FIP/2-­‐5	  Liquid lithium loop system to solve challenging
technology issues for fusion power plant
M. Ono1, R. Majeski1, M.A. Jaworski1, Y. Hirooka2, R. Kaita1, T.K. Gray3, R. Maingi1,C. H. Skinner1, M. Christenson4, D. N. Ruzic4 and the NSTX Research Team
1 Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543, USA.
2 National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, Japan.
3 Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831, USA.
4 University of Illinois at Urbana-Champaign, IL 61801, USA

M Ono IAEA 2016 FIP/2-5

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Talk	  Outline• Motivation of lithium (Li) in fusion reactor• Divertor heat and particle handling using radiativeliquid lithium divertor (RLLD, Active RLLD)•	  Liquid	  Lithium(LL)	  -­‐loop	  system	  	  General	  descripBon	  T	  ~	  	  0.5	  g/s	  Dust	  removal	  TriBum	  recovery	  methods	  Fusion	  Power	  Plant	  Plasma	  Core	  TriBum	  Inventory	  LL	  and	  triBum	  clean-­‐up	  •	  Summary	  	  Return	  to	  divertor	  RLLD	  

Lithium-­‐Loop	  	  Dust	  ﬁlter	  T	  Extractor	  LL	  ﬂow	  ~	  1	  l	  /s	  M Ono IAEA 2016 FIP/2-5

2Why	  lithium	  (Li)	  for	  fusion	  energy	  development?	  
Some Li related facts and information• Li is the lowest Z = 3 metal – compatible with hot fusion plasmas• Liquid lithium (LL) has a wide operating temperature range of180 -1344°C – expected LL operating temperature of 200 – 400°C (lowlithium vapor pressure) – compatible with fusion reactor environment• Lithium reacts and captures D, T, O – Can provide strong
divertor particle pumping – e.g., low-recycling radiative divertor•	  Lpossible!
ithium	  as	  PFC	  were	  invesBgated	  by	  many	  fusion	  devices	  with	  mostly	  beneﬁcial	  results	  on	  plasmas:	  TFTR,	  NSTX	  and	  LTX	  (PPPL,	  USA),	  FTU	  (ENEA,	  Italy),	  T-­‐11M	  (Trinity,	  RF),	  T-­‐10	  (Kurchatov	  InsBtute,	  RF),	  TJ-­‐II	  (CIEMAT,	  Spain),	  EAST	  (ASIPP,	  China),	  HT-­‐7	  (ASIPP,	  China),	  DIII-­‐D	  (GA,	  USA),	  ISTTOK	  (IPFN,	  Portugal)	  and	  KTM	  (NNC	  RK,	  Kazakhstan).	  	  •	  The	  fourth	  InternaBonal	  Symposium	  on	  Liquid	  Metal	  ApplicaBon	  for	  Fusion	  Device	  (ISLA-­‐2015)	  :	  held	  in	  September	  2015	  at	  Granada,	  Spain.	  	  hbp://www-­‐fusion.ciemat.es/quixplorer/	  M Ono IAEA 2016 FIP/2-5

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Radiative LL Divertor Concept (RLLD, ARLLD)
Possibility of Radiative Divertor with Low Recycling!
M. Ono et al., RLLD - NF 2013,
ARLLD - FE&D 2014

•	  ConvenBonal	  radiaBve	  divertor	  proved	  eﬀecBve	  in	  reducing	  divertor	  heat	  ﬂux	  Divertor	  heat	  &	  parBcle	  ﬂux	  –	  however	  oeen	  caused	  plasma	  conﬁnement	  degradaBon.	  

