. 30th JSST Annual Conference(2011) 197.
[4] S. Saito, M. Tokitani, and H. Nakamura: Springer CCIS, volume:474 (2014) 176.

- 193 -

P2-20Hydrogen gas-driven permeation through F82H steel
coated with vacuum plasma sprayed tungsten
Y. Xu 1,*, Y. Hirooka 1,2, T. Nagasaka 1,2, H. Bi 1, J. Yagi 1,2
1

The Graduate University for Advanced Studies, 322-6 Oroshi, Toki 509-5292, Gifu, Japan
2
National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292, Gifu, Japan

Hydrogen isotopes transport behaviour through tungsten coated reduced activation ferritic steels
such as F82H has attracted increasing attention in the fusion engineering research community [1].
However, limited amount of literature data is available on the transport parameters related todiffusivity, permeability and solubility of hydrogen isotopes in tungsten coated F82H.
This paper reports on these parameters using a laboratory-scale plasma device: VEHICLE-1 [2]to investigate hydrogen gas-driven permeation behaviour in vacuum plasma sprayed tungsten(VPS-W) coatings on substrate of F82H at a temperature range of 200-500℃. A permeation
membrane is prepared. The thicknesses of the tungsten coating and the F82H membrane are 0.09
and 0.5 mm, respectively. It has been found that i) hydrogen permeation through VPS-W coated
F82H is diffusion-limited, ii) tungsten coatings have been demonstrated the efficacy to reduce thehydrogen permeability by about one order of magnitude under all the examined temperatures andiii) the values of the effective diffusivity for tungsten coated F82H is between that of the bare
F82H and tungsten and the activation energy is increased by ~40% compared to bare F82H (asshown in Fig. 1), which is in agreement with the indication of the multi-layer permeation theory[3].

Fig.1 Temperature dependence of hydrogen effective diffusivity for VPS-W coated F82H (W-F82H). The
thicknesses of the W coating and the F82H membrane are 0.09 and 0.5 mm, respectively. Literature data
[4-6] are shown for comparison.

[1] R. A. Anderl, D. F. Holland, G. R. Longhurst, J. Nucl. Mater. 176–177 (1990) 683-689.
[2] Y. Hirooka, H. Ohgaki, Y. Ohtsuka, M. Nishikawa, J. Nucl. Mater. 337–339 (2005) 585-589.
[3] J. Crank, The Mathematics of Diffusion, 2nd Edition. Clarendon Press Oxford: London, 1975.
[4] H. Zhou, Y. Hirooka, N. Ashikawa, et al., J. Nucl. Mater. 455 (2014) 470-474.
[5] E. Serra, G. Benamati, Mater. Sci. Technol. 14 (1998) 573-578.
[6] R. Frauenfelder, J. Vac. Sci. Technol. 6 (1969) 388-397.

- 194 -

P2-21
Tritium accumulation on carbon deposition layer and its decontamination
by glow discharge
Y. Torikaia), N. Ashikawab), K. Noborioa), Y. Uedac), K. Katayamad), Y. Nobutae),
N. Matsuyamaa) and K. Nishimurab)aHydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japanb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanc
Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871, Japan.
dDepartment of Advanced Energy Engineering Science, Kyushu University,
6-1, Kasugakoen, Kasuga-shi, Fukuoka 816-8580, Japane
Faculty of Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo, Hokkaido 060-8628,
Japan

E mail address: y2torik@ctg.u-toyama.ac.jp
From a safety point of view, assessment of tritium inventory in plasma-facing materials andselection of methods for tritium decontamination of the materials are important issues due torequirement of the tritium inventory limitation inside fusion machines. In this study, tritium
trapping in a carbon layer deposited on the 316L stainless steel (SS316L) was investigated.
Besides, methods of tritium decontamination of the carbon deposition layer by thermal releaseand glow discharge were tested.
Carbon layers were deposited onto SS316L substrates, 15×15×0.5 mm3 in size, bymagnetron sputtering. The carbon deposited SS316L specimens were placed in ultra-highvacuum system and was initially pre-heated at 423 at background pressure of (1.3 - 2.6)×10-6
Pa. Then the specimens were exposed at 423 K to gaseous deuterium - tritium mixture (7.2 %T) at the total pressure of 1.2 kPa for 3 h. For determination of the tritium content in thenear-surface layer of the carbon deposition, the imaging plate (IP) technique (Fujifilm
BAS-TR) was employed. The intensity of the photo-stimulated luminescence (PSL) was
analyzed with a laser scanner (Fujifilm FLA-7000). The detectable depth range of thistechnique for tritium in carbon is estimated to be < 1 µm.
Decontamination of the tritiated carbon deposited SS316L specimens were performed bydeuterium glow discharge.
Figure shows the results of the tritium decontamination of the carbon deposited SS316L
specimen due to D glow discharge. Left photo shows the carbon deposited SS316L specimen.
Part 1) shows the IP image of tritium loaded specimen, 2) shows the IP image after exposure
to D glow discharge for 30 min and 3) shows the IP image after exposure to D glow discharge
additional 60min. After D glow plasma discharges, surface tritium was removed incorresponding with surface carbon deposition layer was removed. Additionally, tritium
trapped on the SS316L surface was also removed by the D glow discharge. Therefore, it canbe concluded that glow discharge decontamination is a good method to remove tritium fromthe surface of the plasma-facing wall.
This work was supported by
T conc. high1)
2)
3)NIFS budget NIFS14KOBE031.
T conc. low
PSL intensity

Fig. Picture of a carbon deposited SS316L and IP imagesof the tritium contaminated SS316L. 1) : before
irradiated to D plasma, 2) : after irradiated to D plasma
for 30 min, 3) : after irradiated to D plasma for 90 min.

- 195 -

P2-22
Measurement of tritium sorbed in tungsten deposition layer
by imaging plate technique
K. Katayama, Y. Torikaia, M. Noguchi, N. Ashikawab, S. Fukada
Department of Advanced Energy Engineering Science, Kyushu University,
6-1, Kasugakoen, Kasuga-shi, Fukuoka 816-8580, Japana
Hydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japanb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Understanding of tritium (T) behavior in plasma-facing wall of a fusion reactor is an
important issue from viewpoints of T economy and safety. The plasma facing surface ischanged with plasma operation time by various mass transfer phenomena such as sputteringand deposition. The deposited material has different structure from the original material.
Therefore, the investigation of T behavior in not only original materials but also depositedmaterials is necessary. Tungsten (W) is a candidate material for plasma-facing components
because of low solubility and low sputtering yield for hydrogen isotopes. It has been observedthat W deposition layer formed by hydrogen plasma sputtering retains a large amount ofhydrogen during sputtering-deposition process [1]. However, basic behavior of hydrogenisotopes in W deposition layer is not understood significantly. In this study, W deposition layer
formed by hydrogen plasma sputtering was exposed to gaseous T and the reduction of T levelby heating under vacuum condition was investigated by the imaging plate (IP) technique.
Samples of W deposition layer were formed on W substrates by hydrogen RF plasma
sputtering. Prepared samples were separately put in a reaction tube of the T exposure apparatus
installed in University of Toyama, and heated at the preset temperature for 3 hours under thevacuum condition in order to remove adsorbed water on the sample surface. Then, the samples
were exposed to the T/D mixture (7.2% T/D) for 3 hours. After the T exposure process, theclosed reaction tube in which the samples were contained was transported into a glove box
filled with argon. Then, T level of the samples were investigated by IP without air exposure.
After that, the samples were put in the reaction tube again and heated under the vacuum
condition. IP measurement and heating process were performed repeatedly. For comparison, W
substrates without W deposition layer were exposed to T/D mixture at the same conditions.
As an example, IP image from the W deposition
layer and the W substrate exposed to T at 573K areshown in Fig.1. The initial intensities ofphoto-stimulated luminescence (PSL) were 1029
PSL/mm2/h for W deposition layer and 81
PSL/mm2/h for W substrate. T level on W deposition
layer was 10 times larger than that on W substrate. A
part of retained T was released at a low temperatureof 95 oC. These results suggest that the formation of
Fig.1 IP image of W deposition layer
W deposition layer increase T inventory in the vessel(left) and W substrate (right) exposed
and a certain amount of T can be removed at about
to gaseous T.
100 oC by in-vessel baking.
This work was performed with the support of the NIFS Collaboration Research program(NIFS14KOBF031).
[1] K.Katayama et al., Fusion Sci. Technol., 54 (2008) 549-552.

- 196 -

P2-23Tungsten-surfaces-structure Dependence of Sputtering Yield for Noble Gas
H. Nakamuraa, b, S. Saito c, A. M. Ito a and A. Takayama aaNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
Nagoya University, 332-6 Oroshi, Toki 509-5292, Japanc
National Institute of Technology, Kushiro College, 2-32-1 Otanoshike-Nishi , Kushiro, Hokkaido,
084-0916, Japanhnakamura@nifs.ac.jp

Tungsten nano structure, e.g. a bubble or a fuzz structure, has been studied recently by a lot ofplasma-wall interaction researchers [1, 2]. The purpose of these researches is to avoid thedecrease of the mechanical property and the heat conductivity of the original tungsten material.
To achieve this purpose, it is necessary to reveal the mechanism of the tungsten nano structureformation. We, recently, proposed the scenario of fuzz structure formation using computer
simulation [3] which is composed of the binary collision approximation (BCA) [4], density
functional theory (DFT) [5], molecular dynamics (MD), and Monte-Carlo (MC) calculations.
Especially our MD-MC hybrid code can reproduce the "fuzz"-like structure in the surface oftungsten bulk under the helium irradiation [4]. Thus, we demonstrated that our simulation code
is a powerful tool to calculate the interaction between the tungsten and the gases (hydrogenisotopes and noble gases). In this paper, we estimate the sputtering yield and the absorptivity of
the noble gas for the tungsten with the nano-structure surface using BCA simulation code AC∀T (atomic collision in any structured target) [4,6,7]. Though these physical quantities for thebulk tungsten are already well-known, those of the tungsten with the nano-structure surface are
not obtained yet. We, therefore, show the sputtering yield and the absorptivity of the hydrogenisotopes and the noble gases for the tungsten with the nano-structure surface. Moreover, we
introduce the tungsten structure dependence of those quantities.

[1] S. Takamura, N. Ohno, D. Nishijima and S. Kajita, Plasma Fusion Res. Vol. 1, 051 (2006).
[2] S. Kajita, T. Saeki, Y. Hirohata, M. Yajima, N. Ohno, R. Yoshihara and N. Yoshida: Jpn. J.
Appl. Phys. Vol. 50, p. 08JG01 (2011).
[3] A.M.Ito, et al., J. Nucl. Mat. in press. doi:10.1016/j.jnucmat.2015.01.018.
[4] S. Saito, A. Takayama, A.M. Ito, H. Nakamura, J. Nucl. Mater. Suppl., Vol. 438 p. S895
(2013).
[5] A. Takayama, A.M. Ito, S. Saito, N. Ohno, H. Nakamura, Jpn. J. Appl. Phys., Vol.52 p.
01AL03 (2013).
[6] A. Takayama, S. Saito, A.M. Ito, T. Kenmotsu and H. Nakamura: J. Appl. Phys. Vol. 50,
p.01AB03 (2011).
[7] S. Saito, A. Takayama, A.M. Ito and H. Nakamura: Proc. Int. Conf. Model. Simulation
Technol., p.197 (2011).

- 197 -

P2-24
Color measurements of the first wall toward the systematic evaluation of
the thickness distribution of deposition layer in LHD
G. Motojima, N. Yoshida1), S. Masuzaki, R. Sakamoto, M. Tokitani, H. Tanaka, T. Murase,
D. Nagata, K. Matsumoto2), M. Miyamoto3), M. Yajima, M. Sakamoto4), H. Yamada,T. Morisaki, and LHD Experiment Group
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan
1) Research Institute for Applied Mechanics, Kyushu Univ., 6-1, Kasuga, Fukuoka 812-8580, Japan2) Honda R & D Co. Ltd., 4630 Shimotakanezawa, Haga-machi, Tochigi 321-3393, Japan
3) Department of Material Science, Shimane University, Matsue, Shimane 690-8504, Japan
4) Plasma Research Center, University of Tsukuba, 1-1-1 Tennodai, Tsukuba, Ibaraki 305-8577, JapanAn innovative color measurement technique is employed in the Large Helical Device(LHD). This study provides a method for obtaining in broad spatial extent and in great detailthe color information of the first wall relating to the thickness of the deposition layer. The
RGB (Red, Green, and Blue) value, mainly of the stainless steel plates on the helically twisted
coil, is measured by the handy color analyzer.
In the LHD, long-term exposed sample analysis reveals that the deposition layer is
primarily composed of carbon formed on the first wall, and may be a contributing factor towall retention. The total amount of the deposition layer formed in the vacuum vessel
influences the wall retention rate. Thus, reliable evaluation of the total amount of thedeposition layer is crucial. The color depends on the composition, the structure, and thethickness of the deposition layer. In LHD, the composition of the deposition layer is primarily
carbon. Although the surface roughness on the deposition layer presumably affects thereflection coefficient at each RGB, the color is almost certainly related to the thickness of thedeposition layer assuming the same surface structure. In this study, a method for obtaining in
broad spatial extent and in great detail the color information of the first wall has been
provided using an innovative concept for the color measurement [1].
The first wall plates in the vacuum vessel of LHD are stainless steel (SUS316L), while
isotropic graphite plates are installed in the divertor section. The former is the main material
in LHD (700 m2), while the graphite area (30 m2) constitutes only about 5% of the total
plasma facing area. We measure the RGB mainly of the stainless steel plates on the helicallytwisted coil in one of 10 toroidal sections of the vacuum vessel. The number of measured
stainless steel plates totaled 530. On the outer torus side, the colors of almost all stainless steel
plates are close to black. On the inner torus side, all plates are close to white, which indicatesthe reflection coefficient of around 1, except for those neighboring the divertor plates. These
results suggest that the outer torus side is deposition-dominant, while the inner torus side isprimarily erosion-dominant. Characterization of the deposition layer by this study isqualitatively consistent with the previous results from the direct specimen analysis.
In this presentation, the thickness distribution of deposition layer will be shown fromthe analysis of samples amounted on the same toroidal section with color measurements aspart of the long-term exposed sample analysis, using a focused ion beam system and atransmission electron microscope (TEM) etc.
This work is performed with the support and under the auspices of the NIFS
Collaboration Research program (NIFSUMPP003-1).
[1] G. Motojima et al., Plasma Fusion Res. 10, 1202074 (2015).

- 198 -

P2-25
Linear stability analysis of Alfvén eigemodes with arbitrary
velocity distribution functions in tokamak plasmas
Taisei Futakuchi, Atsushi Fukuyama, Hideo Nuga
Graduate School of Engineering, Kyoto University
Kyoto-Daigaku-Katsura, Nishikyo-ku, Kyoto 615-8540, Japan

futakuchi@p-grp.nucleng.kyoto-u.ac.jp

The excitation of Alfvén eigenmodes (AE) driven by fast ions in tokamak plasmas has been
extensively studied theoretically and numerically, since it may cause the loss of fast ions andreduce the heating power to maintain the fusion reaction. Since the heating scheme andabsorbed power density strongly aﬀects the velocity distribution function of fast ions, thestability of AE has to be evaluated for arbitrary velocity distribution function. The full wave
code TASK/WM has been used to analyze the linear stability of Alfvén eigenmodes driven by
fast ions with isotoropic Maxwellian and slowing-down distributuions. Recently we have developed a module to numerically calculate drift-kinetic electric susceptibility of fast ions with
arbitrary radial and velocity distribution function. With this moudule, complex eigenmode
frequencies were obtained by solving Maxwell’s equation. Examples of eigenmode frequencies
in complex frequency plane are shown in Fig. 1. In the case of isotropic Maxwellian distribution function, TAE and lower-frequency eigenmodes exist and the groth rate of eigenmodes
increases with the increase of density or eﬀective temperature of fast ions. Parameter dependence of the eigenmode frequencies and wave structures for arbitrary distribution function
will be demonstrated.
a)

b)

0.03

0.004

0.02

Im(f) [MHz]Im(f) [MHz]

0.002

0.000

-0.0020.01

0.00-0.004
0.04

0.06

0.08

-0.01

0.10

Re(f) [MHz]

0.04

0.06

0.08

0.10

Re(f) [MHz]Fig. 1 Eigenode frequencies without fast ion components (a), and with fast ion components
nF 0 = 1 × 101 7m−3 , TF 0 = 0.5MeV (b)- 199 -

P2-26 Toki Conference
25th International
3-6 November 2015

Bifurcation of the interchange mode growth rate androtation frequency due to the perpendicular heat
conductivity in stellarator plasmas
T. Nicolas1 , K. Ichiguchi1,2 , M. Sato1 , Y. Todo1,2 , Y. Suzuki1,2 ,A. Ishizawa1,2 , and S. Sakakibara1,2
1

National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, JapanThe Graduate University of Advanced Study, Sokendai, Toki 509-5292, Japan2

The linear stability against interchange modes of heliotrons like the Large Helical Device has not been fully understood yet. The interchange mode is the dominant instability
in nearly current-free 3D configurations. Recently, experiments in the so-called inwardshifted configuration have shown the machine can be operated safely up to hβi ∼ 5%
without major MHD event [1]. Here hβi is the volume averaged ratio between kinetic andmagnetic pressure. The rotation of the mode, which seems to play a role in the stability [2],is also not understood. Namely, the modes are ob×10served to rotate in the electron direction [3], whereas
3
Bifurcation
extended MHD theory without heat conductivity
but including electron and ion diamagnetic effects
2
predicts rotation in the ion direction for Mercier
1
unstable modes and almost no rotation for Mercier
stable modes. [4].
0
In this contribution, we present a new effect on−1
Growth Ratethe linear Mercier unstable interchange growth rate
Frequency
χccaused by the perpendicular heat conductivity χ⊥ .
−2
10
10
When χ⊥ is equal to a critical value χc , the growthχ⊥
rates of the two most unstable modes become equal. Figure 1: Frequency and growth
For χ⊥ > χc , the two modes have same growth rate rate of the first two most unstableand opposite frequency, as shown in Fig. 1. This modes
degeneracy is removed by the diamagnetic effects.
In the domain χ⊥ > χc , the diamagnetic effects can be destabilizing and the rotation
can be in the electron direction. The rotation direction is determined by the value of
perpendicular viscosity.
This effect, which sheds a new light on the stability of the interchange mode in stellarators, is found analytically and confirmed both with a 3 field linear eigenvalue solver
and with 3D non-linear MHD simulations utilizing the MIPS code [5].
Growth rate, Frequency−2−5References[1] H. Yamada et al. Nucl. Fusion 51(9), 4021 (2011).
[2] Y. Takemura, et al. Nucl. Fusion 52(10), 102001 (2012).
[3] Y. Takemura, et al. Plasma and Fusion Research 8(0), 1402123 (2013).
[4] S. Gupta, et al. Phys. Plasmas 9(8), 3395 (2002).
[5] Y. Todo, et al. Plasma and Fusion Research 5, S2062 (2010).

Corresponding author: Timothee Nicolas timothee.nicolas@nifs.ac.jp- 200 -

−4

P2-27Comparison of bootstrap current calculation in helical plasmas
among different types of approximations in drift-kinetic
equation
B. Huanga , S. Satakea,b , R. Kannoa,b , S. Matsuokac
a Sokendai (The Graduate University for Advanced Studies),322-6 Oroshi-cho, Toki 509-5292,

Japan
b National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb Research Organization for Information Science and Technology, 6F Kimec-Center Build.,

1-5-2 Minatojima-minamimachi, Chuo-ku, Kobe, 650-0047 Japan

In toroidal plasma, neoclassical (NC) theory describes radial NC flux and the bootstrap
current. The bootstrap current depends on Magnetohydrodynamics (MHD) equilibrium
and collisionality. The bootstrap current is unessential for MHD equilibria of helicalplasma. However, in high-beta and high-temperature plasma, the bootstrap current could
be large enough to affect equilibrium field. Therefore, a self-consistent method is required
to track MHD equilibrium and bootstrap current. For example, Wendelstein 7-X (W7X) is optimized to have good MHD instabilities and low bootstrap current in order toimprove NC confinement.[1] Here W7-X design fellowed the local NC code.[2] Conventionally, mono-energy and local approximation are employed to reduce computation timeand resource. Here, ”local” means neglecting the higher order magnetic drift terms in thedrift-kinetic equation. The recent NC studies[3][4] indicate that conventional local approximations are deficient for some conditions in radial neoclassical transport. It showsthat keeping tangential magnetic drift terms leads to large differences in the radial NC
flux where E × B rotation is small. The tangential magnetic drift effect on bootstrap current is still an open question. This work performs the benchmark between conventionaland new local code with the tangential magnetic drift and discusses its contribution on
bootstrap current in LHD and W7-X.
References[1] J. Geiger, C. D. Beidler, Y. Feng, H. Maaberg, N. B. Marushchenko, and Y. Turkin, “Physics inthe magnetic configuration space of w7-x,” Plasma Physics and Controlled Fusion, vol. 57, no. 1,
p. 014004, 2015.
[2] D. I. van Rij and S. P. Hirshman, “Variational bounds for transport coefficients in three-dimensional
toroidal plasmas,” Phys Fluids B; (United States), vol. 1:3, Mar 1989.
[3] S. Matsuoka, S. Satake, R. Kanno, and H. Sugama, “Effects of magnetic drift tangential to magnetic
surfaces on neoclassical transport in non-axisymmetric plasmas,” Physics of Plasmas, vol. 22, no. 7,
p. 072511, 2015.
[4] M. Landreman, F. I. Parra, P. J. Catto, D. R. Ernst, and I. Pusztai, “Radially global f computationof neoclassical phenomena in a tokamak pedestal,” Plasma Physics and Controlled Fusion, vol. 56,
no. 4, p. 045005, 2014.