AcBve	  Li	  InjecBon	  (ARLLD)	  •	  Li	  with	  non-­‐coronal	  radiaBon*	  could	  signiﬁcantly	  reduce	  divertor	  heat	  ﬂux.	  	  Li	  injecBon	  shown	  eﬀecBve	  in	  heat	  ﬂux	  reducBon	  in	  EAST.	  
*S.	  V.	  Mirnov	  et	  al.,	  PPCF	  (2006)	  LL	  Li	  T,	  D	  ImpuriBes	  •	  Li	  can	  provide	  strong	  parBcle	  pumping	  for	  low	  recycling	  divertor.	  
•	  LL	  provides	  renewable	  protecBve	  layer	  Li	  	  non-­‐coronal	  radia5on	  Li	  inﬂux	  (RLLD)	  From	  	  LL-­‐loop	  Thin	  	  ~	  0.2	  mm	  	  LL	  coated	  high	  Z	  divertor	  wall	  To	  	  LL-­‐loop	  for	  high-­‐Z	  substrate. 	  
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A LL-Loop as a part of Fusion Reactor “Ecosystem”
Continuously Removing Dust, T, D, H, He, O, & Other Impurities
D2,	  T2	  ~	  	  0.5	  g/s	  needed	  for	  ~	  3	  GW-­‐fusion	  power	  with	  1%	  burn	  	  	  First	  Wall	  Temperature	  600	  –	  700	  °C	  keeps	  wall	  clean	  Fusion	  Power	  Plant	  Plasma	  Core	  T2 Recycling

T2, D2, He, H2, O2	  Flowing LL cancapture and carryout dust, T, D, H,He, O, & otherimpurities
RadiaBve	  Liquid	  Lithium	  Divertor	  Chamber	  Dust	  ﬁlter	  for	  solid	  waste	  400°C	  T2 ~ 0.5 g/s

TriBum	  Separater	  T	  ~	  	  0.5	  g/s	  D2, He, &Other impurities

Return	  to	  divertor	  0.4%	  T	  ~	  	  2.0	  g/s	  0.5%	  T	  ~	  	  2.5	  g/s	  200°C	  TriBum	  Removal	  (Surface	  Cold	  traps	  Centrifugal	  separator)	  LL	  ﬂow	  ~	  1	  l	  /s	  M Ono IAEA 2016 FIP/2-55

Dust Generation Identified as a Serious Issue for Fusion Reactor*
0.5 g of dust generation per second results in 17 tons per year!

*G.	  Federici	  et	  al.,	  NF	  2001	  •	  With	  0.1%	  dust	  fracBon	  by	  weight	  in	  1	  l/sec	  LL	  ﬂow,	  can	  carry	  away	  17	  tons	  of	  dust	  in	  a	  year!	  	  	  	  •	  While	  LL	  may	  also	  reduce	  the	  dust	  generaBon	  in	  divertor,	  the	  philosophy	  is	  to	  remove	  the	  dust	  as	  much	  as	  possible	  whenever	  generated.	  	  	  “Uchi-­‐mizu”	  in	  Kyoto	  Flowing LL cancapture and carry
Return loop
Return loop
out the “dust”
Flowing	  LL	  ﬁlm	  along	  1	  l	  /	  sec	  over	  ,	  50	  m	  x	  1.5	  m	  divertor	  wall	  for	  ~	  0.2	  mm	  thick	  LL	  ﬁlm	  pumping	  and	  protec5on	  ~	  10	  cm/s	  ﬂow	  speed	  ~	  15	  sec	  transit	  Bme	  	  	  Dust/particle filters are locatedbelow divertor and dust are
carried to filter by gravity
action and if needed jxB and,themoelectric actions.

Dust /
particlefilter

M Ono IAEA 2016 FIP/2-5

Tritium/impurity
removalSurface Cold Trap /Centrifuge
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A LL-Loop as a part of Fusion Reactor “Ecosystem”
Continuously Removing Dust, T, D, H, He, O, & Other Impurities
D2,	  T2	  ~	  	  0.5	  g/s	  needed	  for	  ~	  3	  GW-­‐fusion	  power	  with	  1%	  burn	  	  	  First	  Wall	  Temperature	  600	  –	  700	  °C	  keeps	  wall	  clean	  Fusion	  Power	  Plant	  Plasma	  Core	  T2 Recycling

T2, D2, He, H2, O2	  Flowing LL cancapture and carryout dust, T, D, H,He, O, & otherimpurities
RadiaBve	  Liquid	  Lithium	  Divertor	  Chamber	  Dust	  ﬁlter	  for	  solid	  waste	  400°C	  T2 ~ 0.5 g/s

TriBum	  Separater	  T	  ~	  	  0.5	  g/s	  D2, He, &Other impurities

Return	  to	  divertor	  0.4%	  T	  ~	  	  2.0	  g/s	  0.5%	  T	  ~	  	  2.5	  g/s	  200°C	  TriBum	  Removal	  (Surface	  Cold	  traps	  Centrifugal	  separator)	  LL	  ﬂow	  ~	  1	  l	  /s	  M Ono IAEA 2016 FIP/2-5