- 201 -

P2-28Chaos of nonlinear ion acoustic waves with vorticity
Yuji Ohno and Zensho Yoshida
Graduate School of Frontier Sciences, The University of Tokyo,
5-1-5 Kashiwanoha, Kashiwa, Chiba 277-8561, Japan

While there are a lot of works on the ion acoustic waves, especially on the solitons, they havenot focused on the vortex of the wave fields. For example, the Korteweg–de Vries (KdV)
equation considers a spatially one-dimensional system, so the fields do not have a freedom tomake a vorticity finite. The two- or three- dimensional generalization of the KdV equation is
the Kadomtsev–Petviashvili (KP) equation, which, however, is set to make the vorticity zero.
In this research, we show that the series of KP equations (including higher order series) for
ion acoustic waves cannot have vorticities. We also prove that entropy must be homogeneous andbaroclinic effects cannot produce any vorticities when we derive the equations. Thus, we cansay that the reductive perturbation method, which is used to derive the KP equation, eliminates
vortex.
This result motivates us to formulate and analyze a generalized system that has a finite
vorticity. We change the expansion of velocity in the reductive perturbation method and derivea new system of equations with a finite vorticity. It consists of the modified three-dimensional
KP equation


1
1 3
(1)
@x @ t  C @x  C @x  C Œ;  C ?  D 0
2
2

hkx i

and the two-dimensional Euler vorticity equation for ? .
Painlevé test is widely used to investigate whether a differential equation is integrable. The
test shows that KdV equation and 2D-KP equation are integrable, but 3D-KP equation is not
integrable [1]. Thus, the equation (1), which is a generalization of 3D-KP equation, is also
non-integrable.
We investigate chaotic property by solving the equation (1) with .y; ´/ D a sin.2y=L/
a cos.2´=L/ (a steady-state solution of thep Euler vorticity equation). As the initial condition
2of  we use a line-soliton 3 sech Œ.x 2y/= 2, the soliton solution of the 2D-KP equation.  ishomogeneous in ´-directon at t D 0, and bea = 0.3, κ = 2
a = 1.0, κ = 2
comes inhomogeneous at t > 0 because of .
a = 0.5, κ = 2
a = 1.0, κ = 4
a = 0.8, κ = 2
a = 1.0, κ = 8
When is weak  becomes twisted, and when
0.7
is strong  becomes divided and spread.
0.6
We evaluated these observation by a calcula0.5
tion of average wave numbers (Fig. 1). We
0.4
can see that larger makes hkx i larger, even
0.3
though is independent of x. hky i and hk´ i
0.2
0.1
also grow. The growth of hki means that large
0.0
scale structures are divided into small scale
0
20
40
60
80
100
structures. Thus, we can consider that this ret
sult shows the nonintegrability due to vorticity
Figure 1: Average wave number hkx i.
in the new equation (1).
[1] H.-y. Ruan, S.-y. Lou, and Y.-x. Chen, J. Phys. A: Math. Gen. 32 (1999) 2719.

- 202 -

P2-29
Electron impact ionization cross sections of the fusion plasmahydrocarbon gas: methane
G. Purohit, P. Singh and V. Patidar
Department of Physics, Sir Padampat Singhania University, Bhatewar, Udaipur-313601,
India
e-mail: ghanshyam.purohit@spsu.ac.in

The study of electron impact ionization of atoms, ions and molecules has been of interest
since the early days of quantum physics. Particularly, study of electron impact ionization
cross section of various targets is important as it gives information about the delicate
collision dynamics and wave functions [1]. Electron-impact ionization cross sections ofatoms and molecules are important not only for the development of basic collision theory,
but also in a wide range of practical applications. For example, the modeling ofsemiconductor manufacturing by plasma processing requires ionization cross sections offluorocarbons used as the feed gas, their ions and fragments. Hydrocarbon molecules,
such as methane, are found in the vicinity of divertors in a fusion device, making it
necessary to include ionization of such molecules to model low-temperature plasmas intokamaks. Ionization cross sections near the ionization threshold are also used inmodeling combustion of flammable gases. Thus, needs for cross section data have been
identified in the fields of plasma physics such as astrophysics and fusion plasmas, as wellas in the fields of applications: gas discharge lasers, semiconductor manufacturing
processes and plasma diagnostics.
We have calculated differential cross sections for the various targets including xenon
atoms and water molecule [2, 3]. We report in this communication the electron impact
ionization cross section data for methane which has importance in the fusion plasma
related processes [4]. We use semi-classical approach to calculate the ionization crosssection of methane. We compare our results with the available experimental andtheoretical results.
References
[1]
[2]
[3]
[4]

H. Ehrhardt, M. Schulz, T. Tekaat, K. Willmann, Phys. Rev. Lett. 22 (1969) 89.
P. Singh, G. Purohit, V. Patidar, J. Phys. B: At. Mol. and Opt. Phys. 46 (2013)
115207.
P. Singh, G. Purohit, C. Champion, V. Patidar, Phys. Rev. A 89 (2014) 032714.
W. L. Morgan, Plasma Chem. Plasma Process.12 (1992) 477.

- 203 -

1

P2-30Kinetic dynamics simulation of the detached plasma
T. Pianpanita , S. Ishiguroa,b , H. Hasegawaa,ba Department of Fusion Science, SOKENDAI, 322-6 Oroshi-cho, Toki 509-5292, Japanb National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan

theerasarn.pianpanit@nifs.ac.jp

The detached plasma has been proposed to reduce the heat flux to the divertor. Fluid
code has been widely used to investigate the detached plasma [1] but the cooling of plasma,trapped particle effects, and other kinetic dynamics in the detached plasma has not been
well understood. Particle-in-Cell (PIC) simulation with the Monte Carlo collisions [2] andthe cumulative scattering angle coulomb collision [3] are carried out to study dynamical
kinetic behavior of the plasma. The constant pressure and constant temperature of neutral
gas box in front of the divertor target model has been used in the simulation. The simulation
has been carried out by using the real mass ratio and the self-consistent sheath effect canbe observed. The low plasma density and the small system size compare to the real fusion
devices have been used, thus by modifying the Coulomb collision frequency, νc′ = 100νc , theresults show the strongly decrease in electron temperature inside the neutral gas box. The
electron temperature in front of the divertor target also becomes less than 1 eV, which is oneof the conditions for detached plasma.

[1] S. Nakazawa, et al., Plasma Phys. Control. Fusion 43 (2000) 401[2] V. Vahedi and M. Surendra, Comput. Phys. Commun. 87 (1995) 179[3] K. Nanbu, Phys. Rev. E 55 (1997) 4642

- 204 -

P2-31
Local approximation of the drift kinetic equation
H. Sugama,1,2 S. Matsuoka,3 S. Satake,1,2 and R. Kanno1,2
1 National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
2 SOKENDAI (The Graduate University for Advanced Studies), Toki 509-5292, Japan
3 Research Organization for Information Science and Technology, Kobe 650-0047, Japan

Conventional calculations of neoclassical transport fluxes are done applying radially local
approximation to solving the drift kinetic equation, in which vd · ∇f1 are often neglected asa small term of higher order in the normalized gyroradius parameter δ ∼ ρ/L. (Here, vd ,f1 , ρ, and L represent the guiding center drift velocity, the deviation of the guiding center
distribution function from the local Maxwellian equilibrium distribution, the gyroradius, andthe equilibrium scale length, respectively.) However, in stellarator and heliotron plasmas,
this vd · ∇f1 term is known to be influential on the resultant neoclassical transport because
it significantly change orbits of particles trapped in helical ripples. Therefore, at least, theE × B drift part vE · ∇f1 in vd · ∇f1 has been kept in most studies of neoclassical transport inhelical systems. Recently, it was shown by Matsuoka et al.[1] that the neoclassical transport issignificantly influenced by retaining the magnetic drift tangential to flux surfaces in vd ·∇f1 forthe magnetic configuration of LHD especially when the radial electric field is weak. However,as pointed by Landreman et al.[2], stationary solutions of the drift kinetic equation with theradially local approximation used require additional artificial sources (or sinks) of particles
and energy when the above-mentioned drift terms are retained. In this work, a novel radiallylocal approximation of the drift kinetic equation is presented. The new drift kinetic equationthat includes both E × B and tangential magnetic drift terms is written in the conservative
form and it has favorable properties for numerical simulation that any additional terms forparticle and energy sources are unnecessary for obtaining stationary solutions. Besides, it isshown that the intrinsic ambipolarity condition for neoclassical particle fluxes are satisfied
in the presence of quasisymmetry of the magnetic field strength.

[1] S. Matsuoka, S. Satake, R. Kanno, and H. Sugama, Phys. Plasmas 22, 072511 (2015).
[2] M. Landreman, et al., Phys. Plasmas 21, 042503 (2014).

- 205 -

P2-32Modification of neutron emission spectra and its effect on determination offuel ion ratio nd/nt in beam injected DT plasma
Y. Kawamoto, H. Matsuura
Applied Quantum Physics and Nuclear Engineering, Kyushu University,
744 Motooka, Fukuoka 819-0395, Japan

To control the DT fusion reactor, it is important to accurately obtain the fuel ion ratio nd/nt. In
DT plasma, there are neutrons which have 14.1 MeV peak caused by DT reaction and have 2.45
MeV peak caused by DD reaction. The method of obtaining the fuel ion ratio nd/nt by measuring
peaks of these neutron spectra has been proposed [1]. This method, so far, has been valued for
ITER under certain conditions and it has been shown that it is possible to obtain the fuel ion
ratio nd/nt in reactor core region by the method with sufficient accuracy [2]. However the
scattered neutrons caused by DT reaction are detected as noise for DD signals. Hence, use of
Neutron Beam Injection (NBI) for neutron spectrometry is considered in order to improvemeasurement accuracy of DD signals, because DD reactivity is enhanced by NBI. NBI
influences not only reactivity but also neutron emission spectra from gauss distribution [3]. Insuch a case, ratio of DD/DT neutron peak also changes. Therefore it needs to correlate neutron
signals measured with the fuel ion ratio nd/nt when beam-injection is made. The purpose of thispaper is to clarify a correlation between neutron signals measured and the fuel ion ratio nd/nt by
evaluating the influence of beam-injection. In addition, we evaluated a noise to examine
whether the above method can be used for plasma diagnostics in blanket structure such as ITER,
Water Cooled Solid Breeder (WCSB), and LiPb blanket.
We assumed the DT plasma with deuteron beam-injection. Deuteron and triton distributionfunctions are evaluated on the basis of the Boltzmann-Fokker-Planck model and neutron
emission spectra are obtained by using the distribution functions. Using the neutron emission
spectra obtained, neutron flux in reactor core region is calculated with the continuous-energyMonte Carlo transport code MVP [4]. It is shown that reassessing of correlation between
neutron signals measured and the fuel ion ratio nd/nt according to plasma conditions is necessary
when beam is injected, and NBI is useful for increasing Signal/Noise (S/N).

[1] K. Källne et al., Rev. Sci. Instrum 62 (1991) 2871.
[2] F. Moro, et al., Fusion Eng. Des. 86 (2011) 1277.
[3] H. Matsuura, et al., Plasma Fusion Res. 8 (2013) 2403064.
[4] Y. Nagaya, et al., JAERI (2005) 1348

- 206 -

P2-33A verification scenario of knock-on tail formation due to nuclear elastic
scattering using γ-ray-generating 6Li+D reaction
in deuterium plasmas
H. Matsuura, S. Sugiyama, Y. Nishimura, Y. Kawamoto
Applied Quantum Physics and Nuclear Engineering, Kyushu University,
744 Motooka, Fukuoka 819-0395, Japan
mat@nucl.kyushu-u.ac.jp

It is well known that in a thermonuclear plasma energetic ions create knock-on tails [1] in fuel-ion
velocity distribution functions due to nuclear elastic scattering (NES) [2]. In a thermonuclear plasma
3.52-MeV α-particles are continuously produced and use of beam injection with energy more than 1
MeV is considered. In such a case, NES effects on slowing down of the energetic ions may not be
negligible compared with those due to Coulomb collisions. We have evaluated the NES effects onburning plasma properties in beam-injected plasmas [3,4] on the basis of the Boltzmann-Fokker-Planck(BFP) model. It is important to validate the analysis model and ascertain the NES effects previously
evaluated via the experimental data.
In this paper, a verification scenario of knock-on tail formation in deuteron velocity distribution function
by energetic protons using γ-ray-generating 6Li+D reaction is examined for proton-beam-injected
deuterium plasmas. Previously we roughly evaluated the 6Li+D reaction rate coefficient assuming auniform plasma [5]. In this paper we re-evaluate the rate coefficient using more precise model supposing
the future experiment. We take into account behaviours of beam-injected protons in a magnetic
confinement device. Using volume-averaged 2D velocity distribution function of beam-injected protons,
which is obtained by means of the ion trajectory analysis, knock-on tail formation in deuteron velocity
distribution function due to NES is examined in 2D velocity space using 2D Boltzmann collision term.
From the 2D deuteron velocity distribution function obtained enhancement of the emission rate of
0.48-MeV γ-rays by 6Li(d,p)7Li*, 7Li*→7Li+γ reaction [6,7] due to NES is evaluated. To grasp the
potential diagnostic capabilities, ratio of 0.48-MeVγ-rays to background noise component is also
evaluated on the basis of Monte-Carlo simulation for neutron transport equation. A possible scenario toverify the BFP numerical model and NES effects including knock-on tail formation is studied on thebasis of a newly-developed model, and discussion is made for the implementability of the scenario.
[1] L. Ballabio, et al., Phys. Rev. E, 55 (1997) 3358.
[2] J. J. Devaney and M.L.Stein, Nucl. Sci. Eng. 46 (1971) 323.
[3] H. Matsuura, Y. Nakao, Phys. Plasmas, 13 (2006) 062507.
[4] H. Matsuura, Y. Nakao, Phys. Plasmas, 14 (2007) 054504.
[5] H. Matsuura, et al., Plasma Fusion Research, 2 (2007) S1078.
[6] V. T. Voronchev, et al., Phys. Rev. E, 63 (2001) 26413-1.
[7] V. T. Voronchev, et al., Mem, Fac. Eng. Kyushu Univ., 51 (1991) 63.

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P2-34
Transmission Mode Analysis in Polarizer Miter Bend
Y. Fujitaa , S. Ikunob , S. Saitoc , T. I. Tsujimurad , S. Kuboa,d , and H. Nakamuraa,da Nagoya University, 322-6 Oroshi-cho, Toki 509-5292, Japanb Tokyo University of Technology, 1404-1 Katakura, Hachioji 192-0982, Japanc Kushiro National College of Technology, 2-31-1 Otanoshike Nishi, Kushiro 084-0916, Japand National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan

In the electron cyclotron resonance heating (ECRH), the corrugated waveguide is employed
to reduce the deterioration of the transmission eﬃciency due to induced current density. The
miter bend (MB) is also employed in ECRH for changing the wave propagation direction.
Moreover, the millimeter waves injected into the plasma should be polarized to accurately
resonance with electrons in the plasma. The polarizer miter bend (PMB) is implemented bycarving grooves in MB mirror. The polarization can be controlled by rotating the reflectorof PMB [1, 2]. The structure of the PMB is very complex because of grooves on the surface. In addition, other modes than the input mode are generated due to the reflection at
MB/PMB reflector [3, 4]. However, the electromagnetic wave propagation phenomena in the
MB/PMB have not been suﬃciently verified [5, 6]. It is necessary to investigate the influenceof the groove shape of PMB reflector on the polarization. Besides, the ratio of other modes
that generated by MB/PMB to the transmission mode should be confirmed for the eﬃcientheating.
The purpose of the present study is to develop the numerical code for analyzing the electromagnetic wave propagation phenomena using the finite-diﬀerence time-domain (FDTD)
method [7, 8, 9]. Furthermore, the existing modes that generated by MB/PMB are numerically investigated by using the results [3, 10].

[1] J. L. Doane, Int. J. Infrared Millimeter Waves 13 (1992) 1727
[2] K. Nagasaki et al., Rev. Sci. Instrum. 66 (1995) 3432
[3] K. Ohkubo et al., Fusion Sci. Technol. 62 (2012) 389[4] Y. Fujita et al., Proc. of the Compumag 2015, Montreal, Canada[5] T. Ii et al., Rev. Sci. Instrum. 86 (2015) 023502[6] B. Plaum et al., J. Infrared Milli. Terahz. Waves 32 (2010) 482[7] Y. Tamura et al., J. Phys.: Conf. Ser., 410 012029[8] H. Nakamura et al., J. Phys.: Conf. Ser., 410 012046[9] H. Nakamura et al., ITC24 (poster), Gifu, Japan, Nov. 4-7[10] J. L. Doane, in Infrared and millimeter waves (Academic Press, Orlando, 1985) Vol. 13,
Chap. 5, p. 123.

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P2-35
Kinetic full wave analysis of electron cyclotron wave mode
conversion in tokamak plasmas
S. A. Khan1 , A. Fukuyama1 , H. Igami2 , H. Idei3
1 Department of Nuclear Engineering, Kyoto Univerisity, Kyoto, Japan
2 National Institute for Fusion Science, Toki, Japan
3 Research Institute for Applied Mathematics, Kyushu University, Kasuga, Japan

shabbir.khan.36m@st.kyoto-u.ac.jp

For heating and current drive in high-density core plasmas of tokamaks, electromagnetic
waves with electron cyclotron (EC) range of frequencies have been extensively studied theoretically and experimentally. The propagation and absorption of EC waves are usually
analyzed by the ray tracing method based on geometrical optics for waves with short wave
length. In a plasma with high density or low magnetic field, however, the presence of cutoff
layer may prevent the waves from penetrating into the central part from the low field side. In
this case, the full wave analysis of EC waves is required for evaluating the absorption profileand optimizing the wave launching conditions. In the present analysis, two schemes of thefull wave analysis in which the boundary value problem of Maxwell’s equation is solved arepresented. The first one is a one-dimensional kinetic analysis of the X-O-B mode conversion
using the TASK/W1D code. The finite gyroradius effects are represented by an integral formof the dielectric tensor derived by integrating along particle orbits and introducing variabletransformation from velocities to spatial coordinates. The mode conversion to the electron
Bernstein waves and the absorption near the cyclotron harmonic resonance are successfully
described. Parameter dependence of the absorption rate and the deposition profile is discussed. The second one is a two-dimensional analysis near the cutoff-resonance layer using
the TASK/WF2D code. Using the integral formulation of the dielectric tensor, the mode
conversion to the electron Bernstein wave near the upper hybrid layer is demonstrated inaddition to the reflection and tunneling near the cutoff-resonance layer. Comparison withthe analysis using a cold plasma model will be presented.

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P2-36
Benchmark of electromagnetic gyrokinetic codes in high performance
fusion plasma
S. Maeyama, T.-H. Watanabe, H. Doerka, M. Nakatab, A. IshizawabaNagoya University, Furo-cho, Nagoya 464-8602, Japan
Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching, Germanyb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Since the high plasma β value is desirable to make a fusion reactor efficiently andeconomically, JT-60SA is being developed to demonstrate and study high-β discharge for
DEMO reactor. At the same time, to examine the finite β effects and improve the achievable β
value, high-β experiments have been conducted in various fusion devices.
Electromagnetic gyrokinetics simulation is regarded as a fundamental tool to investigate
the finite β effects on micro-turbulence. As the β value increase, typically, electrostatic ion
temperature gradient modes (ITGs) are stabilized, while electromagnetic modes aredestabilized such as kinetic ballooning modes (KBMs) and micro-tearing modes (MTMs).
The dominant instability and their nonlinear dynamics depend not only on the β value or
density/temperature gradients but also on the equilibrium magnetic configuration andCoulomb collision. Thus, in order to understand the experimentally-relevant turbulence,comprehensive gyrokinetic simulations including the realistic magnetic geometry and themultiple-species collision operator are required. For this purpose, a verification study basedon a code-code benchmark is helpful.
We have carried out a benchmark between the electromagnetic gyrokinetic codes GKV[1] and GENE [2]. By using the ASDEX Upgrade H-mode discharge data [3], both codes areused to analyze micro-instabilities in a local flux-tube model. It is confirmed that constructed
field-aligned coordinates and dispersion relation show good agreements. MTMs and ITGs
coexist in the examined parameter, where MTM is significantly driven by the collisionality.
In the presentation, more detailed and extended comparison will be reported.
[1] T.-H. Watanabe, and H. Sugama, Nucl. Fusion 46, 24 (2006).
[2] F. Jenko, W. Dorland, M. Kotschenreuther, and B. N. Rogers, Phys. Plasmas 7, 1904
(2000).
[3] H. Doerk, M. Dunne, F. Jenko, F. Ryter, P. A. Schneider, E. Wolfrum, and The ASDEX
Upgrade Team, Phys. Plasmas 22, 042503 (2015).

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P2-37
Moment extracted formulation for describing kinetic Alfven waves
T.-H. Watanabe, S. Maeyama, and Y. Watanabe
Department of Physics, Nagoya University, Nagoya 464-8602

Electromagnetic gyrokinetic equations have been widely used for describing drift waves,kinetic ballooning modes, and kinetic Alfven waves in fusion, space, and astrophysicalplasmas, where kinetic models of ions and electrons are dealt with. In describing theelectromagnetic perturbations with frequencies much lower than the ion gyrofrequency,
however, separation of space and time scales for the two species leads to numericaldifficulties (such as a sever restriction on the time step size and a demand of high phase-spaceresolution). In order to develop an efficient numerical scheme for the electromagnetic
gyrokinetic equations, we discuss a formulation of the kinetic Alfven waves convenient tonumerical simulations.
In the gyrokinetic simulations with ions and electrons, the fast parallel motion ofelectrons often restrict the time-step size. The semi-Lagrangian (SL) method may be useful torelax the numerical restriction [1, 2]. However, in case with electromagnetic perturbations,characteristics of the shear Alfven waves need to be decoupled from the parallel electron
motion. Otherwise, the SL method may suffer from numerical instability.
Before application of the SL scheme to the electromagnetic gyrokinetic equations, weconsider a simple drift kinetic system of electrons with the ion polarization effect. This is theminimal set for describing the kinetic Alfven waves. Then, we rewrite the governing
equations so that the low-order moment equations are extracted from the drift kinetic
equations. The electron distribution function is modified to satisfy constraints given by the
low-order moments. Numerical properties in regard to the new formulation will also be
discussed at the conference.
[1] S. Maeyama, et al., Plasma Fusion Res. 6, 2401028 (2011).
[2] S. Maeyama, et al., Comput. Phys. Comm. 183, 1986-1992 (2012).

- 211 -

P2-38Equilibrium reconstruction of plasma discharges in the Aditya tokamak
Deepti Sharma, Santanu Banerjee, A. K. Singh, R. Srinivasan, D. Raju, R. L. Tanna, J.
Ghosh, Y. Shankara Joisa, P. K. Atrey, S. K. Pathak and the Aditya teamInstitute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat, India
External magnetic measurements with flux loops and magnetic pick-up coils intokamaks have provided vital information on the shape of the plasma column and also global
current profile parameters, such as the sum of the poloidal beta (βp) and the internal
inductance (ℓi) [1]. Such a reconstruction needs to be fast and sufficiently accurate such that it
can be used routinely as a complementary input with other experimentally measured
parameters for any sort of physics analysis of the plasma discharges.
Here we present a method which can be used to proficiently reconstruct the current
profile parameters, the plasma shapes, and a current density profile satisfying the MHD
equilibrium constraint, reasonably conserving the external magnetic measurements. A GradShafranov (GS) equation solver, named as IPREQ [2], has been developed in IPR to search
for the best-fit current density profile. GS equation is a nonlinear elliptical differential
equation describing axisymmetric toroidal equilibria. Ohmic transformer current (OT),
vertical field coil current (BV) along with the plasma pressure (p) and current (Ip) profiles areused as inputs to the IPREQ code to reconstruct the equilibrium and the poloidal flux,
plasma shape, βp and the safety factor (q) are inferred.
At the four corners of the square cross-section vacuum vessel of Aditya, there are four
magnetic pick-up coils aligned to measure the poloidal magnetic field (Bθ) during a plasma
discharge. Further, there are two toroidal flux loops at the shadow of the limiter on the highfield side to measure the loop voltage inside the vacuum vessel. Vacuum shots with OT and
BV and no fill gas are used to calibrate these coils and loops. Measurement from these coils
and flux loops are used to reconstruct the equilibrium consistently with the peak density andtemperature measurements. Finally, the reconstructed equilibria are validated against thevisible images from the fast visible imaging diagnostic on Aditya.
[1]. L. L. Lao et al., “Reconstruction of current profile parameters and plasma shapes intokamaks,” Nucl. Fusion, 25, 1611 (1985).
[2]. Tokamak equilibrium code-IPREQ, R. Srinivasan and S. P. Deshpande IPR/RR393/2007 (August, 2007)- 212 -

P2-39Two dimensional hybrid simulation of a translated field-reversed
configuration
M. Yanagi, D. Adachi and T. Takahashi
Gunma University, 1-5-1 Tenjin-cho, Kiryu 376-8515, JapanThe field-reversed configuration (FRC) is a compact torus with extremely high , and its
magnetic field structure is simply connected. This feature lets a FRC move in an axial
direction; the motion is called translation. Translation provides FRC experiments withadvantages. One of them is that the formation region and the confinement region can beseparated by using translation technology, which is still used by recent experiments or
projects like C-2 [2] and FRC-HX [3].
However, there are many unclear phenomena observed in FRC translation experiments. So
we have developed a simulation code for the FRC translation research. The simulation uses
2D hybrid simulation method and we modeled the temporal change of external field in orderto reproduce the axial acceleration control in FRC translation experiments. In our presentation,
we will present our simulation results for several types of external field time-variation.
[1] M. Tuszewski, Nucl. Fusion 28 (1988) 2033.
[2] H. Y. Guo et al., Phys. Plasmas 18 (2011) 056110.
[3] C. Grabowski et al., IEEE Trans. Plasma Sci. 42 (2014) 1179.

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P2-40Influences of Hall effect on MHD behavior of two axially colliding
FRCs
K. Matsuzaki, S. Koike, M. Yanagi, N. Mizuguchi1, and T. Takahashi
Division of Electronics and Informatics, Gunma University, Kiryu, Gunma 376-8515, Japan
1

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

As one of the effective ways to improve lifetime of Field-Reversed Configurations(FRCs), there is a merging formation technique that is introduced in the C-2 experiment.
By colliding two accelerated FRCs, the merged FRC plasma exhibited a better
confinement performance [1, 2]. In the report from merging experiments, the structure
and profiles of the merged FRC plasma after completion of the merging process were
presented intensively. However, detailed study for a transient process which involves
magnetic reconnection phenomena is needed to improve the refueling efficiency for
application to a nuclear reactor design.
We report here the simulation results of colliding two accelerated FRCs. In our
simulation code, we modeled to separate the external magnetic field and the field
generated by plasma. By using time-variable external magnetic field, we reproduced thetranslation of FRCs. Meanwhile, the Hall effect is to play an important role in themagnetic reconnection research. For this reason, we employed both the 2-D resistive
MHD model and the 2-D resistive Hall-MHD model. We compare the simulation results
and report on the changes in the property of the plasma. Detailed results will be reported
in our presentation.
[1] M. W. Binderbauer et al., Phys. Rev. Lett. 105 (2010) 045003.
[2] H. Y. Guo, M. W. Binderbauer, D. Barnes et al., Phys. Plasmas 18 (2011) 056110.

- 214 -

P2-41

Flux supply by Neutral Beam Injection into Field-Reversed
Configuration plasma with resistive decay
R. Sekiguchi, S. Koike, and T. Takahashi
Gunma University, 1-5-1 Tenjin-cho, Kiryu 376-8515, Japan

Sustainment of Field-Reversed Configuration (FRC) plasma by NBI is most effective,
because the motion of injected beam ions is compatible with kinetic characteristics of
FRC [1]. Although several numerical works showed current drive [2] and power
deposition [3] by NBI, target FRC plasmas were supposed to be in a stationary or
equilibrium state. In the present study, we consider NBI into an FRC plasma withresistive decay; it is reproduced by 2-dimensional resistive MHD simulation. Neutral
Beam (NB) particles are injected tangentially to the toroidal plasma current, and their
ionization processes are reproduced by a Monte-Carlo method. From theparticle-tracking calculation of beam ions, we obtain the density and current for beamions. Since the magnetic field generated by the beam current affects the MHD behavior
of FRC plasma, the MHD calculation and particle-tracking calculation are done
alternately.
In our presentation, we will show the shine-thorough ratio of NB particles,
augmentation of the poloidal flux, and the orbit loss ratio of beam ions.
[1] T. Asai, Y. Suzuki, T. Yoneda et al., Phys. Plasmas 7 (2000) 2294.
[2] A. F. Lifschitz, R. Farengo, and N. R. Arista, Nucl. Fusion 42 (2002) 863.
[3] T. Takahashi, Y. Kondoh, Y. Hirano et al., J. Plasma Fusion Res. 82 (2006) 775.

- 215 -

P2-42Estimations of Beam-Beam Fusion Reaction Rates
in the Deuterium Experiment Plasma on LHD
M. Homma, S. Murakami and H. Yamaguchi
Department of Nuclear Engineering, Kyoto University, Kyoto 615-8540, Japan

Email: homma@p-grp.nucleng.kyoto-u.ac.jp

Deuterium plasma experiment campaign from 2017 is planned in the Large Helical Device(LHD). During deuterium discharges, 1 MeV tritons are produced by D-D fusion reactions,and then 14 MeV neutrons are generated by triton burn-up. The D-D fusion reaction rates
in the LHD deuterium plasma have been calculated by the FIT3D-DD code[1], which is anextension version of FIT3D. In the FIT3D-DD code, we have evaluated the fraction of tritons
which are produced only through the reactions between deuterium beams from the Neutral
Beam Injection (NBI) heating and deuterium thermal ions (beam-thermal reactions). In
the deuterium experiment plasma, however, beam-beam reactions in which both reacting
ions are injected NBI deuterons would also occur. Therefore, it is necessary to estimate thecontribution of the beam-beam fusion reactions for more precise prediction of triton burn-up
ratios and neutron production rates.
In this study, we estimate beam-beam fusion reaction rates between the NBI deuteron beams
in the deuterium experiment plasma of LHD. We obtain the deuterium beam distributionfunctions by the GNET (Global NEoclassical Transport) code[2,3], in which the drift kinetic
equation of energetic particles is solved in five-dimensional phase space, and the beam-beamfusion reaction rates are calculated. The radial profiles of triton and neutron production
rates are evaluated and are compared with that of the beam-thermal reaction rates. We also
investigate their dependences on plasma parameters.

[1] M. Homma et al., Plasma Fusion Res. 10, 3403050 (2015)[2] S. Murakami et al., Nucl. Fusion 40, 693 (2000)[3] Y. Masaoka et al., Nucl. Fusion 53, 093030 (2013)- 216 -

P2-43
Simulations of Energetic Particle Driven Geodesic Acoustic Modein 3-dimensional LHD Equilibrium
H. Wanga , Y. Todoa,b , and Y. Suzukia,ba National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanb The Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki 509-5292, Japanwanghao@nifs.ac.jp

The first simulation results of energetic particle driven geodesic acoustic mode (EGAM) in
3-dimensional LHD equilibrium are presented. MEGA, which is a hybrid code for energetic
particles interacting with a magnetohydrodynamic fluid, is used for the simulations. The
equilibrium data is based on LHD shot #109031 and generated by HINT2 code. The simulation parameters are also based on the same shot. The energetic particle velocity distribution
is slowing-down type with charge exchange, and pitch angle distribution is Gaussian type.
The simulated EGAM frequency is 61kHz, which is the same as the experimental observation[1]. The mode time evolution is shown in Fig. 1(a). The mode amplitude vθ increases
with γ/ωEGAM = 19%. At t = 0.15ms, the mode amplitude reaches the maximum value
and starts to be saturated. The mode is located near the magnetic axis, and the poloidal
mode number is 0 for vθ , as shown in Fig. 1(b). The velocity is high in the low field side andlow in the high field side. In the linear growth phase, the mode propagates outward, but inthe saturated phase, the propagation direction changes with time. In addition, the poloidal
mode number is 1 for density and pressure, and 2 for perturbed magnetic field strength.
The simulated mode numbers are consistent with theoretical prediction and experimental
observation.
0.002
0.0015(a)ve / vA

0.001
0.00050-0.0005-0.001-0.0015-0.0020

0.05

0.1 0.15
Time [ms]

0.2

0.25

Figure 1: (a) Time evolution of mode amplitude vθ . (b) Velocity in poloidal cross section at
t = 0.15ms.

[1] M. Osakabe, T. Ido, K. Ogawa et al, 25th IAEA Fusion Energy Conference, St. Petersburg, Russia.

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P2-44
Advances in applying the gyrokinetic heat diﬀusivity model totransport simulations for helical plasmas
S. Toda1 , M. Nunami1 , A. Ishizawa1 , S. Murakami2 , T. -H. Watanabe3 and H. Sugama1
1 National Institute for Fusion Science, Oroshi-cho 322-6, Toki, Gifu, 509-5292, Japan
2 Department of Nuclear Engineering, Kyoto University, Nishikyo, Kyoto, 615-8540, Japan
3 Department of Physics, Nagoya University, Furo-cho, Nagoya, Aichi, 464-8602, JapanA reduced transport model for the turbulent ion heat diﬀusivity due to the ion temperature gradient(ITG) mode was obtained from the gyrokinetic simulation using the GKV-X code for the high-Ti
Large Helical Device (LHD) plasmas [1]. This model is given by the function of the normalized
zonal flow decay time τ̃ZF and the mixing length estimate γ̃k̃y /k̃y2 integrated over the poloidal
wavenumber space, where γ̃k̃y is the normalized linear growth rate of the ITG mode. The nonlinear
gyrokinetic transport simulation results are quantitatively reproduced by the reduced model, which
we call the gyrokinetic heat diﬀusivity model. However, it is costly to carry out linear calculationsof the growth rate by the gyrokinetic simulation at each time step of the transport simulation.
How to apply the turbulent heat diﬀusivity model derived from the gyrokinetic simulations to
the dynamic transport simulation with a low computational cost was shown [2]. This additionalmodeling for the turbulent ion heat diﬀusivity has been applied to the integrated code, TASK3D.
The TASK3D simulation of the high Ti plasmas with the carbon pellet injection in LHD was
performed [3] using the gyro-Bohm+∇Ti model for the turbulent ion transport. In this study,
the method to apply the reduced model to the transport simulation is used for the LHD shots
diﬀerent from that in [2]. Dependence of the density scale length on the turbulent heat diﬀusivity
is examined in a case of the pure hydrogen plasma. We examine the ion temperature profiles before
and after the C pellet injection for applying the gyrokinetic heat diﬀusivity model to TASK3D.
The neoclassical transport is numerically estimated in the TASK3D simulation. Introduction of
the carbon impurity density remains to be improved in the turbulent transport studies of the LHD
high Ti plasmas. The simulation results without the eﬀects of the carbon density on the turbulent
heat diﬀusivity are compared with the experimental results. The eﬀect of the multi-ion species
plasma on the heat diﬀusivity after the C pellet injection is investigated.
[1] M. Nunami, et al., 2013 Phys. Plasmas 20 092307[2] S. Toda, et al., 2014 J. Phys.: Conf. Ser. 561 012020[3] S. Murakami, et al., 2015 Plasma Phys. Control. Fusion 57 054009- 218 -

P2-45Development of TASK3D/WM module for evaluation the ICRF wave
propagation and absorption in LHD
R. Seki, A. Fukuyamaa), T. Seki, K. Saito, H. Kasahara, S. Kamio,
M. Yokoyama, K. Y. Watanabe and T. Mutoh
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japana
Graduate School of Engineering Kyoto University, Kyoto 606-8501, Japan

In the ICRF (ion cyclotron range of frequencies) in the LHD, the maximum value of
input power reached about 4 MW and the helium plasma of the relatively high density with acenter electron density ~ 6x1019 m-3 can be maintained by using the hydrogen minority heating.
To investigate the transport characteristic or to optimize the heating condition of ICRF, it isimportant to evaluate the heating power profile of ICRF properly. In particular, since the
ray-tracing cannot be used because of long wave length of ICRF wave, to calculate the wave
propagation from the Maxwell’s equation is required.
The TASK3D/WM [1] has been developed as the wave propagation and absorption
evaluation code of ICRF in the non-axially symmetric 3-dimensional magnetic configurationsuch as LHD. In the TASK3D/WM, the Fourier expansion is used in the poloidal and toroidal
directions, and the Maxwell’s equation where finite difference method is used in the minor
radius is solved for electric field as a boundary value problem. Each electric field components(Erho, Etheta, Ephi) are expressed in the torus magnetic coordinates.
In calculations of the wave absorption of ICRF heating using the TASK3D/WM in
LHD, the cold plasma approximation is assumed in the dielectric constant tensor [2]. In thecold plasma approximation, since the thermal motion of plasma is not included, the wave
propagation can be calculated only in the region where a phase velocity of the wave is muchlarger than a thermal velocity of plasma, i.e. the region except for near the resonance layer. The
plasma heating from the wave near resonance layer cannot be evaluated properly. Therefore, in
the TASK3D/WM, the wave collisional dumping model is introduced so that the heating powerfrom the wave to plasma near the resonance layer can be evaluated. In this case, only thelocations where particles resonate with the wave can be calculated but the evaluation of heating
power is not properly. In addition, the wave propagation near the resonance layer can becalculated properly because the cold plasma approximate is assumed. Thus, in order to evaluate
the wave propagate and absorption properly, to introduce the dielectric constant tensor
including the thermal motion of plasma is required
However, in the case assumed the dielectric constant tensor including the thermal
motion of plasma, non-physical minus heating is calculated. Its minus heating is thought to becaused a numerical error when the electric field components are transformed to the Stix frame
defined in the basis of the field line direction.
In order to reduce the numerical error, we change the electric field components fromthe torus magnetic coordinates to the coordinates like the Stix frame, Erho, Eperp, Epara in the
TASK3D/WM. In addition, the TASK3D/WM where finite element method is used in theminor radius has been developed too. In the conference, the modeling and algorithm of
TASK3D/WM will be explained in detail. The primary results of two codes will be shown.
[1] A. Fukuyama and T. Tohnai, in 5th IAEA Technical Committee Meeting on Alpha Particles
in Fusion Research, IAEA, Vienna, 1997.
[2] R. Seki, et al., Proc. of plasma conference 2012, 23P095-P.