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“Cold” Trap at 200 °C Could Remove T, D, H, and O
Cold trap can be regenerated at higher temperatures

Surface	  crystallizaBon	  in	  supersaturate	  soluBon	  1	  ~	  0.7%	  by	  wt	  Super	  saturaBon	  crystallizaBon	  104	  0.1	  ~	  0.08%	  by	  wt	  700	  500	  300	  200	  Temperature	  (°C)	  T	  Solubility	  (ppm)	  T	  Solubility	  (wt	  %)	  105	  103	  “rock	  candy”	  P.F.	  Adams	  JLCM	  1975	  •	  At	  200	  °C,	  hydrogen	  can	  be	  reduced	  toward	  0.08%	  level.	  
•	  Oxygen	  is	  also	  eﬀecBvely	  reduced	  with	  cold	  trap	  (e.g.,	  IFMIF).	  
•	  Nitrogen	  would	  require	  separate	  hot	  nitrogen	  trap	  if	  needed.	  
M Ono IAEA 2016 FIP/2-58

Surface Cold Trap (SCT) system to provide
Large surface area to facilitate LiT crystallization
1	  m3	  SCT	  can	  provide	  100	  m2	  of	  collecBon	  surfaces	  	  Hole	  or	  slit	  LL-­‐in	  LL	  Flowing	  LL	  layer	  ~	  0.1	  mm	  thick	  Flow	  speed	  	  ~	  5	  cm/s	  Total	  LL	  volume	  ~	  10	  l	  	  Or	  1%	  of	  SCT	  volume	  0.1	  mm	  	  LL	  ﬂow	  LiT	  crystal	  layer	  	  LiT	  Diﬀusion	  τ	  ~	  few	  sec	  Advantage	  of	  SCT:	  •	  Reduce	  the	  cold	  trap	  LL	  volume	  and	  T	  inventory.	  	  •	  Facilitate	  LL	  draining	  and	  regeneraBon	  of	  T.	  

LL-­‐out	  LiT	  collecBon	   •	  Provide	  robust	  structural	  design	  regeneraBon	  cycle.	  
plates	  •	  Minimize	  probability	  of	  LL	  clogging.	  	  •	  Facilitate	  the	  CT	  maintenance	  due	  to	  its	  simplicity.	  	  	  M Ono IAEA 2016 FIP/2-5

9Modest heating of SCT can release T from LiT
Release of D2 shown by heating LiD film to ~ 400 - 600 °C
D2	  release	  from	  thin	  LiD	  ﬁlm	  occurred	  at	   T	  recovery	  from	  SCT:	  relaBvely	  lowTemperature	  ~	  400°C	  •	  The	  LL	  ﬂow	  is	  diverted	  to	  another	  SCT.	  
•	  LL	  in	  the	  SCT	  is	  then	  drained	  •	  The	  accumulated	  LiT	  on	  the	  surface	  of	  SCT	  is	  then	  heated	  to	  ~	  400	  –	  600°C	  to	  release	  T	  	  •	  The	  released	  T	  pumped	  out	  for	  recovery.	  
D2,	  T2	  400°C	  T2,	  D2	  A.M.	  Capece,	  JNM	  (2015)	  Similar	  observaBon	  by	  E.	  Oyarzabal	  et	  al.,	  JNM	  (2015)	  To	  triBum	  separator	  	  T2	  D2	  QuanBﬁcaBon	  of	  Li-­‐H	  crystallizaBon	  and	  H-­‐isotope	  release	  are	  being	  invesBgated	  in	  collaboraBon	  with	  University	  of	  Illinois	  –	  U.C.	  
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Centrifugal separation of LiD, LiT also appears feasible
LiT, LiD ~ 1 g/cc is twice as heavy as LL ~ 0.5 g/ccCentrifuge	  Separator	  Operate	  at	  ~	  190°C	  LiD,	  LiT	  LiD,	  LiT	  Processed	  LL	  Enriched	  slurry	  of	  LiD,	  LiT	  ◆◆Separation via cascaded centrifuges– Enriched slurry of LiD, LiT removed
continuously at periphery– Need to prevent LiD, LiT accumulation
Commercial units handle 60,000 liters/hour