- 219 -

P2-46
Analysis of High Energy Electron Tail and Its Effect on the H- Productionin the Linac4 H- Source
S. Mochizuki1, S. Mattei 2, K. Nishida1, A. Hatayama1 and Jacques Lettry2
1

Faculty of Science and Technology, Keio University, 3-14-1 Hiyoshi, Kohoku-ku, Yokohama 223-8522,
Japan
2
CERN, 1211 Geneva 23, Switzerland

A numerical simulation code based on the Electromagnetic Particle in Cell
(EM-PIC) Model with Monte Carlo method for Collision processes (MCC) has
been developed to understand the Radio Frequency Inductively Coupled Plasmas(RF-ICP) [1]. The code has already been applied to the simulation of the Linac4
H- source plasma. The code is able to obtain the Electron Energy Distribution
Function (EEDF) and Ion Energy Distribution Function (IEDF) in RF-ICP. Thisis the most distinct feature of the code compared with the conventional fluid
model. The code has been improved step by step by including various effects, suchas i) capacitive component of the electric field[2], ii) Coulomb collision[3] and iii)
coupling to the Collisional Radiative (CR) model[4,5] for the calculation of Balmer
emission lines from the source plasmas. Most of these studies, however, werefocused on the macroscopic property of RF plasmas and detailed discussion of the
EEDF in RF plasmas has not been given so far.
The purpose of this study is to clarify the characteristic features of EEDFs in
RF H- Sources. We have performed the analysis of EEDF with simple theoretical
approach [6] based on the Boltzmann equation. Comparison between the EEDFs
derived from analysis mentioned above and those obtained by numericalsimulation shows that the low energy parts of EEDFs become Maxwellian
distribution. In addition, the temperature calculated by the simulation agrees
well with theoretically calculated one. On the other hand, it has not been clarified
the relationship between the high energy part of EEDF and plasma source
parameters (RF driven frequency, RF-coil current, etc.). High energy part of
EEDF plays a key role to enhance the dissociation of molecules. The resultant
enhancement of the atomic flux onto the PG leads to the increase in the surface
H- production. The parameter survey for the dependence of the high energy partof EEDF on these parameters is now underway and the results will be presented.
[1] S. Yoshinari, T. Hayami, R. Terasaki, A. Hatayama and A. Fukano, Rev. Sci. Instrum. 81,
02A728 (2010)[2] S. Mattei, M. Ohta, M. Yasumoto, A. Hatayama, J. Lettry and A. Grudiev, Rev. Sci.
Instrum. 85, 02B115 (2014)[3] S. Mochizuki, S. Mattei, T. Shibata, K. Nishida, A. Hatayama and J. Lettry, AIP Conf.
Proc. 1655, 020016 (2015)[4] T. Yamamoto, T. Shibata, M. Ohta, M. Yasumoto, K. Nishida, A. Hatayama, S. Mattei, J.
Lettry, K. Sawada and U. Fantz, Rev. Sci. Instrum, 85, 02B118 (2014)[5] T. Shibata, S. Mattei, K. Nishida, A. Hatayama and J. Lettry, AIP Conf. Proc. 1655,
020008 (2015)[6] V. E. Golant, A. Zhilinskii, S. A. Sakhavov, Fundamentals of Plasma Physics, (Wiley,
New York, 1977)- 220 -

P2-47
Effect of Coulomb Collision on the Negative Ion Extraction Mechanism
in Negative Ion Sources
I. Gotoa, K. Miyamotob, S. Nishiokaa, S. Abea, A. HatayamaaaFaculty of Science and Technology, Keio University, 3-14-1 Hiyoshi, Kohoku-ku, Yokohama
223-8522, Japanb
Naruto University of Education, 748 Nakashima, Takashima, Naruto-cho, Naruto-shi, Tokushima
772-8502, Japan

Negative hydrogen ion (H-) beam has a wide range of applications, such as fusion plasma
heating, accelerators for medical use and high energy particle physics. However, the extraction
mechanism of negative ions from the ion sources has not been fully understood yet. To clarify
the extraction mechanism of H- ions from H- negative ion sources is an important subject for
optimizing the negative ion beam extraction. In our previous study, numerical simulations using
the 2D3V-PIC (two dimension in real space and three dimension in velocity space)·(Particle In
Cell), and 3D3V-PIC model, aiming to understand the H- extraction mechanisms[1-3].
In this study, Coulomb collisions such as electron-H-, H--H+ have been taken into account toour 2D3V-PIC to (a) clarify the energy relaxation of the surface produced H- ions and (b) toestimate the detailed effect of electron transport across the filter magnetic field.
Initial energy of the surface produced H- ions is set to be TH-SP0 = 1.0 eV for the energy of thesurface produced negative ion because the energy of H- ions produced on the PG surface isestimated to be ~ 1 eV from the theoretical formula[4]. The energy relaxation process isimportant because the energy of surface produced H- ions strongly affect the beam emittance.
The preliminary result of the simulation including the Coulomb collision process shows that theenergy of H- ions in the region 8 mm far from the inner surface of the PG has been reduced TH= 0.66 eV due to the Coulomb collision with the lower energy particles.
Electron diffusion across the magnetic field was introduced by a simple model derived fromdiffusion equation with some approximations in the previous PIC model of the extraction
region[6]. In the present model, electron diffusion across the filter magnetic field can be taken
into account naturally without such a simple model, because we include the full effect ofCoulomb collision. Comparison of the results by the previous simple diffusion model with thepresent model including the full effect of Coulomb collision will be presented. The effect ofelectron diffusion across the magnetic field on the potential structure near the PG and theresultant H- extraction will be also discussed in this paper.

[1] K. Miyamoto, S. Okuda, and A. Hatayama, Appl. Phys. Lett., 100, 233507(2012).
[2] K. Miyamoto, S. Okuda, S. Nishioka, and A. Hatayama, J. Appl. Phys., 114,
103302/1-7(2013).
[3] S. Nishoka, K. Miyamoto, S. Okuda, I.Goto, A. Hatayama and A. Fukano, Rev. Sci. Instrum.
85, 02A737/1- 3(2014).
[4] R. McAdams, A. J. T. Holmes, D. B. King, and E. Surrey, Plasma Sources Sci. Technol. 20,035023-1 (2011).
[5] S. Nishoka, K. Miyamoto, S. Okuda, I.Goto, A. Hatayama and A. Fukano, Rev. Sci. Instrum.
85, 02A737/1- 3(2014).

- 221 -

P2-48Saturation mechanism of ion temperature gradient driven trapped electronmode turbulence
Y. Idomura
Japan Atomic Energy Agency, Wakashiba 178-4, Kashiwa, Chiba 277-0871, Japan

Full-f gyrokinetic simulations are important tools in studying stiff temperature profiles
imposed by non-local transport, non-diffusive momentum transport leading to intrinsic
rotation, interplay between neoclassical and turbulent transport, and plasma size scaling of
turbulent transport [Idomura, NF09, CSD12, POP14a, POP14b, PFR14]. However, most of
existing full-f gyrokinetic simulations are limited within electrostatic ion temperature gradientdriven (ITG) turbulence simulations with adiabatic electrons, and electron transport problems
have been still open in fill-f approaches. In this work, we develop a kinetic electron model forelectrostatic ITG and trapped electron mode (TEM) turbulence simulations in the Gyrokinetic
Toroidal 5D full-f Eulerian code, GT5D. In the model, a full kinetic electron model including
both trapped and passing electrons is solved for collisional processes and radial electric fields.
In the axisymmetric limit, this model imposes a neoclassical equilibrium with an ambipolar
condition. The neoclassical transport is computed using a multi-species linear Fokker-Planck
collision operator [Sugama,POP09], and the particle transport, the ion and electron heattransport, the bootstrap current, and the radial force balance condition are verified against thestandard moment approach [Hirshman,NF87]. On the other hand, turbulent fluctuations arecomputed by kinetic (adiabatic) response of trapped (passing) electrons in order to avoid ahigh frequency mode, which appears as the electrostatic limit of kinetic Alfven waves. It isnoted that unlike a bounce-averaged trapped electron model [Idomura,JPFR04], this model
involves collisional and turbulent trapping and de-trapping processes, which are essential asphysically relevant source and sink models for the trapped electron distribution. The linear
ITG-TEM stability is verified against former benchmark results [Rewoldt,CPC04], and linear
critical gradient parameters of the ITG mode are confirmed. In this work, we address
saturation mechanisms in decaying ITG (adiabatic electrons) and ITG-TEM (kinetic electrons
with mi/me=100) turbulence simulations using the same Cyclone parameters [Dimits,POP00]
with a small plasma size parameter 1/ρ*~150. In both simulations, density and temperature
profiles evolve during nonlinear saturation processes, and turbulent transport is quenchedafter the nonlinear saturation phase. What is important here is that in the nonlinear
quasi-steady phase, the temperature profiles are sustained above linear critical temperature
gradient parameters of the ITG mode. This kind of nonlinear critical gradient is often
explained by stabilizing effects of turbulence driven zonal flows. In fact, in the ITG
turbulence, a radial electric field Er contains turbulence driven zonal flows, and its shearing
rate ωE~0.1vti/Ln is comparable to the linear growth rate of the ITG mode. However, it isfound that this mechanism is not the case for the ITG-TEM turbulence. A remarkable featureof the full-f ITG-TEM turbulence simulation is that a corrugated density profile is generated
in the saturation process, and Er is determined by a radial force balance relation, where adensity gradient term is dominant. The resulting Er also shows a fine corrugation, and its
shearing rate is quite high ωE~vti/Ln. In the simulation, it is found that the density profile
structure is related to passing electrons transport near low order mode rational surfaces, where
non-adiabatic response of passing electrons may become important [Dominsky,POP15].

- 222 -

P2-49Kinetic equilibrium of a non-adiabatic trap
T. Takahashi, H. Momotaa
Division of Electronics and Informatics, Gunma University,1-5-1 Tenjin-cho, Kiryu 376-8515, Japana
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

Quest for an economically-competitive nuclear reactor concept is necessary for post-ITER
phase. A non-adiabatic has attractive properties to be a candidate of a next-generation compact
reactor. It consists of the solenoid and mirror coils, and a field-canceling coil (the Helmholtzcoil). The magnetic field of the trap vanishes widely near its center region. As a result, a
confined plasma is in a high-beta state. Moreover, the confined plasma is expected to be stableagainst interchange modes because the field lines surrounding the core plasma have good
curvature. Since plasma ions move stochastically in the zero-field region, the mirror
confinement concept breaks and the probabilistic nature dominates an end-loss process in the
non-adiabatic traps. One of the authors has proposed a new concept of the multi-linked
non-adiabatic traps [1], where plasma particles escaping a trap are re-trapped in neighboringtraps. The resultant overall particle confinement time can be extended in proportion to thesquare of the number of traps; it offers drastic improvement of confinement.
We study here kinetic equilibrium state of the plasma confined in a non-adiabatic trap by
2-dimensional (2-d) hybrid simulation in order to find the 2-d profiles of the diamagnetic
current and the magnetic field structure. Detailed results will be reported at our presentation.

[1] H. Momota, G. H. Miley, and O. Motojima, J. Fusion Energ. 27(2008) 77.

- 223 -

P2-50Kinetic behavior of electron magnetohydrodynamic structures
M. Hata, H. Sakagamia, A. Dasb
Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita 565-0871, Japana
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
Institute for Plasma Research, Bhat, Gandhinagar 328428, India

The Electron Magnetohydrodynamic (EMHD) fluid model (simplified fluid depiction of fastelectron time scale phenomena for plasmas) supports exact nonlinear solutions in the form ofelectron vortex currents with associated induced magnetic field in two-dimensional uniform
plasma [1,2]. The consideration of immobile ion and zero temperature employed in the
EMHD formalism, however, is pretty restrictive. Recent simulation studies have shown thatsuch structures can be created as a result of laser-plasma interaction and these structures play
important roles in the beam electron energy dissipation [3,4]. Understanding the kinetic
behavior of EMHD structures become more important, but these properties have not beenclarified yet. Therefore kinetic behavior of EMHD structures has been explored in the present
study. According to recent research, kinetic effects on robustness of EMHD monopole
solution have been studied and its robust condition was investigated [5]. In this study, kinetic
effects on propagation of EMHD dipole structures in homogeneous and inhomogeneous
plasma are investigated.
For this purpose, Particle-In-Cell (PIC) simulations that include kinetic effects are conducted.
The exact dipole solution derived in EMHD model, which approximately consists of a
combination of two counter-rotating electron vortex currents and their induced magnetic
fields, is used as initial conditions. This solution has traveling property and it can travel tohigher density region depending on the situation. Electron velocities are given to satisfy
Maxwell velocity distribution with the given electron temperature and then thecurrent-derived drift velocity is added to each electron. Changing the electron temperature, we
perform PIC simulations to investigate finite temperature effects. Furthermore we also
investigate effects of density inhomogeneity on dipole behavior.
[1] M. B. Isichenko and A. M. Marnachev, Sov. Phys. JETP 66 (1987) 702.
[2] A. Das, Plasma Phys. Controlled Fusion 41 (1999) A531.
[3] T. Nakamura, S. V. Bulanov, T. Z. Esirkepov, and M. Kando, Phys. Rev. Lett., 105(2010) 135002.
[4] Q. Jia, K. Mima, H. B. Cai, T. Taguchi, H. Nagatomo, and X. T. He, Phys. Rev. E, 91.2
(2015) 023107.
[5] M. Hata, H. Sakagami, and A. Das, Phys. Plasmas, 20.4 (2013) 042303.

- 224 -

P2-51
Simulation study on helium in the vicinity of tungsten grain boundary
A. Takayamaa, A. M. Itoa,b, H. Nakamuraa,caNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
Sokendai,322-6 Oroshi-cho, Toki 509-5292, Japanc
Nagoya University,322-6 Oroshi-cho, Toki 509-5292, Japan

Tungsten is one of plasma facing materials in magnetic confinement nuclear fusion reactors. In
the surface region of plasma facing materials, high concentrations of hydrogen isotopes andhelium as a nuclear fusion product can build up and interact with the materials. These
interactions will induce changes in the surface morphology and the bulk microstructure and thus
in the mechanical properties. Therefore, the mechanical properties and the structural strength oftungsten under exposure to plasma irradiation have been intensively investigated bothexperimentally and theoretically.
In our previous work [1], the binding energies of helium and hydrogen clusters interstitiallytrapped in bcc tungsten are evaluated by first-principles calculations based on density functional
theories (DFT). It is shown that helium-rich interstitially-trapped clusters have the positive
binding energies and the low electron-density region expands as the number of helium in thecluster increases. These results indicate that the helium-rich interstitially trapped clusters canact as a trapping site for hydrogen, and interstitially trapped helium interrupts or disturbs thehydrogen diffusion in tungsten.
Tungsten material is polycrystal in reality, containing grain boundary. In the present study,
helium atom(s) in the vicinity of tungsten grain boundary is investigated by use of DFT
calculation. Difference between helium behavior in bcc tungsten and one in the vicinity oftungsten grain boundary will be discussed.

[1] A. Takayama, A. M. Ito, Y. Oda, H. Nakamura, J. Nucl. Mater. 463 (2015) 355.

- 225 -

P2-52Three-dimensional numerical analysis of shear ﬂow eﬀects on
MHD stability in LHD plasmas
K. Ichiguchia,b , Y. Suzukia,b , Y. Todoa,b , M. Satoa , T. Nicolasa , B. A. Carrerasc , S.
Sakakibaraa,b , Y. Takemuraa , S. Ohdachia,b , Y. Narushimaa,ba National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb SOKENDAI, The Graduate University for Advanced Studies, Toki, 509-5292, Japanc Universidad Carlos III, 28911 Leganes, Madrid, Spain

In the Large Helical Device (LHD), experiments for high beta achievement has been extensively conducted. The highest value of hβi = 5.1% was obtained in the conﬁguration with
Rax = 3.6m and γc = 1.20 [1]. Here, Rax and γc are the horizontal position of the vacuum
magnetic axis and the parameter of the aspect ratio of the helical coils, respectively. On
the other hand, the plasma in this conﬁguration is predicted to be unstable with respect tothe Mercier criterion. It follows that some stabilizing eﬀects work on the LHD plasma forthe achievement of the high beta. In order to design the coil conﬁguration in the heliotron
DEMO accurately, it is crucial to identify the stabilizing mechanism in the viewpoint of themagnetohydrodynamic (MHD) stability. In the recent experiments, the phenomenon like alocked mode was observed [2,3]. In this case, the mode grows rapidly just after the mode
rotation stops, and brings a partial collapse of the proﬁle in the electron temperature. Thisphenomenon indicates that the shear ﬂow of the plasma may be a candidate which can suppress the growth of the mode. Thus, we numerically study the eﬀects of the shear ﬂow on thestability against the pressure driven modes in the LHD plasmas. For this purpose, we utilize
three-dimensional (3D) numerical codes. It has not been established how to obtain a 3D
equilibrium consistent with ﬁnite ﬂows. Hence, we utilize a static equilibrium, and then, we
set a ﬁnite ﬂow in the stability examination, as the ﬁrst step of the ﬂow analysis. In the equilibrium and the stability calculations, the HINT2 code [4] and the MIPS code [5] are utilized,
respectively. In both codes, rectangular grids are employed in the coordinates (R, φ, Z). The
MIPS code examines the stability by following the time evolution of the plasma. We set theshear ﬂow v satisfying v · ∇Ψ = 0 as the initial condition of the MIPS code. Here, Ψ denotes
the label of the equilibrium ﬂux surface. The proﬁle of the ﬂow close to the experimental
data is assumed. It is reported how the shear ﬂow aﬀects the linear growth rate and thenonlinear dynamics of the pressure driven modes.

[1] H. Yamada, et al., 2011 Nucl. Fusion 51 pp.094021.
[2] Y. Takemura, et al., 2012 Nucl. Fusion 52 pp.102001.
[3] S. Sakakibara, et al., 2013 Plasma Phys. Control. Fusion 55 pp.014014.
[4] Y. Suzuki, et al., 2006 Nucl. Fusion 46 pp.L19.
[5] Y. Todo, et al., 2010 Plasma and Fusion Res. 5 pp.S2062.

- 226 -

P2-53Pressure anisotropy analyses based on MHD equilibrium
calculations and magnetic diagnostics in LHD
R. Uedaa , Y. Suzukia,b , K.Y. Watanabea and W. A. Cooperc
a National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb SOKENDAI, The Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki 509-5292,

Japan
c Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Researches en Physique desPlasmas, Association Euratom-Confederation Suisse, CH1015 Lausanne, Switzerland

Large Helical Device (LHD) is a heliotron-type plasma confinement experiment device
which has achieved the volume averaged β that exceeds 5% [1]. The high-β plasmas areproduced by the tangential Neutral Beam Injection (NBI) in the low density and the low
magnetic field. For the conditions, the beam pressure enhances the parallel component to
the magnetic field line, which leads to the pressure anisotropy. The pressure anisotropy is
expected to aﬀect on the characteristics of transport and stability. Additionally, for the
anisotropic pressure, the shape of the magnetic flux surface does not coincide with the pressure isosurface.
However, the plasma pressure of the future reactor device is predicted to be isotropic
because the α-heating, which is a major heating system in a fusion reactor, produces theisotropic velocities. To extrapolate the knowledge obtained by LHD high-β discharges tofuture device designs, we should investigate the pressure anisotropy of the LHD experimentplasmas.
In previous works, an identification method for the pressure anisotropy has been proposed[2], which has been studied using ANIMEC code [3]. The code is an extended version
of VMEC, and can treat the anisotropic pressure. The study shows the anisotropy canbe predicted tolerably even if we use VMEC code. The equally weighted averaging β ofthe perpendicular and parallel component, βeq , and the perpendicular component, β⊥ , areobtained by the saddle loop flux and the diamagnetic flux, respectively. Then, the parallel
component, β∥ , is calculated by β∥ = 2βeq − β⊥ . The identification of the pressure profile iscarried out by investigating the relationships between the saddle loop flux and the location of
the magnetic axis. However, the application to the actual LHD experiment plasma has notbeen studied in detail. In this study, we apply the method to a number of LHD discharges
and show the characteristics of the equilibrium focusing on the anisotropy and the pressure
profile.

[1] A. Komori et al., Nucl. Fusion 49 (2009) 104015.
[2] Y. Asahi el al., Phys. Plasmas 20 (2013) 022503.
[3] W. A. Cooper et al., Comput. Phys. Commun. 180 (2009) 1524.

- 227 -

P2-54
Benchmark of multi-ion-species collision operator for δfMonte-Carlo neoclassical simulation
S. Satakea,b , M. Nakataa,b , T. Pianpanitb , S. Matsuokac , M. Nunamia,b , H. Sugamaa,ba National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japanb Sokendai, 322-6 Oroshi-cho, Toki 509-5292, Japanc Research Organization for Information Science and Technology (RIST), 1-5-2

Minatojima-Minamimachi, Chuo-ku, Kobe 650-0047, Japansatake@nifs.ac.jp

To study and evaluate the neoclassical transport process in burning plasma in nuclear fusion
reactor, it is required to develop a neoclassical transport simulation code for multiple ion
species plasmas which treats the Coulomb collisions between unlike ion species. We plan
extend the neoclassical simulation code FORTEC-3D[1,2] by implementing the unlike-species
collision operator which keeps the self-adjointness nature even if the temperature of two
ion species are diﬀerent[3]. In δf neoclassical transport simulation, the linearized collision
operator for the perturbation part of the distribution function of particle species a, δfa =fa − fM a , should satisfy the following properties.
∫

∫

3

∫

ma

T
d vCab(δfa )

F
d3 vCab(δfb ) = 0,=T
d3 v{v∥ , v 2 }Cab(δfa ) = −mb∫∫F
d3 v{v∥ , v 2 }Cba
(δfa ),

(1)
(2)

∫

δfa T
δg T
dv
Cab (δg) =d3 v
C (δfa ),faM
faM ab
∫
∫
δfa F
δgb F
Ta d3 v
Cab (δgb ) = Tb d3 v
C (δfa ),faM
fbM ba3(3)(4)

where Eqs. (1)-(3) represents the conservation property and the last two are the adjointnessof the operator. In this presentation, a numerical method to implement the collision operator
in the δf Monte-Carlo particle code is explained. Benchmark of the new collision operator
with the same one implemented in a gyrokinetic continuum code[4] and the other one in afull-f PIC code, based on Nanbu-method[5] are to be reported. Dependence of the numericalaccuracy of the operator on the mass-ratio is studied.
To accelerate the collision calculation among many particles species, we test the implementation of the simulation code on the Intel Phi accelerator. Optimization method of the codefor the Phi architecture is explained.

[1] S. Satake, R. Kanno, and H. Sugama, Plasma Fusion Res. 3 (2008) S1062[2] S. Matsuoka et al., Phys. Plasmas 22 (2015) 072511[3] Sugama et al., Phys. Plasmas 16 (2009) 112503[4] M. Nakata, et al., ”Improved collision operator for plasma kinetic simulations with multispecies ions and electrons” Comput. Phys. Comm. (2015) in press[5] K. Nanbu, Phys. Rev. E 55 4642 (1997)- 228 -

P2-55
Improvement of a Multi-Hierarchy Simulation Model for Magnetic
Reconnection Studies
S. Usamia, H. Miuraab, H. Ohtaniab, R. Horiuchiab, M. DencaNational Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanThe Graduate University for Advanced Studies (SOKENDAI), 322-6 Oroshi-cho, Toki 509-5292,
JapancNational Institute of Information and Communications Technology, 4-2-1 Nukui-Kitamachi, Koganei,
184-8795, Japanb

A fusion plasma consists of multiple hierarchies. It is characterized by the wide range ofspatiotemporal scales from macroscopic transports and instabilities covering the entire system
to microscopic processes relating to the dynamics of each electrons and ions. Furthermore, afusion plasma is composed of not only multiple hierarchies but also multiple sciences, such asplasma physics, material science, and nuclear physics. In fusion plasma studies with computer
simulations, different hierarchies or physics are treated independently by different models.
Studies on each hierarchy or physics are advancing remarkably [1]. On the other hand, it ispoorly understood that how and where multiple hierarchies or physics couple each other. For thecomplete comprehension of fusion plasmas, simulation models which deal with multiple
hierarchies or physics self-consistently and simultaneously are required.
We have developed a multi-hierarchy simulation model which couples an MHD code and a PIC
code [2]. Using our model, we have studied on collisionless magnetic reconnection, sincemagnetic reconnection is one of typical multi-hierarchy phenomena and is believed to bestrongly involved with sawtooth oscillations, which cause tokamak disruption. Our
multi-hierarchy model employs the domain decomposition method, namely real space in asimulation domain is divided into the following three parts: an MHD domain to deal with global
plasma dynamics away from reconnection points, a PIC domain to solve kinetic physics in thevicinity of the neutral sheet, and an interface domain to commute the MHD and PIC informationand interlock the MHD and PIC domains. In 2014, by means of our multi-hierarchy simulations
we have investigated on the influence of macroscopic dynamics on microscopic physics ofmagnetic reconnection and reported the first results [3].
In the near future, we have a plan to apply our model to sawtooth oscillations. In fusion devices
such as tokamak, the finite Larmor radius and Hall effects play important role in the entire
region of a device. As ongoing activities, we improve our multi-hierarch model as follows. An
extended MHD code [4] including non-ideal terms such as the finite Larmor radius and Hall
effects, and a PIC code are interlocked (In the current model, an ideal MHD code and a PIC
code are coupled). In our presentation, we would like to introduce the development plan of ourmulti-hierarchy model and demonstrate simulation results performed with the improved modelas the first step.
[1] R. Horiuchi and NSRP members, Plasma Fusion Res. 6 (2011) 2101055.
[2] S. Usami, R. Horiuchi, H. Ohtani, and M. Den, Phys. Plasmas 20 (2013) 061208.
[3] S. Usami, R. Horiuchi, H. Ohtani, and M. Den, Journal of Physics: Conference Series 561(2014) 012021.
[4] R. Goto, H. Miura, A. Ito, M. Sato, and T. Hatori, Phys. Plasmas 22 (2015) 032115.

- 229 -

P2-56
Neutron incident angle and energy distribution to vacuum vesselin beam-injected LHD deuterium plasmas
S. Sugiyama, H. Matsuura, T. Gotoa
Department of Applied Quantum Physics and Nuclear Engineering,
Kyushu University, 744 Motooka, Fukuoka 819-0395, Japana National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

In toroidal devices, it is known that neutron wall loading and incident flux to first wall dependon poloidal angular wall position because of the torus wall shape [1, 2]. As the vacuum vesselof helical devices has a dumbbell-shaped poloidal cross section, the shape of the first wall
would cause additional nonuniformity of the neutron incident flux. The neutron flux has been
evaluated for neutron measurement, e.g. in situ calibration for neutron monitor [3], using
Monte Carlo simulation code in the Large Helical Device (LHD) approximately incorporating
the complex shape of the vacuum vessel. Neutron emission profile can be taken into accountfor these calculations [4]. However, analysis of the neutron flux for eﬀects on structural
materials has not been studied. From this point of view, it is important that the dependenceof the neutron energy spectra on incident angle at all wall positions is clarified. In addition to
the geometrical eﬀects of the first wall, anisotropic high-energy ion tail formation in fuel ion
velocity distribution function [5, 6] also aﬀects the neutron flux distribution, energy spectra,
and incident angle distribution.
In this study, the neutron incident angle distribution and energy spectra at all wall positions
are evaluated assuming deuterium-beam-injected deuterium plasmas confined in the LHD
considering exact shape of the vacuum vessel and the fuel ion velocity distribution function.
It is presented that the neutron spectra significantly diﬀer by the wall position and incident
direction.

[1] P. P. H. Wilson, et al., Fusion Eng. Des. 83 (2008) 824.
[2] J. C. Rivas, et al., Fusion Sci. Technol. 64 (2013) 687.
[3] Y. Nakano, et al., Rev. Sci. Instrum. 85 (2014) 11E116.
[4] M. Homma, et al., Plasma Fusion Res. 10 (2015) 3403050.
[5] H. Matsuura, Y. Nakao, Phys. Plasmas 16 (2009) 042507.
[6] H. Matsuura, et al., Plasma Phys. Control. Fusion 53 (2011) 035023.

- 230 -

P2-57
Particle-in-cell simulations of a kink instability
in a pinched current filament
T. Haruki
Graduate School of Science and Engineering, University of Toyama,3190 Gofuku, Toyama, Toyama 939-8555, Japan

haruki@eng.u-toyama.ac.jp

Current filaments are often observed in astrophysical and laboratory plasmas. It is wellknown that a filament pinches [1] and/or becomes unstable against magnetohydrodynamic(MHD) instabilities. Understanding the physics of the current filaments is important to avariety of applications, including electric discharge, solar flares and a pinch device.
Haruki et al. already reported the dynamics of a current filament in three dimensional,
relativistic, fully electromagnetic particle-in-cell (PIC) simulations [2]. The plasma column
located at the center of system is initially accelerated by an externally applied electric field
to produce a current filament. The filament then starts to pinch, due to the Lorentz force,and becomes unstable initially against a sausage instability and then a kink instability. Highenergy particles are produced by these two types of instabilities.
However, further analysis showed that out-going electromagnetic waves were reflectedfrom the numerical boundaries of the system and affect the current filament at the center.
These reflected waves enhanced the current filament instabilities.
In this study, more accurate simulations of a pinched current filament were performed
with absorbing boundary conditions applied perpendicular to the axial direction. As a result,the enhancement of instabilities are suppressed and the current filament pinches only weakly.
For the strong current filament, it starts to pinch while oscillating and becomes unstableagainst a kink instability (a sausage instability is not observed here). The positive andnegative isosurface levels of the axial component of an electric field and current density
showed double helical structures. After the kink instability, the current filament diffuses andthe ions are heated.
These results are important to understanding not only the basic plasma physics but
also to the newly proposed concept of focus fusion [3], which uses a dense plasma focus
device, improved upon from a pinch device, to produce a p-11 B reaction without neutrons(aneutronic).

[1] W. Bennett, Phys. Rev. 45, 890 (1934)[2] T. Haruki et al., Phys. Plasmas 13, 082106 (2006)[3] E.J. Lerner et al., Phys. Plasmas 19, 032704 (2012)- 231 -

P2-58
Basic Comparison of the Numerical Method to Solve Fluid Equationsfor SOL/Divetror Plasmas
Ryoko Tatsumi1, Yuki Homma2, Shohei Yamoto1, Kazuhiro Ishibashi1,
and Akiyoshi Hatayama1
1

Graduate School of Science and Technology, Keio University, 3-14-1, Hiyoshi, Kohoku-ku,
Yokohama 223-8522, Japan
2
Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japan

Understanding plasma transport in the SOL/Divertor region is one of the most important
issues in order to reduce heat and particle loads to the divertor plate in future fusion reactors.
For this purpose, several numerical codes based on the fluid plasma equations have beendeveloped so far. Among them, there are mainly two types of method, finite difference method
like B2-EIRENE(SOLPS)[1], and Monte Carlo method like EMC3[2]. The former seems to beharder for 3D (three dimensional) large scale simulations, while the latter seems to havedifficulty on setting boundary conditions. Therefore, the objective of our study is to solve thebenchmark problem with both methods and compare them.
The plasma transports can be described as the following convection-diffusion equation[3]:∇∥ ∙ [𝛼∥ 𝑓 − ∇∥ (𝛽∥ 𝑓)] + ∇⊥ ∙ [𝛼⊥ 𝑓 − ∇⊥ (𝛽⊥ 𝑓)] = 𝑆(1)
where 𝑓 stands for density, momentum, and temperature of the plasma. (Other notations are
given in the Ref. [3]) In one-dimensional case where 𝛽 = 0, 𝑆 = 0 for density, 𝛽 = 0, 𝑆 =−∇𝑝 for momentum, and 𝑆 = 0 for temperature, the analytic solution can be obtained as:
𝑚𝑣02 + 𝑇0 + √(𝑚𝑣02 + 𝑇0 )2 − 4𝑚𝑣02 𝑇(𝑥)(2)𝑣(𝑥) =, 𝑛(𝑥) = 𝑛0 𝑣0 /𝑣(𝑥)2𝑚𝑣0
where 𝑥 is the distance along the magnetic field, 𝑛, 𝑣, and 𝑇 correspond to the density,
velocity, and temperature, respectively.
As the first step, we have developed a simple 1D simulation model using the Monte Carlo
method and conducted the calculation in the case mentioned avobe with the boundary
conditions, 𝑣0 = (𝑇0 /𝑚)1/2 and 𝑞 = −𝜅𝜕𝑇/𝜕𝑥(0) = 10MW/m2 . The results agree with theanalytic solutions (see Fig. 1). The comparison between the Monte Carlo method and finite
difference method will be presented at the conference as the next step.
(a)

(b)

Fig. 1. (a) Convergence property of the iteration [Relative change of the residuals for each equation of (1)](b) Numerical solution of f (n, v, and T) at the iteration number N=15 for 1D simulation along the field line.

[1] R. Schneider et al., J. Nucl. Mater. 196-198, 810 (1992).
[2] Y. Feng, et al: Contrib. Plasma Phys. 54 (2014) 426-.
[3] Y. Feng, et al: J. Nucl. Mater. 266-269 (1999) 812-818.

- 232 -

P2-59Multi-scale simulation of magnetic reconnection using particle-in-cell withmagnetohydrodynamics on adaptively refined mesh
T. Ogawa, S. Usamia, R. Horiuchia, M. Denb, K. Yamashitac
Kitasato University,1-15-1 Kitasato, Sagamihara 252-0373, Japan
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb
National Institute of Information and Communications Technology,4-2-1 Nukui-kitamachi, Koganei
184-8795, Japanc
University of Yamanashi,4-4-37 Takeda, Kofu 400-8510, Japana

Magnetic reconnection is a phenomenon during which magnetic field topology changes andmagnetic energy is transformed into plasma kinetic or thermal energy. Both macroscopic
plasma dynamics and microscopic particle kinetics play important roles there. In numerical
study, magnetohydrodynamics (MHD) is an efficient method to deal with macroscopic plasma
dynamics, but it needs some artificial resistivity or non-MHD effects to model the phenomenon[1, 2]. While particle methods can bring microscopic kinetics, they need higher computationalcost. Usami et al. developed an efficient multi-scale simulation model of magnetic reconnection.
They used particle-in-cell (PIC) method near a magnetic neutral sheet and MHD distance fromthe sheet [3, 4, 5]. In our previous study, we extended their MHD part to a hierarchical mesh
controlled by adaptive mesh refinement (AMR) technique and calculated Alfvén wave
propagation to test the code [6].
We apply the code to a magnetic reconnection in this study (Figure). PIC is localized near theneutral sheet while a hierarchical mesh for MHD spreads outer region. In addition, we importexternal physical quantities onto outer boundary. It enables to decide outer boundary conditions
by a global structure of another simulation. We aim to link a global MHD model, such as amodel of Earth’s magnetosphere, and a PIC calculation smoothly by the AMR-controlledhierarchical mesh sandwiched as a bridge.
This work was supported by JSPS Grant-in-Aid for Scientific Research (B) 23340182.

Resulting density distribution of the magnetic reconnection model.

[1] M. Ugai, Phys. Fluids 4 (1992) 2953.
[2] M. Yamada, Phys. Plasmas 14 (2007) 058102.
[3] S. Usami, H. Ohtani, R. Horiuchi, and M. Den, Commun. in Comput. Phys. 11 (2012) 1006.
[4] S. Usami, R. Horiuchi, H. Ohtani, and M. Den, Phys. Plasmas 20 (2013) 061208.
[5] S. Usami, R. Horiuchi, H. Ohtani, and M. Den, J.Phys.: Conf. Ser. 561 (2014) 012021.
[6] T. Ogawa, S. Usami, R. Horiuchi, M. Den, K. Yamashita, JPS Conf. Proc. 1 (2014) 016013.

- 233 -

P2-60Three dimensional magnetic reconnection ofstressed X-point collapse
N. Tsujine, T. Harukia
Graduate School of Science and Engineering for Education, University of Toyama, 3190 Gofuku,
Toyama 930-0885, Japana Graduate School of Science and Engineering for Research, University of Toyama, 3190 Gofuku,
Toyama 930-0885, Japan

m1571113@ems.u-toyama.ac.jp

Magnetic reconnection in plasmas is an important mechanism for converting magnetic fieldenergy into thermal plasma heating and non-thermal particle acceleration. In particular,a stressed X-point collapse is a well know mechanism for explaining solar flares [1]. The
magnetic field configuration for an X-point is
Bx = y, By = α2 x, Bz = 0,
where α is the stress parameter, and the x-point is located at the origin.
This model has been numerically simulated using a two dimensional, relativistic, fully
electromagnetic, particle-in-cell (PIC) code [2]. The stress (plasma pressure) causes magnetic
reconnection in the vicinity of the X-point, via the formation of a current sheet. Magnetic fieldenergy is dissipated (about 20% in the strongly stressed case), followed by the productionof high energy plasmas generated by the magnetic reconnection process. However, three
dimensional simulations of the stressed X-point collapse model have not yet been studied.
Including three dimensional eﬀects into the numerical simulation is necessary to understandrealistic magnetic reconnection.
In this study, three dimensional stressed X-point collapse was investigated by using athree dimensional PIC code. The X-point starts to collapse as in the two dimensional PIC
simulations. A similar current sheet is also produced in the vicinity of the X-point. However,magnetic field lines are distorted in the z-axis direction - an eﬀect not seem in the twodimensional case. These results are useful for understanding the physics in three-dimensional
magnetic reconnection.

[1] E. Priest, T. Forbes, Magnetic Reconnection: MHD Theory and Applications (Cambridge
University Press, Cambridge, 2000)[2] D. Tsiklauri, T. Haruki, Physics of Plasma. 14, 112905 (2007)- 234 -

P2-61
Simulation study on resistive instabilities in a small aspect ratio reversed fieldpinch
Yasuo Nagamine and Masamitsu Aizawa
Institute of Quantum Science, Nihon University
1-8-14 Kanda-Surugadai,Chiyoda-ku,Tokyo 101-8308, JapanA reversed-field pinch (RFP) plasma is the result of a self-organization process, which is
produced by a dynamo effect due to magnetohydrodynamic (MHD) instabilities. In the recent
progress of the RFP study, improvement of the confinement properties by the transition to asingle-helical-axis state is observed in the large experimental device. Also recent numericalsimulations show the possibility of reproducing such single-helical-axis states [1]. As for a low
aspect ratio RFP with A≈2, formation and rotation of the helical structure such as quasi-singlehericity states have been observed in experiments [2]. Towards the understanding of thisstructure formation mechanism, analysis using the detailed three-dimensional MHD simulation
through the comparison with the experiment has been conducted [3].
In this study, from an analogy with the spherical tokamak (ST), we focus our attention on
the RFP configuration having an elongated cross section with a small aspect ratio A < 2, in otherwords, spherical RFP. The basic characteristics of the equilibrium configuration and dynamical
behavior of resistive instabilities are examined by using the MHD simulation model. Regarding
the spherical RFP concept, its characteristic already has been introduced briefly in the first
theoritical study on the spherical tokamak [4].
Since a stabilizing shell located near the plasma surface is necessary to sustain the RFP
configuration, it is considered that it is not easy to develop spherical RFP device with a noncircular cross section like the ST. In addition, as the reactor, the extreme lowering of the aspectratio becomes disadvantageous from the viewpoint of neutron irradiation of the torus centerregion. However, considerably high beta plasmas were attained in the ST than the conventional
standard tokamak. With respect to the RFP, it is considered that the theoretical prediction ofplasma properties such as the resistive instabilities and the search for the beta limit condition inthis small aspect ratio regime are much important.

[1] Onofri M, Nuclear Fusion 51, 112003 (2011).
[2] Oki K, et al., J. Phys. Soc. Jpn. 77, 075005 (2008).
[3] Mizuguchi N, et al., Plasmas Fusion Res. 7, 2403117 (2012).
[4] Peng Y-K M and Strickler D J, Nuclear Fusion 26, 769 (1986).

- 235 -

P2-62Large-scale simulation of contactless crack detection in HTS films:
application of H-matrix method to fast matrix-vector
multiplication
T. Takayama, A. Saitoh, A. Kamitani, H. Nakamuraa
Yamagata University, 4-3-16 Johnan, Yonezawa, Yamagata 992-8510, Japana National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan

takayama@yz.yamagata-u.ac.jp

The practical applications of the high-temperature superconductors (HTSs) require themeasurement of a critical current density jC . Recently, Hattori et al. proposed a contactless
and a high-speed measuring method of jC in an HTS film [1]. In the following, this is calleda scanning permanent magnet method. In the method, a permanent magnet is moved along
the HTS surface. As a result, they found that a spatial distribution of jC can be estimatedfrom the distribution of an electromagnetic force acting on the film. However, note that thescanning permanent magnet method has been not yet applied to the crack detection.
In order to reproduce a scanning permanent magnet method numerically, a numericalcode was developed for analyzing the time evolution of a shielding current density in an HTS
film with multiple cracks. After an initial-boundary-value problem of the shielding current
density is spatially discretized with the finite element method (FEM), the resulting ordinary
diﬀerential system is solved by using the backward Euler method and the Runge-Kutta (R-K)
method with an adaptive step-size algorithm. Consequently, it is found that the R-K method
is a useful tool with increasing the number of nodes for the FEM. However, it is necessary
to consider the further high-speed of the shielding current analysis for a large-scale problem.
The purpose of the present study is to develop a numerical code analyzing the time
evolution of the shielding current density in an HTS film with cracks by using the H-matrix
method [2]. Also, by using the code, we simulate the scanning permanent magnet method
for measuring jC and detecting cracks by FEM with using large number of nodes.

[1] K. Hattori, A. Saito, Y. Takano, H. Yamada, T. Takayama, A. Kamitani, S. Ohshima,
Physica C, 471 (2011) 1033
[2] M. Bebendorf, Springer-Verlag (2008)- 236 -

P2-63
Performance Improvement of Extended Boundary Node Method
for Solving Elliptic Boundary-Value Problems
A. Saitoh, T. Takayama, A. Kamitani and H. Nakamuraa
Yamagata University, 4-3-16, Johnan, Yonezawa, Yamagata 992-8510, Japana National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japansaitoh@yz.yamagata-u.ac.jp

Recently, the boundary node method (BNM) [1] which is one of boundary-type meshlessmethods has been proposed. As the feature of the BNM, a boundary does not need to be
divided into a set of elements before executing the BNM code. However, integration cells
must be used for calculating matrix elements.
In order to resolve the above demerit, the BNM has been reformulated without using
integration cells. Throughout the present study, the method is called the extended BNM(X-BNM) [2]. The results of computations have shown that the accuracy of the X-BNM ismuch higher than that of the dual reciprocity boundary element method [3, 4].
In spite of a high usefulness of the X-BNM, the calculation speed for obtaining the numerical solution is extremely slow. This is because the coeﬃcient matrix of the resulting linear
system becomes non-symmetric and dense. In eﬀect, this means that the X-BNM is diﬃcult
to be applied to a large-scale simulation.
The purpose of the present study is to develop the fast numerical code for solving theelliptic boundary value problems on the basis of the X-BNM and to numerically investigate
its performance. The numerical results will be shown in the conference.