M Ono IAEA 2016 FIP/2-5

11Tritium (T) Inventory Control in Fusion Power Plant~ 2.5 kg with 1100 l LL-loop systemT	  ~	  	  0.5	  g/s	  T	  inventory	  in	  LL-­‐loop	  system:	  •	  0.5	  %	  (by	  weight)	  T	  concentraBon	  	  LL	  contains	  ~	  2.5	  g	  of	  T	  /	  l	  •	  LL	  inside	  VV	  may	  contain	  up	  to	  100	  l	  of	  LL	  or	  0.25	  kg	  •	  LL	  in	  LL-­‐loop	  before	  and	  in	  cold	  traps	  (~	  0.5%)	  may	  contain	  500	  l	  of	  LL	  or	  1.25	  kg	  •	  LL	  aeer	  cold	  traps	  (~	  0.4%)	  may	  contain	  500	  l	  of	  LL	  or	  1.0	  kg	  100	  l	  Fusion	  1	  %	  burn	  LL	  ﬂow	  ~	  1	  l	  /s	  0.5%	  T	  ~	  	  2.5	  g/s	  500	  l	  500	  l	  •	  Total	  triBum	  inventory	  in	  RLLD/ALLD	  may	  contains	  2.5	  kg	  of	  T.	  

T	  inventory	  is	  quite	  manageable;	  ~	  2.5	  kg	  with	  ~	  0.5%	  T	  in	  1100	  l	  LL	  system.	  	  Less	  with	  lower	  T%.	  	  	  M Ono IAEA 2016 FIP/2-5

Dust	  ﬁlter	  T	  ~	  	  0.5	  g/s	  T	  Extractor	  Return	  to	  RLLD/ARLLD	  0.4%	  T	  ~	  	  2.0	  g/s	  T	  %	  in	  VV	  0.7	  0.6	  0.5	  0.4	  0.3	  T(kg)	  3.6	  3.05	  2.5	  1.95	  1.4	  12

Li	  –	  T	  Clean-­‐up	  Before	  Shutdown	  In preparation for refurbishment/maintenance• In preparation for shut down, drain LL
into Li reservoir while keeping VV anddivertor chamber at ~ 600 °C to releaseresidual gas and Li.
• Continue pump and collect residual Treleased.
• Collect any residual Li into Li reservoir
operating at ~ 200°C. Li vapor condenses
on coldest surfaces which is the L reservoir.

Maintain	  VV	  and	  divertor	  chamber	  	  at	  ~	  600	  °C	  To	  Pump	  T	  Extractor	  Dust	  Filter	  Li	  reservoir	  at	  200	  °C	  • Shut down once all LL and T are collected.

M Ono IAEA 2016 FIP/2-513

Summary•	  	  LL	  provides	  renewable	  surfaces	  and	  eﬀecBve	  medium	  to	  protect	  high-­‐Z	  divertor	  PFC	  substrates	  from	  steady-­‐state	  and	  transient	  heat/parBcle	  ﬂux	  in	  reactor.	  
•	  	  Lithium	  can	  provide	  an	  eﬀecBve	  means	  of	  reducing	  divertor	  heat	  ﬂux	  through	  non-­‐coronal	  radiaBon	  without	  increasing	  recycling	  (RLLD	  and	  ARLLD)	  •	  	  A	  modest	  LL-­‐loop	  of	  ~	  1	  l	  /	  sec	  could	  provide	  a	  means	  to	  remove	  dust/
impuriBes	  from	  reactor	  divertor	  chamber.	  
•	  	  The	  LL-­‐loop	  system	  could	  provide	  an	  eﬃcient	  way	  to	  remove	  triBum	  and	  control	  triBum	  inventory	  in	  fusion	  reactor.	  
•	  	  Surface	  cold	  trap	  and	  centrifugal	  systems	  are	  proposed	  for	  real	  Bme	  triBum	  removal	  in	  the	  LL-­‐loop	  system.	  
•	  	  NSTX-­‐U	  will	  conduct	  basic	  RLLD/ARLLD	  experiments	  in	  H-­‐mode	  plasmas.	  
•	  	  LL-­‐loop	  and	  T	  extracBon	  R&D	  cam	  be	  performed	  with	  relaBvely	  modest	  investment.	  
14
M Ono IAEA 2016 FIP/2-5	  In NSTX, Li Improved H-mode Performance