[1] Y.X. Mukherjee, S. Mukherjee, Int. J. Numer. Methods Eng. 40(6) (1997) 797.
[2] A. Saitoh, S. Nakata, S. Tanaka and A. Kamitani, Information 12(5) (2009) 973 [inJapanese].
[3] A. Saitoh, N. Matsui, T. Itoh, A. Kamitani, IEEE Trans. Magn. 47(5) (2011) 1222.
[4] A. Saitoh, K. Miyashita, T. Itoh, A. Kamitani, T. Isokawa, N. Kamiura, N. Matsui, IEEE
Trans. Magn. 49(5) (2013) 1601.

- 237 -

P2-64
Stability analysis on node arrangement in Meshless Time-Domain
MethodY. Ohi and S. Ikunoa
RIKEN Advanced Institute for Computational Science, Kobe, Hyogo 650-0047, Japana School of Computer Science, Tokyo University of Technology, Hachioji, Tokyo 192-0982, Japan

yoshiharu.ohi@riken.jp

The meshless time-domain method (MTDM) based on the radial point interpolation method
is employed for electromagnetic wave propagation simulations [1]. In the MTDM, sincemeshes of a geometrical structure does not require, nodes of electric and magnetic fields only
are scattered in an analysis domain. However, appropriate arrangement of nodes is not clear.
In this paper, influence of node arrangement on stability is numerically investigated in order
to find appropriate arrangement of nodes.
Four types of node arrangement are proposed, and schematic views of each node distribution
are shown in Fig. 1, (a): the standard arrangement based on an orthogonal mesh, (b):nodes for Hx move to positions of nodes for Hy , (c): nodes for Hy move to positions ofnodes for Hx , (d): nodes for Hx and Hy move to furthest positions from nodes for Ez .
Numerical simulations of electromagnetic wave propagation by using these node arrangement
are performed. Contour lines of Ex = 0 are shown in Fig. 2. We can see from these resultsthat anisotropy of propagation direction is observed in case of (b) and (c). In addition, since
contour lines crowd, solutions includes vibration except case of (a). If nodes of Hx and Hy
are located at the same position, a numerical simulation may be unstable. It is indispensablethat numerical experiments are performed with various parameters. Details of numericalresults will be shown in ITC25.
1

1

Ez
Hx
Hy

1

Ez
Hx
Hy

1

Ez
Hx
Hy

0.6

0.6

0.6

0.6

0.4

0.4

0.4

0.4

0.2

0.2

0.2

0.2

0

0

0

0

0.2

0.4

0.6

0.8

1

0

0.2

0.4

x [m]

0.6

0.8

1

Ez
Hx
Hy

y [m]

0.8

y [m]

0.8

y [m]

0.8

y [m]

0.8

0
0

0.2

0.4

x [m]

0.6

0.8

1

0

0.2

0.4

x [m]

0.6

0.8

1

x [m]

(a)
(b)
(c)
(d)
Figure 1: Schematic four types of node arrangements ( : Ez , 4: Hx , 2: Hy ).
10

10

10

10

8

8

8

8

6

6
y [m]

0

2

4

6

x [m]

(a)

8

10

6
y [m]

6
y [m]

y [m]

4

4

4

4

2

2

2

2

0

0

0

0

2

4

6

8

10

x [m]

0

2

4

6

8

10

x [m]

(b)
(c)
Figure 2: Contour lines of Ex = 0.

0
0

2

4

6

8

10

x [m]

(d)

[1] Y. Ohi, Y. Fujita, T. Itoh, H. Nanamura, S. Ikuno, Plasma Fution Res., 9:3401144(2014).

- 238 -

P2-65Structure transition in forced magnetohydrodynamic turbulencewith uniform background magnetic field
H. Miura, K. Arakia
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japana Okayama University of Science,1-1 Ridaicho, Kita-ku, Okayama, Okayama 700-0005 , Japan

Direct numerical simulations (DNS) of forced magnetohydrodynamic (MHD) turbulencewith uniform background magnetic field are carried out to elucidate small-scale natures inmagnetized MHD turbulence and provide basic information to construct sub-grid-scale (SGS)model for a Large Eddy Simulation (LES). DNSes are carried out for the two cases, (1) with
the Hall term and (2) without the Hall term. The Hall term is a part of the two-fluid effect in
the generalized Ohm’s law in the extended MHD equations, which have been widely used to
study instabilities including torus plasmas[1-4]. The Hall term restricts the time step width
in a numerical simulation severely and excite high-wave-number coefficients significantly[4].
Consequently, the introduction of the Hall term brings about very small step width andincreases a demand for a very high resolution in a simulation. While the former difficulty
can be managed by the use of an implicit time-marching technique, the latter remains as
a controversial problem. Because of the nonlinearity of the equations, artificial truncationof the short-wave-length components (insufficient grid width) often deteriorates numericalconvergence in a simulation. In order to overcome the difficulty, an LES (See [5], for example)
can be a reasonable compromise between the correctness of a simulation and a computationalcost. Numerical simulations presented here provide basic database to construct numericalmodels required for LESes.
In our study, DNSes of the magnetized MHD turbulence with the uniform backgroundmagnetic field B0 are carried out. Number of grid points N 3 is 10243 . It is shown thatthe introduction of the Hall term brings about an essential change in the small-scales ofturbulent field. While an intese vorticity region is sheet-like (so-called vortex-sheets) in
MHD turbulence, it is tubular in Hall-MHD turbulence, showing that the dominant fluid
motion is changed from the shearing motions to the swirling motions associated with theintroduction of the Hall term. While the transition of the vortex structure has been reported
in Refs.[3,4], we find that some of current sheets at large scales also become tubular due tothe Hall effect. Mechanism of the structure transitions are discussed. We will also estimate
the applicability of the Smagorinsky-type SGS model proposed in Ref.[6] and discuss further
development of the model to a torus plasma instability.
[1] W. Park et al., Phys. Plasmas 6 (1999) 1796.
[2] C.R. Sovinec et al., J. Comput. Phys. 195 (2004) 355.
[3] H. Miura, R. Goto, et al., 25th IAEA FEC, TH/P5-17 (Oct. 13-18 2014, St. Petersberg,
Russia).
[4] R. Goto, H. Miura et al., Phys. Plasmas 22 (2015) 032115.
[5] F. Hamba and M. Tsuchiya, Phys. Plasmas 17 (2010) 012301. [6] H. Miura and K.
Araki, Phys. Plasmas 21 (2014) 072313.

- 239 -

P2-66
Extension of integrated transport analysis suite, TASK3D-a,
to be capable for analysis of fast ion particles and plasma flows in LHD
M. Sato, R. Seki, M. Yokoyama, M. Osakabe,
Numerical Simulation Reactor Research Project, and LHD Experiment Group
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan

The integrated transport analysis suite, TASK3D-a has been developed and applied mainly to
NBI-heated LHD plasmas [1]. Recently, extension has been conducted such as including ECH
ray-tracing codes and the module for creating ascii files to be registered on to the International
Stellarator-Heliotron Confinement Database [2]. In this paper, further extension of theTASK3D-a to make possible the analyses of fast ion particles and plasma flows in LHD is
described. For analysis of the fast ion particles, FIFPC code [3], which solves Fokker-Planck
equation, has been linked to VMEC, NEWBOZ, HFREYA and MCNBI codes (see eg., [4]) as
shown in Fig.1(a). This calculation flow can evaluate the time evolution of the distributionfunction of the fast ion particles in velocity space (v, θ) for NBI-heated plasmas as shown in
Fig.2, where θ is the angle between the velocity vector and the magnetic field line. For
analysis of plasma flows, DKES [5] and PENTA [6] codes have been integrated to
TASK3D-a as shown in Fig.1(b). From this calculation flow, the neoclassical plasma flows on
flux surfaces can be evaluated in addition to such as the neoclassical radial diffusion andneoclassical ambipolar electric field. The application examples for LHD plasmas with thisfurther extended TASK3D-a will also be reported in the conference.
Extensive cooperation by Dr. D. A. Spong (Oak Ridge National Laboratory) for providing PENTA andassisting its implementation into TASK3D-a has been highly acknowledged.

[1] M. Yokoyama et al., Plasma and Fusion Research, 9 (2014) 3402017.
[2] M. Yokoyama et al., to be presented at the 20th International Stellarator-Heliotron
Workshop (Greifswald, Germany, 5-9 October 2015).
[3] R. H. Fowler et al., Comput. Phys. Comm. 13 (1973) 323.
[4] M.Yokoyama for task3d users and developers, NIFS-MEMO-61 (Nov.2012).
[5] W. I. Van Rij and S. P. Hirshman, Phys. Fluids B1 (1989) 563.
[6] D. A. Spong, Phys. Plasmas 12 (2005) 056114.
itv=0001(a) Flow chart for analysis
of fast ion particles(b) Flow chart for analysisof plasma ﬂows

VMEC

VMEC

t=0.5ms

1.2
1
0.8
0.6

v⊥ / v 0

0.4
0.2

itv=0040
0.000001
0.000000

0
-1

-0.5

0

0.5

1

v! / v 0

NEWBOZBOOZXFORM

t=20ms

1.2
1
0.8
0.6

HFREYAMCNBIv⊥ / v 0

0.4

DKES

0.2
itv=0200

0.000020.00001
0.00000
-1

-0.5

0

v! / v 0

0
0.5

1

1.2

t=100ms 1

PENTA

0.8
0.6

v⊥ / v 0

0.4

FIFPC

0.2
0.00004
0.000030.000020.00001
0.00000

0-1

Fig.1. Flow chart for analysis of (a) fast ion
particles and (b) plasma flows.

- 240 -

-0.5

0v! / v 00.5

1

Fig.2. Time evolution of a distributionfunction of fast ion particles for a test
simulation obtained by FIFPC code.

P2-67
Validation of Tungsten Transport Model in JT-60U Plasmas
Y.Shimizua, T.Fujitaa, A.Okamotoa, H.Arimotoa,
T.Nakanob, K.Hoshinoc, N.Hayashib, M.HondabaNagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603, Japan
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japanc
Japan Atomic Energy Agency, 2-166, Omotedate, Obuchi, Rokkasho 039-3212, Japanbshimizu-yuusuke14@ees.nagoya-u.ac.jp
Tungsten (W) is planned to be used as plasma-facing material in ITER divertor, but
accumulation of tungsten in the core plasma is concerned because of its large radiation loss due
to its high atomic number (high Z). In JT-60U H-mode plasma, tungsten accumulation wasobserved to be enhanced as the toroidal rotation opposite to the plasma current (CTR rotation)
increased [1]. This phenomenon cannot be explained by the conventional neoclassical transport.
From theoretical considerations, three effects related to the toroidal rotation (i.e., PHZ pinch,
Er pinch, and centrifugal force) are proposed. Hoshino et al. have proposed PHZ pinch and Er
pinch [2]. The PHZ pinch is caused by changes in the ion charge state along its drift orbit. The
Er pinch is caused by the effect of the radial electric field through Coulomb collisions. The
centrifugal force makes poloidal asymmetry in heavy impurity ion density, which effects theneoclassical transport. In this study, we introduce these three pinch models into transport code
TOTAL [3] and study dependence of the tungsten accumulation on the toroidal rotation. We
simulate the experimental data of the 5 cases in JT-60U H-mode plasmas [1].
The tungsten transport is solved with the following fluxes(1)𝛤𝑘 = 𝛤𝑘 𝑁𝐶 − 𝐷𝑘 𝐴𝑁 𝜕𝑛𝑘 ⁄𝜕𝜌 + 𝑉𝑘 pinch 𝑛𝑘 ， 𝛤𝑘 𝑁𝐶 = −𝐷𝑘 𝑁𝐶 𝜕𝑛𝑘 ⁄𝜕𝜌 + 𝑉𝑘 𝑁𝐶 𝑛𝑘
where 𝛤𝑘 is particle flux of impurities in the charge state k, 𝑛𝑘 is density of impurities in the
charge state k. The subscript NC and AN represent neoclassical and anomalous transports,
respectively. PHZ pinch and Er pinch models are introduced in the 𝑉𝑘 pinch term. Centrifugal
force effects are introduced in the 𝐷𝑘 𝑁𝐶 and the 𝑉𝑘 𝑁𝐶 terms. The tungsten accumulation 𝑐𝑤(𝑐𝑤 = (𝑛𝑤 /𝑛𝑒 )) is evaluated when the steady state is reached in the simulation.
In previous research [4], we introduced the PHZ pinch and the Er pinch, but investigated theeffect of the PHZ pinch only because the Er pinch model equation was not valid in theexperimental conditions. By introducing the PHZ pinch, the larger accumulation observed
experimentally at higher toroidal rotation was reproduced in the simulation. However, thedependence on toroidal rotation was weaker in the simulations than in the experiments. In thisstudy, we develop a new Er pinch model equation that is valid in the experimental conditions,and also newly introduce the centrifugal force model ([5] for instance). By combining the threeeffects (PHZ pinch, Er pinch and centrifugal force), we will study if the dependence of thetungsten accumulation on the toroidal rotation observed in the experiments is quantitatively
reproduced in the simulation.

[1] T. Nakano, J. Nucl. Mater. 415, S327-S333 (2011)[2] K. Hoshino, et al., Nucl. Fusion 51, 083027 (2011)[3] K. Yamazaki and T. Amano, Nucl. Fusion 32, 633 (1992)[4] Y. Shimizu, et al., Plasma and Fusion Res. 10, 3403062 (2015)[5] K.G. McClements, and R. J. Mckay, Plasma Phys. Control. Fusion 51, 115009 (2009)- 241 -

P2-68Study of Plasma Coherent Structure Transport Dynamics
with Particle-in-Cell Simulation
Hiroki Hasegawaa,b and Seiji Ishiguroa,ba National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 Japanb Department of Fusion Science, SOKENDAI, 322-6 Oroshi-cho, Toki 509-5292 Japan

hasegawa.hiroki@nifs.ac.jp

Recent experiments of magnetic confinement fusion devices have shown the evidence of nondiﬀusive radial transport in the scrape-oﬀ layer (SOL) by intermittent filamentary plasma
coherent structures along the magnetic field line (such structures are called “plasma blob”).
Motivated by the experiments, many theoretical and computational studies about the blob
dynamics have been made [1, 2]. However, most of those studies are based on two-dimensional
reduced fluid models and such kinds of macroscopic models treat kinetic eﬀects (e.g. sheathformation between plasma and divertor plates, velocity diﬀerence between ions and electrons,etc.) as some adjustable parameters under some empirical assumptions. Thus, we have investigated kinetic dynamics on blob transport with a three dimensional electrostatic plasma
particle simulation [3, 4]. In our previous study, we studied kinetic eﬀects on plasma blob
dynamics in the system where the periodic boundary condition is applied in the direction
parallel to the magnetic field line and found that the symmetry breaking in a blob profile occurs by the kinetic eﬀect [5]. In this study, we have applied the particle absorbing boundaries
to the ends in the direction parallel to the magnetic field line and studied such kinetic eﬀects
with the plasma sheath. In the simulation, not only the symmetry breaking shown in theprevious study but also other properties which were not found in the periodic boundary casehave been observed. Furthermore, we will report results of simulation of the “hole” (which
is a filamentary structure with lower density than that of surrounding plasma) propagation.

[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 andreferences 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, 23rd International Toki Conference (2013) P1-28.

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P2-69Intuitive interface for visualizing numerical data using gesturerecognition
Y. Tamura , H. Nakamuraa and S. Fujiwarab
Konan University, Kobe 658-8501, Japana National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb Kyoto Institute of Technology, Kyoto, 616-8354, Japan

Our purpose of this study is to propose an intuitive and natural interface for visualizing
numerical data, for example, numerical simulation result and experimental result. Scientiﬁcvisualization is indispensable technique for recognizing complex numerical data. In nuclear
fusion research, large scale visualization systems using virtual reality technology have beenused [1], and developed visualization interface [2].
On the other hand, an inexpensive virtual reality system, for example, head mounted display(HMD) widely comes to be used to watch 3-dimentional scene. The HMD is useful for scientiﬁc visualization in that the user can look around the scene in 360 degrees and other users
don’t disturb scene; however the user cannot see his/her own hand. Therefore, we introduce
an intuitive gesture recognition system to a visualization system. Automatic extraction ofhuman intuitive movement from natural behavior is diﬃcult task and it is diﬃcult to detectgesture correctly. We introduced singular spectrum transformation for extracting gesture [3].
This method has disadvantage in that computational speed is slow, but can stably detecthuman behavior. At ﬁrst, a key frame from time series of human behavior is extracted using
singular spectrum transformation method. After that, the recognition engine speciﬁes whichpart of the body moves intentionally. Finally, a function assigned to a gesture is executed.
By using this system, the user can naturally control numerical data in 3-dimentional scene.

[1] A. Kageyama, Y. Tamura and T. Sato, Trans. Virtual Reality Societ of Japan 4, 4 (1999)[2] A. Kageyama, Y. Tamura and T. Sato, Progress of Theoretical Physics Supplement 138(2000)[3] J. B. Elsner and A. A. Tsonis, Plenum Press (1996)- 243 -

P2-70Applications of the virtual-reality technology to Plasma Physics
and Fusion Plasmas
H. Ohtania,b and S. Ishiguroa,ba National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japanb The Graduate University for Advanced Studies, 322-6 Oroshi-cho, Toki 509-5292, JapanIn 1997, the National Institute for Fusion Science (NIFS), Japan, installed the CompleXcope
virtual-reality (VR) System based on CAVE system [1] as an instrument for scientiﬁcally analyzing simulation results. We report the recent two developments in visualization researches
at NIFS in this paper.
We have developed the visualization system which displayed in the VR world the threedimensional trajectories of dust particles observed by the stereoscopic fast framing cameras
in Large Helical Device (LHD). The trajectories, equilibrium plasma simulation data anddevice data were displayed integrally in the one VR space [2]. Their three-dimensional
relative positions can be investigated. This system will promote viewer’s understanding.
It is important to consider how to assemble and remove parts in the reactor design. Especially
it is diﬃcult to remove the part through the small hole with rotating it, since the informationof depth is lost. However, by using the VR system, a viewer can watch and move the parts
three-dimensionally. We constructed the system which displayed CAD data by the desktopVR system (Fig.1, right ﬁgure). By using this system, the viewer can grasp and move theparts, and can perform collision detection when assembling.
VR technology is powerful equipment for analyzing simulation and experimental data, anddesigning the reactor. We believe that the buildup in this paper will boost up the researchof the plasma physics and fusion plasmas.
This work is supported by the NIFS Collaboration Research program (NIFS15KNTS037) andthe NIFS/NINS under the project of Formation of International Scientiﬁc Base and Network.
The authors thank Prof. Ohno, Prof. Kageyama and Prof. Tamura for useful discussions.
The HINT2 data, the dust data and CAD data were provided by Dr. Suzuki, Dr. Shoji and
Dr. Tamura of NIFS, respectively.

Figure 1: Left: Integrated VR visualization of dust trajectory data, equilibrium plasma
simulation result and the inner surface of the LHD vacuum vessel. Right: The viewer using
the desktop-VR system. This picture is a composite photograph and only the viewer canwatch the CAD data in the VR world.
[1] C. Cruz-Neira, et al. ACM SIGGRAPH 93 (1993) 135[2] H. Miyachi, et al. IEEE Comp. Soc., (2005) 530- 244 -

P2-71
Direct observation of electron heating by electron Landau damping of
Alfvén ion cyclotron waves with Thosmon scattering system in the tandem
mirror GAMMA 10/PDX
M. Yoshikawa, R. Ikezoe, K. Ohta, X. Wang, J. Kohagura, Y. Shima, M. Chikatsu,
K. Ichimura, M. Sakamoto, M. Ichimura, T. Imai, R. Minami, and Y. Nakashima
Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577, JapanIn GAMMA 10/PDX tandem mirror, the ion cyclotron range of frequency (ICRF) waves
have been used for plasma production, heating and sustaining MHD stability. Maximum ion
temperature has reached 10 keV and the temperature anisotropy (which is defined as thetemperature ratio of perpendicular , T⊥,	  to parallel, T∥, to the magnetic field line) becomes
more than 10. In such a strong temperature anisotropy condition, a slow Alfvén ion cyclotron(AIC) wave was spontaneously excited in the central cell of GAMMA 10/PDX. When the
power of heating ICRF is increased, the diamagnetism increases and the anisotropy becomesstrong. In the experiments, saturation of the diamagnetism is sometimes observed. The
excitation of the AIC modes, which can scatter the hot ions confined in the magnetic mirror
field into loss cone of the velocity space in the central cell, is one of the possible mechanismsof confinement degradation. On the other hand, the parametric decay of the heating ICRF
waves is discussed for the saturation mechanism of the diamagnetism. Low-frequency (LF)magnetic fluctuations with beat frequencies between the heating ICRF waves and discrete
peaks of the AIC modes are clearly detected in GAMMA 10/PDX. The AIC modes haveseveral discrete peaks in the frequency spectrum just below the ion cyclotron frequency.
There is no resonance with ions, and resonant interaction with electrons can be extracted. The
electron acceleration by using waves of ICRF caused by the electron Landau damping. We
directly detect the accelerated electrons at a machine end by making use of the open field
configuration of GAMMA 10/PDX and examining the effects of the AIC mode on electrons.
The AIC mode is normally observed with reflectometers and magnetic probes. The AIC mode
makes an electron temperature heating, and end loss electron flux and temperature increases,
which is measured by the loss electron energy analyzer (LEE).
In this presentation, the direct observation of electron heating by the interaction of AIC
wave is reported. The electron temperatures were directly measured by using a Thomson
scattering system with changing the laser injection time every 10 ms shot-by-shot. In theformer study, the electron temperatures were estimated indirectly by using LEE signals. The
electron temperature increase along with the increase of diamagnetism. However, the peakelectron temperature reached faster than that of diamagnetism. At that time of the electron
temperature peak, the AIC wave spontaneously occurred and scattered the electrons to theparallel directions. From the analysis of the transient feature of the electron temperature,electron Landau damping is attributed to this interaction.

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P2-72
New Proposal on Development of Machine Protection Functionin Conventional Control System Using Information Technologies
T. Yamamoto, E. Yatsuka, T. Hatae, K. Otaa , Y. Hashimotoa, K. Nakamuraa,
T. Sugie, M. Takeuchi, S. Kitazawa, H. Ogawa, Y. Kawano, K. Itami
Japan Atomic Energy Agency, 801-1, Mukouyama, Naka-shi, Ibaraki 311-0193, JapanaJapan Expert Clone Corp., 2-8-6, Shinjuku, Shinjuku-ku, Tokyo 160-0022, Japan

Machine protection is an important function to ensure high reliability and safe operation in alarge experimental facility [1]. ITER edge Thomson scattering system (ETS) is being
developed in Japan. ETS measures electron density and temperature profiles using high energy
laser more than 5 J [2]. Therefore, ETS has to contain interlock system to protect opticalcomponents from damage caused by laser malfunctions. As same as interlock system,
conventional control also should sustain components good conditions. We proposed new
method that templates for program codes of conventional control were generated from interlock
logics using information technologies such as semantic web technique. Information of plantsystem described in digital data allows plant characteristic data to generate other information
by transformation and inference.
The software library EPICS (experimental physics and industrial control system) [3] isrecommended to develop instrumentation and control systems in ITER project. Input / output
controller (IOC) of EPICS manages signal values and provides programming interface tosignals. We proposed a template generator from interlock logics to IOC source codes (that is
database file). The generator describes program codes, which inhibit or permit output to signal,
automatically. This template avoids missing actions for machine protection in the conventional
control.
In addition, we applied the plant description data to risk analysis. HAZOP (hazard andoperability study) is a useful method to indentify hazards and problems in a system [4]. In
HAZOP, guide words are used for investigating risks. To describe plant functionalities andbehaviors using ontology for functional knowledge [5], list of possible risks are extracted fromthe description by applying substitution and/or inference rules.
The views and opinions expressed herein do not necessarily reflect those of the ITER
Organization.

[1] G. Raupp, et al., “Real-time exception handling – Use cases and response requirements”,
Fusion Engineering and Design 87 (2012) 1891-1894.
[2] E. Yatsuka, et al., “Progresses in development of the ITER edge Thomson scatteringsystem”, Journal of Instrumentation 8 (2013) C12001.
[3] EPICS: http://www.aps.anl.gov/epics/.
[4] T. Kletz, “Hazop & Hazan: Identifying and Assessing Process Industry Hazards”
4th edition, Philadelphia, Taylor & Francis (1999).
[5] Y. Kitamura, et al., “A Framework for Systematization of Functional Knowledge
based on Ontological Engineering”, Transactions of the Japanese Society for
Artificial Intelligence 17 (2002) 61-72.

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P2-73
Studies of AM He Atomic Magnetometer for Application in Fusion Plasmas
H. Kiyonaga, Y. Yamada, T. Fukuda
Graduated School of Engineering, Osaka Univ., A1-214, 2-1Yamadaoka, Suita 565-0871, Japan

λ/4

Probe
beam
polarizer

He Discharge
Lamp

λ/4
oscilloscope

polarizer

He Discharge
Absorption Cell
(sensor head)

Pump
beam

Hollow cathodeHe Discharge
Lamp(AM)

Intensity [μV]

P.D.

OFF

Absorption Cell ON

Emission port

Transmittance rate [%]1. Introduction
The detailed measurement of magnetic field is above all fundamental in fusion research, andvarious approaches, as represented by the Faraday rotation and motional stark spectroscopy
techniques, have hitherto been extensively developed.
An atomic magnetometer [1] allows spatially resolved magnetic field measurements and He
atom suggested as a probing target in this work is produced as a result of the D-T fusion reaction.
Neither heating beam nor impurity species are necessary. This study aims at investigating the
proposed AM-AOM i.e. amplitude modulated atomic magnetometer.
2. Design of Amplitude Modulated He atomic magnetometerFig.1 shows a schematic of the AM-AOM system. The pump and probe beams of which
wavelength is 1083nm (He D1 line), are injected into a He discharge cell. The former isamplitude modulated at frequency fpump and circularly polarized for the spin-polarization of the
He atom, whilst the latter is only circularly polarized. When fpump equals the Larmor frequencyfL(=γB), the resonant property is detected in the transmission rate, and magnitude of themagnetic field at the cell is evaluated. Here, γ is the gyromagnetic ratio; 28[GHz/T] for He.
3. Development of the light source
In order to produce the 1083nm emission efficiently, a hollow-cathode type He discharge lamp
shown in Fig.2 has been manufactured in the lab. The power supply of the discharge lamp has
also been designed and built. This power supply can produce high frequency AC voltage withDC voltage offset which is enough high to discharge the lamp. The frequency range of thispower supply includes the aimed frequency 1.4[MHz]. This frequency is equivalent to thegeomagnetic field (50[μT]) measurement, anticipated in the rudimentary experiment. The AM
He D1 emission was thereby successfully obtained.
4. Status of the experimentWith the developed light source, we have performed an exploratory experiment of geomagnetic
field measurement. The magnetic field measured by a Gaussmeter (475DSP) was about
30-60[μT]. Fig.3 shows the transmitted light intensity and the transmission rate between 0.6-1.6[MHz]. The transmission rate seems to have its peak around 0.85 [MHz] which corresponds to30.4[μT]. Accordingly, it is deduced that the result of preliminary experiment indicates thepossible sign of magnetic field. However, quality of the measurement has yet to be improved.
We believe that (1) elimination of stray magnetic field produced by the surrounding electronics
and (2) calibration of the source power would increase the signal integrity.

AM Frequency [MHz]Fig.1 Schematic of AM-AOM.

Fig.2 He discharged lamp.

[1] W.E.Bell and H.L.Bloom, Phy.Rev.Lett.6, 280.

- 247 -

Fig.3 Transmitted light property.

P2-74
Digital Signal Processing
on FPGA for Plasma Diagnostics
Rachana Rajpal, Hitesh Mandaliya, Pramila Gautam
Institute for Plasma Research, Gandhinagar, Gujarat- 382428
rachana@ipr.res.in
Abstract
The overall goal of this experiment is to establish a technology for real time processing of the signals
coming from various sensors via signal conditioning electronics. One such requirement is forinterferometer diagnostics of Aditya and SST-1 Tokamak. The two signals (Reference and Probe) arephase shifted to each other and for feedback application the phase shift between these two needs tobe measured in real time. The existing system of the interferometer system provides two analog
signals in form of sin and cosine trigonometry functions (IQ modulated). One of the method to
calculate phase difference is to find inverse of tan of these two signals. No Von Neumann
Architecture processor provides the trigonometry functions in hardware. Some high-end 32 bitprocessors provide these function but they use ANSI-C library to perform calculations in software
defined manner, which can not assure real time performance. For feedback application real time
response must be deterministic. FPGA can solve this problem by using CORDIC algorithm. The IP(Intellectual Property VHDL Code) is used to find arc tan of these two signals. The hardware isdesign which mainly consists of ADC, DAC, FPGA and single board computer.
The two signals from I and Q interferometer diagnostic is first digitized through high speed pipeline
ADC and then processed in FPGA for implementation of algorithm. The processed phase data is fed
to DAC in real time as well as acquired by host PC. The output of DAC can be used for real time
control. All the commands required by FPGA is issued through PC 104 bus of onboard single board
computer. The control signals for interfacing various components are generated by FPGA. Ethernet
is used as communication medium between host PC and board.
The paper will high light the details of CORDIC algorithm in FPGA, hardware architecture, resultsof real time response and software architecture of system.

- 248 -

P2-75Progress of KSTAR Thomson Scattering Diagnostic System in 2015
J.H. Leea, S.H. Leea, W.R. Leea, K.P.Kima, I.Yamadab, K. Nariharabab

National Fusion Research Institute (NFRI), Korea
National Institute for Fusion Science (NIFS), Japanjhlee@nfri.re.kr

In KSTAR tokamak, Nd:YAG Thomson scattering diagnostic system was installed five years
ago. Until last year, KSTAR Thomson scattering system used proto-type ITER edge Thomson
YAG laser system by rent from JAEA, but proto-type ITER laser was returned to JAEA in thisyear. For this reason, a commercial Nd:YAG laser system newly setup for the KSTAR
Thomson system. This laser’s specification was specially ordered to the Continuum Laser Co.,
that was 2.5J, 20Hz. In this year, to suppress the stray light and electrical noise, we installed along baffle and UPS (Uninterruptible Power Supply: battery) system. The length of long baffle
is ~505mm, and inside of the baffle 25 carbon rings were installed. In case of UPS, an electric
power switched to the UPS before the plasma shot then all of DAQ system was operated under
the UPS system. The electric power supply switch returned to the normal position when plasma
was turned off. In case of upgrade of optical fiber bundle, three optical fiber bundles wereincreased at core Thomson thus we have 12 for core and 15 for edge in 2015. To measure theelectron temperature and density, we calibrated Thomson system by using Rayleigh calibration
and relative calibration by tungsten lamp. During the Rayleigh calibration experiment, we could
measure the Raman signal also. For the alignment, we used the vertical alignment system in the
optical fiber block both core and edge.
This year is transition point for the KSTAR Thomson scattering system because of the greatchanges of the laser system and electric power system. In my poster, I will present the change ofThomson system moreover we will compare the data with other diagnostic system in KSTAR.

- 249 -

P2-76

  
  
  
  
  
  
  
withdrawal  - 250 -

P2-77
Behavior of non-thermal electrons during ECR pre-ionization at Aditya
tokamakS. Purohit, Y. S. Joisa, J. V. Raval, M. B. Chowdhuri, J. Ghosh, B. K. Shukla, R. Manchanda,
N. Ramaiya, R.Tanna, K.A Jadeja, S.B Bhatt, C. N. Gupta, A. Kumar and Aditya teamInstitute for Plasma Research, Bhat, Gandhinagar, 382 428, India

E-mail: pshishir@ipr.res.in

The non-thermal electrons (NTE) in tokamak plasma have a strong effect over the plasma
properties and on the plasma facing components. This establishes an importance of NTE to bestudied in both flux and energy space. The NTE were studied for Aditya tokamak, by Silicon
Drift Detector (SDD) based X-ray spectroscopic diagnostic for the ECR pre-ionization
experiments. The calibrated spectrometer, which is having energy resolution of 134 eV at 5.89
keV, is mounted on the radial mid plane of Aditya tokamak with a 15 µm thickness beryllium
window. The X-ray spectra were acquired in energy range of 1.5 keV – 26.0 keV. The spectra
for ECR pre-ionization shows X-ray line radiation band on top of the bremsstrahlung. It is
primarily the combination of multiple characteristics lines existed between 5 keV to 7 keV andmostly associated with multiple ionized states of iron, tokamak wall material. This is likely due
to the interaction between the non-thermal electrons, generated from ECR, with the stainlesssteel vessel wall of the Aditya Tokamak. The behaviour of NTE was studied as a functionof vertical field and pre-filled pressure. With the increasing vertical field theinteraction weakens which is most likely due to the reduction of particle drift at
higher vertical field. It was also found that the X-ray line radiation strength reduces
dramatically with higher pre-filled pressure at fixed vertical field. This indicates the
reduction of NTE with increasing pre-filled pressure.

- 251 -

P2-78
Fast ion loss associated with RMP in KSTAR
Jun Young Kima, Junghee Kima,b, T. N. Rheeb, Won-Ha Koa,b,
S. W. Yoonb, Y. M. Jeonb, and M. IsobecaKorea University of Science and Technology, Daejeon, Koreab
National Fusion Research Institute, Daejeon, Korea
c
National Institute for Fusion Science, Toki-shi, Japan

Recent numerical simulations and experimental results have shown that the beam-ion loss
in the presence of non-axisymmetric magnetic perturbation (or resonant magnetic
perturbation), aiming at mitigation or suppression of the ELMs, is enhanced at the toroidally
localized positions [1-5] depending on the configuration of the perturbation field.
Fast ion losses measured through 2011 to 2014 KSTAR campaigns with the application of
the RMP have shown the enhanced or suppressed signals exhibited on the FILD (fast-ion loss
detector). It implies that fast ion loss during RMP is non-axisymmetric and sensitive to theradial perturbation direction. In addition, fast ion loss level is elevated at least one order ofmagnitude higher than that in case of conventional L-mode plasma when the RMP-induced
mode-locking, which shows a big drop of toroidal rotation level, is occurred. Recent study
shows the screening effect on the magnetically perturbed plasmas by the toroidal plasma
rotations. [6]. In this sense, it is necessary to investigate the influence of the toroidal rotation
velocity on the fast ion loss level during RMP application in order to understand the relation
between the toroidal rotation screening the perturbation field and the local fast-ion loss
intensity. Electron heating [7] is used to change plasma rotation speed during n=2 RMP(middle-coil only). As a result, both in L- and H-modes, toroidal rotation dependency on the
fast-ion loss is observed clearly. As rotation level increases, fast-ion loss decreases. In cases of
low rotation or locking-plasmas, fast-ion loss level is enhanced significantly.
Additional fast ion loss detectors at I-port (FILD-2, scintillator type, fixed) and D-port(FILD-3, Faraday-cup, radially movable) are ready to observe the asymmetric and localized
loss according to various RMP spectra during 2015 KSTAR experimental campaign.
[1] K. Tani et al., Nucl.Fusion 52, 013012 (2012).
[2] K. Shinohara et al., Nucl. Fusion 52, 094008 (2012).
[3] J. Kim et al., Rev. Sci. Instrum. 83, 10D305 (2012).
[4] M. Garcia-Munoz et al., Nucl. Fusion 53, 123008 (2013).
[5] J.-Y. Kim et al., 41st EPS Conference on Plasma Physics. P4.067.
[6] Y. Q. Liu et al., Plasma Phys. Control. Fusion 54 (2012) 124013[7] Y.J. Shi et al., Nucl. Fusion 53 (2013) 113031- 252 -

P2-79The temperature calibration for IR camera of the IRVB system
including periscope system for KSTAR
Dongcheol Seoa,b and B. J. PetersoncaNational Fusion Research Institute,169-148 Gwahak-ro, Daejeon 34133, Koreab
University of Science and Technology (UST), Daejeon, Korea
c
National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, JapanAn infrared imaging video bolometer (IRVB)[1] provides a 2-dimensional measurement of thetotal radiation power loss from the plasma. The IRVBs have installed in the Korea
Superconducting Tokamak Advanced Research (KSTAR)[2,3] and LHD and JT60-U. The
IRVB diagnostic consists of periscope optics, Phoenix IR camera (Indigo) and a platinum (Pt)foil for KSTAR. For accurate measurement, temperature calibration technique of the IR camera
has been developed. This new temperature calibration technique was applied to IR camera of
IRVB which is installed on KSTAR. In this conference, it will show the details of the newtemperature calibration method and results of the measurement on KSTAR.

[1] B.J. Peterson, et al., Rev. Sci. Instrum., 70 (2000) 3696
[2] M. Kwon, et al., Nucl. Fusion 51 (2011) 094006[3] G. S. Lee, et al., Nucl. Fusion 41 (2001) 1515

- 253 -

P2-80The effect of thermal history on microstructure of Er2O3 coating layer
prepared by MOCVD process
Masaki Tanaka1, Takayuki Shinkawa 1, Yoshimitsu Hishinuma2,
Teruya Tanaka2, Takeo Muroga2, Susumu Ikeno3, Kenji Matsuda4
and Seungwon Lee4
1

Graduate School of Science and Engineering for Education, University of Toyama,
2
National Institute for Fusion Science3
Hokuriku Polytechnic College
4
Graduate School of Science and Engineering for Research, University of Toyama
E-mail: ikenolab@eng.u-toyama.ac.jp

Erbium oxide (Er2O3) was shown to be a high potential candidate for tritium permeation barrier
and electrical insulator coating for advanced breeding blanket systems with liquid metal or
molten-salt types. Recently, Hishinuma et al. succeeded to form homogeneous Er2O3 coating
layer on large interior surface area of metal pipe using Metal Organic Chemical Vapor
Deposition (MOCVD) process. In this work, evaluating the influence of thermal history on
microstructure of Er2O3 coating layer on stainless steel 316 (SUS 316) substrate via MOCVD
process was investigated using SEM and TEM observation and XRD analysis. The
selected-area electron diffraction (SAED) diffraction pattern and the ring-shaped diffraction
pattern of Er2O3 coating are obtained to each SUS substrates, so it revealed that homogeneous
Er2O3 coating had been formed easily on SUS substrate from these diffraction patterns.
The typical SEM images of the surface area on the Er2O3 coating before and after thermal
cycling up to 700℃ in Ar atmosphere. We confirmed that the columnar-like Er2O3 grains werechanged to granular structure by thermal history. the columnar-like Er2O3 grains were promoted
to change to granular structure by thermal history. From the cross-sectional image of TEM
observations, furthermore, the formation of interlayer between Er2O3 coating and SUS substrate
was also confirmed. In this study, comparisons of microstructure on the Er2O3 coating layer
between with and without thermal history were investigated in detail.

Fig. 1 Typical SEM images of the surface area on the Er2O3 coating before and after thermal
cycling up to 700℃ in Ar atmosphere.
1) Y. Hishinuma, T. Tanaka, T. Tanaka, T. Nagasaka,Y. Tasaki, A. Sagara, T. Muroga: FusionEng. Des., 86, (2011), 2530-2533.

- 254 -

P2-81Development of the In-vessel Components of ITER Microfission Chamber
System for Fusion Power Measurements in ITER
M. Ishikawa, K. Takeda, K. Itami
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, JapanA neutron flux monitor is one of the most important systems in ITER because it provides
total neutron source strength and fusion power of ITER. The in-vessel neutron flux monitor
equipped with Microfission Chambers (MFC) [1] is designed by Japan Domestic Agency.
Since the MFC can measure neutrons without attenuation by a thick vacuum vessel (VV), it
has advantages for accurate measurements compared with the other ex-vessel flux monitors.
On the other hand, in-vessel components of the MFC are exposed to the extreme ITER
environment, such as high temperature, high radiation and high electromagnetic (EM) forces.
Therefore, the in-vessel components need to withstand such ITER environment. In this study,
the in-vessel components of the MFC have been developed in order to apply ITER conditions.
ITER will experience about 500 times of baking of the VV during its lifetime of 20 years.
Since the in-vessel components of the MFC cannot be replaced during ITER operation, thosecomponents must withstand for a lot of thermal cycles. In order to demonstrate that thein-vessel components of the MFC would be able to sufficiently withstand the ITER
environment, the thermal cycle tests which simulate baking of the VV of the ITER were
performed. In the thermal cycle tests, severer thermal stresses (+45 °C /h during warm-up,
and -35 °C /h during cool-down) than the ITER regulation (+5 °C /h during warm-up, and-7 °C /h during cool-down) were exposed to the in-vessel components. After the thermal cycle
tests, soundness of the in-vessel components was investigated through the helium leak test
(bombing method), and the measurements of the insulation resistance, the withstand voltage,
and the impedance. As a result, it was verified there were no problems in the performance of
the in-vessel components as a signal transmission system.
The in-vessel components are further exposed to high radiation condition due to highperformance of the D-T plasma. Such high radation generates negative influence on materialsof the in-vessel components, for example, crack of the insulation material of the signal cabledue to swelling. In order to apply to the extreme ITER environment, proven tri-axial mineralinsulated (MI) cables in several fission reactors have been adopted as in-vessel cables. It wasreported that any problem has not been observed in MI cables even if integrated neutron flux
to the MI cable exceeded ~ 1022 nvt in fission reactors. On the other hand, in the case of ITER,
integrated neutron flux to the MI cable has been estimated ~ 1020 nvt for ITER life. However,radiation conditions such as neutron spectrum in ITER is much difference from those offission reactors. Then, in order to show that the MI cable of the MFC system has radiation
hardness during the ITER life, neutronic analysis has been conducted. In the neutronic
analysis, absorbed dose of the MI cable in ITER was calculated and compared with that infission reactor. As a results, absorbed dose of the MI cable in the ITER condition was
calculated 4.5x1010 Gy. On the other hand, that in fission reactors has been evaluated > 1011
Gy if integrated neutron flux is ~ 1022 nvt. This result indicates that MI cable of the MFC canmaintain the radiation hardness for the ITER lifetime.
In this study, it was verified that the in-vessel components of the MFC can be used in theextreme ITER environments for 20 years without any replacements. Detail conditions,
methods and results of the test and analysis as well as detail structure of the in-vessel
components are described together with other tests and analysis in this paper.
[1] M. Ishikawa et. al., Rev. Sci. Instrum. 79 (2008) 10E507.