Modest amount of Li ~ 0.3 g improved plasma performance• Broader pressure and current
profile for improved MHD stability(H. Kugel, PoP 2008)• Electron energy confinement
improvement by ~ 20 - 30 % (R.
Maingi, PRL 2011 )• Reduction in H-mode power
threshold by ~ 20 - 30%. (S. Kaye, NF

S.	  Kaye,	  NF	  2014	  2011)• ELM stabilization (D. Mansfield, JNM
2009)• Very low Li core dilution observed< 0.1% (M. Podesta, 2012)
M Ono IAEA 2016 FIP/2-5

R.	  Maingi,	  PRL	  2009	  	  15

NSTX Liquid Lithium Divertor (LLD) Experiment

Modest amount of lithium protected LLD metal surfaces• Thin sub-millimeter Li coating protected the LLD surface
even with the strike-point directly on it.
• No sign of LLD surface material (moly) in the plasma.
Before	  NSTX LLD ExperimentAeer	  LLD	  LLD	  0.2	  mm	  thick	  Li	  layer	  protected	  the	  LLD	  surface	  	  M Ono IAEA 2016 FIP/2-5

H.	  Kugel	  FE&D	  2012	  16

Double walled for safety and tritium recovery
Also provide eﬃcient thermal barrier
LL	  Double	  Walled	  System	  Vacuum	  Flowing	  LL	  Vacuum	  Double	  walled	  system	  for	  the	  LL	  loop	  system:	  1. Provide	  extra	  layer	  of	  safety	  protec5on.	  	  Should	  be	  able	  to	  detect	  the	  LL	  leak	  at	  early	  stage.	  
2. 	  Vacuum	  provides	  good	  thermal	  insula5on.	  
3. 	  Any	  tri5um	  migra5on	  into	  vacuum	  can	  be	  collected	  by	  the	  double	  walled	  system,	  essen5ally	  elimina5ng	  tri5um	  loss	  from	  the	  system.	  

M Ono IAEA 2016 FIP/2-5

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NSTX-U Facility Came On Line This year
To demonstrate fully sustained high beta plasmas
2nd	  NBI	  NSTX-­‐U	  M Ono IAEA 2016 FIP/2-5

1st	  NBI	  HHFW	  System	  18

NSTX-U Mission Elements:ITER	  • Explore unique ST parameter
regimes to advance predictive
capability - for ITER and beyond• Develop solutions for plasmamaterial interface (PMI)

Liquid metals / Lithium Snowflake/X• Advance ST as Fusion Nuclear
Science Facility and Pilot Plant

M Ono IAEA 2016 FIP/2-5

ST-FNSF /Pilot-Plant

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Lithium	  applicaBons	  will	  be	  invesBgated	  on	  NSTX-­‐U	  Lithium	  Evaporators,	  Granule	  Injector	  Lithium Evaporator (LITERs)

LITERsGranule injector (GI) for ELM pacing
Successfully tested
on EAST and DIII-D
Granules: Li, B4C, C
f ~ up to 500 Hz
GranularReservoir•	  	  LITERs	  ﬁling	  set	  up	  in	  high	  bay	  south	  of	  NSTX-­‐U	  Test	  Cell.	  
•	  	  Argon	  purge	  system	  implemented	  for	  Li	  safety,	  used	  for	  the	  quick	  argon	  vent.	  

Rotating Impeller• GI is now
operational.

•	  	  Ready	  to	  support	  the	  run.	  	  	  GI	  installed	  on	  NSTX-­‐U	  M Ono IAEA 2016 FIP/2-520

Upward Lithium Evaporator R&D Being Performed
Fast evaporation with much less lithium•	  WAF	  completed	  for	  ULITER	  prototype	  development	  •	  Based	  on	  resisBve	  heaBng	  of	  lithium	  in	  porous	  stainless	  steel	  medium	  for	  fast	  evaporaBon	  •	  Prototype	  can	  be	  used	  on	  NSTX-­‐U	  if	  installed	  using	  power	  supply	  with	  remote	  control	  •	  EvaluaBon	  of	  porous	  medium	  to	  absorb	  liquid	  lithium	  in	  progress	  	  Li21”

M Ono IAEA 2016 FIP/2-5

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