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P2-82Clarification of arcing mechanism on filament-driven high-current arc
discharge in Cs-seeded negative ion source
M. Ichikawa, M. Yoshida, A. Kojima, M. Hanada, M. Kashiwagi, K. Watanabe, N. Umeda,
H. Tobari, R. Nishikiori, J. Hiratsuka and NB Heating Technology Group.
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japan

Reliable long pulse injections of 500keV 10 MW D0 beam are required on thenegative-ion-based neutral beam injector for JT-60SA. For this purpose, a large negative ion
source has being developed to achieve negative ion current of 22A for 100s. As a result of R&D
efforts, 100 s productions have been achieved with a negative ion current of 15 A. The available
current for the 100 s production is limited by arcing in the cesium-seeded filament-driven
plasma generator. The arcing that is an abnormal localized discharge damages tungsten
filaments during high-power long-pulse arc discharge over 150 kW with cesium seeded. Inorder to realize the JT-60SA negative ion source, the suppression of the arcing based on physics
understandings is one of the critical issues. Therefore, the physical mechanism of the arcing in
the negative ion source has been investigated by employing JT-60 negative ion source shown in
Fig. 1.
Arcing is supposed to be driven by secondary particles produced by injection of accelerated
ions to filaments via plasma sheath. It is expected that the impurities in plasma affect arcing.
Therefore, the analysis focusing on relation between arcing rate and amount of impurities was
performed. The arcing rate defined by the number of the arcing in one pulse duration wasplotted as a function of total operation time in Fig. 2. At first, large intensity of OI emission andhigh arcing rate were measured. After 1500 s of sufficient conditioning time of arc discharges,
the arcing rate was maintained to low level even for the high power arc discharge of > 150 kW .
However, after 13000 s from the beginning of the cesium (Cs) injection for the enhancement of
the negative ion production, by when Cs of about 5 g was seeded, the arcing rate increased again
with Cs amount, although the arc power was kept almost same. This result means amount of
Oxygen and Cs correlated to arcing rate.
From these results, it was seemed that one of the possibility of arcing might be the interaction
between the filament and the plasma with impurity ions because the arcing increased withamount of the impurities.

Figure 1: Layout of experimental setup.

Figure 2: Arcing rate, integral injected amount of
Cs and luminescence intensity of oxygen ion in
operational time.

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P2-83
Fokker-Planck Simulation for Runaway Electron Generationin ITER Disruptions
H. Nuga, A. Matsuyamaa , M. Yagia , A. Fukuyama
Kyoto University, Kyoto-Daigaku-Katsura, Nisikyo-Ward, Kyoto 615-8540, Japana Japan Atomic Energy Agency, 2-166, Omotedate-Obuchi, Rokkasho 039-3212, Japan

It is known that the disruptive plasma termination causes severe problems for tokamaks.
One of the problems in disruptions is the generation of runaway electrons (REs). The highenergy REs may damage the the plasma facing components on impact. Moreover, there is apossibility that the number of REs increases exponentially with plasma current by avalanche
eﬀect. Therefore, the RE generation should be suppressed especially in large scale tokamaks.
In this reason, the accurate estimation of the number of REs generated in disruptions isrequired for the the large scale tokamak design and operation scenario developments.
In large scale tokamaks like ITER, the RE generation process seems to be diﬀerent to thatin present tokamaks. Since the thermal quench started from the higher plasma temperatureencourages the formation of the hot-tail of the electron velocity distribution because of the
low collisionality, the seed REs are easy to generate. The RE multiplication by the avalanche
eﬀect becomes remarkable due to the higher plasma current. The impurity gas injection
which is proposed for the heat load suppression to ITER disruptions makes the current
quench faster. The faster current quench may induce the higher toroidal electric field.
The purpose of this study is the development of the three-dimensional (two-dimension in
momentum space and one in radial direction) Fokker-Planck code, TASK/FP[1], for RE generation and the simulation for ITER parameters. Since the non-Maxwell eﬀect like a hot-taileﬀect will be remarkable in ITER disruptions as noted above, the kinetic simulations are
required. Moreover, two-dimensionality in momentum space is required because the information of the pitch angle distribution of REs is important for the RE mitigation operation.
In our simulations, the RE generation and the evolution of the electric field are calculated
self-consistently.
It will be reported that the RE generation simulation in ITER disruptions, especially thecontribution of the hot-tail eﬀect.
[1] H. Nuga, A. Matsuyama, M. Yagi, and A. Fukuyama, Plasma Fusion Res. 10, 1203006(2015).

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P2-84
Effect of UHR location on non-inductive formation of overdense sphericaltokamak by electron Bernstein wave heating and current drive on LATE
M. Uchida, Y. Nozawa, A. Yoshida, K. Kuroda, H. Tanaka, T. Maekawa
Graduate School of Energy Science, Kyoto University,
Kitashirakawa Oiwakecho, Sakyo-ku, Kyoto 606-8502, Japan

Non-inductive startup of spherical tokamak (ST) without the use of central solenoid is a key
issue for realizing compact, economical fusion ST devices since there is only a restricted spacein the center column of ST [1]. In the Low aspect ratio Torus Experiment (LATE) device,
non-inductive formation of ST by electron Bernstein (EB) waves has been explored. It has beenshown that EB wave heating and current drive is quite effective for formation of initial closed
flux surfaces [2] and rapid plasma current ramp-up [3]. Moreover, recent results shows that anextremely overdense ST plasma can be also started-up with EB wave heating and current drive,
where a plasma current reaches ~12 kA and a line-averaged density reaches more than seventimes the plasma cutoff density [4].
To study and improve the mode conversion rate from incident electromagnetic waves to EB
waves in such extremely overdense plasmas, a multi-chord 70 GHz interferometer system withfive vertical and two horizontal chords has been developed. In this paper we report the density
profiles and their influences on mode conversion to EB waves in such plasmas.
Microwaves at 2.45 GHz from four 20kW magnetrons are used for EB heating and current drive,
which are injected from four midplane launchers. When a microwave power of Pinj = 10 kW is
injected under a weak vertical field of Bv = 20 G, a plasma current is initiated and increases up
to ~ 2 kA, resulting in the formation of closed flux surfaces under the steady external field [2].
The plasma current ramps up with ramps of the microwave power and Bv, and finally reaches Ip~ 10.5 kA. The line-averaged electron density at the final steady phase reaches ne = 5.5 ×1017m-3, which is more than 7 times the plasma cutoff density. In this plasma, midplane density
measurements along a tangency radius of Rt =35.5cm shows that the UHR layer is well inside of
the 2nd ECR layer. This indicates that the EB waves are mode-converted at their first
propagation band (between the fundamental and 2nd ECR layer), after which the EB waves
propagate towards the fundamental ECR layer and heat the bulk electrons as well as current
carrying tail electrons.
On the other hand, if we set the ECR layer at a slightly higher field side, the electron density
becomes much lower. In this case the density does not increase significantly as the current
increases, and the line averaged density at the final steady phase is only one third of the higher
density discharge. In this lower density discharge, the UHR layer is estimated to be just outside
the 2nd ECR layer. This suggests that a large portion of the incident wave power ismode-converted to EB waves at their 2nd propagation band and they are absorbed before the
2nd ECR layer. This may significantly reduce the power coupled to EB waves at the first
propagation band, resulting in decreases in the density and the current at the plasma center.
[1] Y.-K. M. Peng et al, Proc. 23rd Int. Conf. on Fusion Energy (Daejeon, Korea, 2010)[2] T. Yoshinaga et al., Phys. Rev. Lett., 96 (2006) 125005.
[3] M. Uchida et al., Phys. Rev. Lett., 104 (2010) 065001.
[4] M. Uchida et al., Proc. 24th Int. Conf. on Fusion Energy 2012 (San Diego, USA, 2012)
EX/P6-18- 258 -

P2-85Inter-strand conductance measurement of ITER TF coil under epoxy
impregnation
H.Kudoh, T.Yagai, T.Matsuda, H.Shoji, H.Kajitania ,T.Obanab, S.Hamaguchib
Sophia University, 7-1, Kioi-cho, Chiyoda-ku, Tokyo, 102-0094, Japan
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka-shi, Ibaraki, 311-0193, Japanb
National Institute for Fusion Science,322-6 Oroshi-cho, Toki-shi, Gifu, 509-5292, Japana

The toroidal field (TF) coils for ITER are developed by Japan Atomic Energy Agency (JAEA).
TF coils are made of Cable in conduit conductor (CICC). Local current distribution in CICC is
unknown and epoxy impregnation is needed in order to fix strands and measure inter-strand
conductance in CICC. We decomposed CICC by each bundles and measured inter-strand
conductance comparing epoxy impregnated with unimpregnated one. We are also fabricating
devices which can specific strand location and measure local contact resistance in liquid heliumtemperature.

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P2-86Preliminary thermal analysis for ITER Lower Port #02 Integration
Engineering in JADA
T. Maruyama, S. Kitazawa, H. Ogawa, K. Itami
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, Japanmaruyama.toshiyuki@jaea.go.jp

Many diagnostic systems will be installed in the ITER through the diagnostic ports. The
diagnostic lower ports will be used to exchange the diverter cassettes and install diagnostic
systems. Therefore the diagnostic systems in the lower ports have to be designed to be removed
and re-installed. In-vessel components of the lower ports will be mounted on a support structure,so-called a diagnostic rack, to be removed and re-installed with this rack. The diagnostic rack isa critical component in the lower port since the rack will be exposed the harshest environmentand has to withstand those environment such as nuclear heating, electromagnetic force andseismic load. The rack should be designed to withstand these environments.
Japan Domestic Agency (JADA) has responsibility to integrate the diagnostic Lower Port #02(LP#02) engineering, and JADA will also provide the diagnostic rack. The conceptual designof the diagnostic rack is already finished with some issues which still have been left, and thenwe performed the structural assessments to clear some issues for future work. In our previous
work [1], we found that the diagnostic rack needs water cooling to withstand nuclear heatingload. Especially the plasma facing front end of the rack which receives the highest heat loadneeds to be intensively cooled. A fabrication method of the water cooling in the rack is very
important for the whole structure and manufacturability.
In this paper, we focused upon a water cooling structure and assessed its fabrication method
for the most effective manufacturability of the rack. A drilling fabrication method and a joining
pipe method have been expected as realistic fabrication methods, and we assessed the cooling
performance of each method by finite element method (FEM) analysis, ANSYS 15.0. We useda simplified analysis model of the front end of the rack. The model is the simple plate structureand has the same size and shape as the rack, and it provides the cooling path by each fabrication
method. Environmental conditions which have been estimated in the lower ports were used.
The thermal load is in range of 0.1 to 0.2 MW/m3. Obtained results indicate that the drilling
fabrication method needed only a simple and short layout of cooling path, on the other hand thejoining pipe method needed a longer layout than the drilling because the thermal contact of thejoining pipe method is lower (about 15 % of the drilling method). By changing the layout of thejoining pipe a little bit, the joining pipe method gives enough cooling performance. In terms offeasibility, the joining pipe method has major advantages because this method is easier tofabricate and it is possible to greatly reduce the welding areas of vacuum boundary which arerequired the non-destructive inspection in ITER. In conclusion, the joining pipe method ispreferred to the diagnostic rack of LP#02.
[1] S. Kitazawa, T. Maruyama, H. Ogawa, K Itami and N. Casal, Plasma Fusion Res. 10,3402044 (2015).

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P2-87
Reduction Rate of MHD potential energy by toroidal plasma current drive
at tokamak plasma formation with stationary direct current of central
solenoidal coil
Osamu Watanabe
Research Institute for Applied Mechanics,6-1 Kasuga Koen, Kasuga, Hukuoka, 816-8850, JapanWM [%]A sufficient stationary direct current applying in a central solenoidal coil (DCCS) can
decrease the MHD potential energy WM of the tokamak plasma to below a vacuum magnetic
field energy WVCB before a tokamak plasma formation[1]. A vertical magnetic field in inner
region of a magnetic axis of the tokamak plasma is reversed by the toroidal plasma current driveat the tokamak plasma formation. In a case of without the DCCS, the vertical magnetic field in
inner region of the central solenoidal coil (CS region) is also reversed. However, the sufficient
DCCS increases the vertical magnetic field only in the CS region. In this case, the verticalmagnetic field energy in the CS region is decreased in proportion to the toroidal plasma current,
and the magnetic field energy in the CS region is reduced approximately in proportion to the
DCCS. As the result, if the reduced magnetic field energy exceeds a thermal stored energy of
the tokamak plasma, the MHD potential energy is decreased by increase of the toroidal plasma
current. In the case of the tokamak plasma with sufficient DCCS, the magnetic field energy in
the CS region must be regain by get of any energy from somewhere in order to decrease thetoroidal plasma current. Therefore, the decrease of the toroidal plasma current is inhibitedenergetically, and a stability of the tokamak plasma is improved.
At a sustainment of the tokamak plasma by microwave heating[2], law of conservationof energy is satisfied by the MHD potential energy including a radiation from the tokamak
plasma and injection of heating powers which is mainly microwave power injection. However,there is a time evolution of the magnetic field energy in the CS region caused by the toroidal
plasma current perturbation, which is same as Joule heating. The magnetic field energy
transition by the toroidal plasma current perturbation is distributed into the toroidal plasma
current drive, an eddy current generation and current modify of poloidal field coils, temporarily.
Then, its energy state shifts easily to a local
DCCS=1 [kA]
minimum again, with change of the radiation.
WM = (WM−WVCB )/WVCB x 100−2
Therefore, to return to same energy state after the
perturbation, an energy minimization shape is better−3
for tokamak plasmas in same thermal stored energy.
Characteristics of an energy reduction rate
−4
HoleGaussianWM by the toroidal plasma current drive at the
0
0.5
1
n1(n1+n2)
tokamak plasma formation is investigatednumerically. In one instance, Figure 1 shows a Figure 1. Energy reduction rate to toroidal
dependence of the energy reduction rate based on a plasma current profile. n1 and n2 meancurrent components of Gaussian and holecurrent distribution, which is defined by a pressure
distributions, respectively.
function P of a magnetic surface function Ψ asfollows.
1

Ψ−Ψ𝑎𝑥

1

𝑎𝑥 −Ψ𝑠𝑒𝑝𝑃(Ψ) = 2𝑇0 [𝑛1 exp {𝜎 (Ψ

1

Ψ−Ψ𝑎𝑥

2

𝑎𝑥 −Ψ𝑠𝑒𝑝)} + 𝑛2 exp {𝜎 (Ψ2) }] (1)When the current distribution has a hole current component 20 %, the energy reduction
rate at the tokamak plasma formation is maximized in this parameter set.
[1] Watanabe, Osamu. Plasma and Fusion Research 8 (2013): 2401026-2401026.
[2] Takase, Y., et al., Nuclear Fusion 53.6 (2013): 063006.

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P2-88Investigation of pressure profile controllability during plasma current
ramp-up with reduced magnetic flux consumption in JT-60SA
T. Wakatsuki, T. Suzuki, N. Hayashi, J. Shiraishi, S. Ide, H. Kubo and Y. Takasea
Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193, JapanThe University of Tokyo, 5-1-5, Kashiwanoha, Kashiwa 277-8561, Japana

Investigation of the plasma current ramp-up with reduced magnetic flux consumption of thecentral solenoid (CS) is important to relax a strong constraint on the required CS size in thetokamak reactor design. Feasibility of plasma current ramp-up in JT-60SA without additional
CS flux consumption after initial plasma formation has been investigated using an integrated
modeling code suite (TOPICS). In our previous study, we developed a scenario in which theplasma current is ramped-up from 0.6 MA to 2.1 MA without additional CS flux consumption
by overdriving the plasma current using neutral beams (NB) and electron cyclotron (EC) waves[1]. The investigation of the ramp-up scenarios with several prescribed density profiles revealed
that a pressure profile with an H-mode pedestal and a wide internal transport barrier (ITB)
whose foot is located at a large minor radius is required in order to obtain a large bootstrap
current within the MHD stability limit. Although time evolution of the temperature profile wascalculated using CDBM model [2,3], time evolution of the density profile was prescribed.
Therefore, the controllability of the pressure profile by changing the heating and current drive
method might be underestimated.
In this study, we introduce a particle transport model according to experimental results of
JT-60U. In reversed shear plasmas on JT-60U, the effective particle diffusivity in the ITB
region was estimated to be 0.04-0.2 times the ion thermal diffusivity when particle pinch effect
was ignored [4]. Thus, we assume an effective anomalous particle diffusivity Dano = 0.2 x ano,i,where ano,i is an anomalous thermal diffusivity of ions calculated using CDBM model. Particle
transport is calculated by assuming that the particle diffusivity is a sum of neoclassical andanomalous diffusivities with zero particle pinch velocity. The neoclassical diffusivity is needed
for including a strong neoclassical diffusion inside the reversed shear region. On the other hand,the neoclassical diffusivity is negligibly small compared to the anomalous diffusivity in thepositive shear region. Particle sources are NBI and the edge gas puff. The volume averaged
density is feedback controlled by the edge gas puff.
As a result, it is shown that the location and strength of the ITB can be modified by changing theheating, current drive and fueling method using NB and EC. An ITB which is strong enough tooverdrive the plasma current can be obtained within the proposed NB and EC capability at
JT-60SA. Ideal MHD instabilities are investigated using a linear ideal MHD stability analysis
code (MARG2D). We will show controllability of the pressure profile considering the ideal
MHD stability.
This work is supported by JSPS Grant-in-Aid for Research Activity Start-up 26889069.
[1] T. Wakatsuki, et al., Plasma Phys. Control. Fusion 57 (2015) 065005.
[2] M. Honda, et al., Nucl. Fusion 46 (2006) 580.
[3] A. Fukuyama, et al., Plasma Phys. Control. Fusion 37 (1995) 611.
[4] H. Takenaga, et al., Nucl. Fusion 43 (2003) 1235.

- 262 -

P2-89The effects of current density profiles on core toroidal rotation in EAST
plasmasFudi Wang1, Bo Lyu1, Ruiji Hu1,2, Jun Chen1,2,Xianghui Yin1,2, Jia Fu1, Manfred Bitter3,
Kenneth Hill3, John Rice4, Patrick Diamond5, Shigeru Morita6, Yi Yu2, Minyou Ye2, Baonian
Wan1, Yuejiang Shi7,2, and EAST team1

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031,China
School of Nuclear Science and Technology, University of Science and Technology of China, Hefei
230026,China
3
Princeton Plasma Physics Laboratory, MS37-B332, Princeton, NJ 08543-0451,USA
4
Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA,02139,
USA
5
CMTFO and CASS, University of California, San Diego, CA 92904, USA
6
National Institute for Fusion Science, Toki 509-5292, Gifu, Japan
7
Department of Nuclear Engineering, Seoul National University, Seoul, 151-742,Korea
2

fdwang@ipp.ac.cn

The strong correlation between toroidal rotation (uφ) and q (safety factor), li (internal
inductance) has been observed in SOC plasmas, in LHCD at 4.6 GHz plasmas, in combinationof LHCD at 2.45GHz and LHCD at 4.6GHz plasmas, in conjunction of LHCD at 4.6GHz and
ICRF at 35 MHz plasmas, in synergism between LHCD at 2.45GHz, LHCD at 4.6GHz and
ICRF at 35 MHz plasmas, and in association between co-NBI, LHCD at 4.6GHz and RMP
plasmas on EAST. In detail, when toroidal rotation (uφ) is increasing in co-current direction
during above plasmas, the q is also augmenting, but li is decreasing. On the contrary, when uφ isdescending, q is also diminishing, but li is enhancing. The uφ is incremental with li reducing andwith q increasing. The above experimental result is consistent with the prediction of entropy
balance theory, for turbulent plasma, associating with residual stress by comparing rates ofentropy production (destruction) due to thermal relaxation (flow) generation.

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- 264 -