Gianfranco Federici

1
Technology Development Status and Prospects for
Divertor PFCs in DEMO

• Gianfranco Federici
• F4E, Barcelona, Spain

Meeting on DEMO Garching, Germany September
29-30, 2009
2
Outline

• Power exhaust in ITER (steady state and transient
  loads)
• ITER design and PFC choices
• Main technology achievements for ITER PFCs
• Remarks on the problem of power exhaust in DEMO
• Status and Development issues of DEMO divertors
• Summary

3
Power Exhaust in ITER Burning plasma with
dominant a heating ? Q10
Pedge radiation

• PF400 MW?Pa80 MW, Padd70MW
• Pout Pwall gt 100 MW Awall 1000 m2
• Divertor ?Awalleff3 m2
• gtqwall 30 MW/m2 (too high!)
• gtgttechnology limit (qwallmax-tech 10 MW/m2) ?
  Need to reduce Pwall by radiation in the
  periphery
• Increase of edge radiation achieved with divertor

Pdiverrtor radiation
Divertor ? edge ionization ? edge radiation ?
Pwall
nzcore ? Pradcore ? Pfus nzedge ? Pradedge ?
Pwall
Radiation must occur in areas of plasma that do
not produce fusion (T ltlt 10 keV)
4
Divertor Power Exhaust in ITER

• Divertor minimizes the influence of plasma wall
  interactions on the main plasma, but concentrates
  the heat and particle flux onto a relatively
  small area.
• Must reduce q-peakPCF 10 MW/m2 for technology
  limits.

• Large SOL radiation ? Divertor ? Prad0.6 PSOL
• Low angle of incidence on target3? q-PCF5
  MW/m2

Detached Plasmas (high Gdiv) high Tdivsurf
Erosion of a carbon targets. Results of MC code
simulations (e.g., ERO) 3D transport of impurities

• Maximum of net-erosion in the SOL ?Outer 1.8
  nm/s (5 mm in 7000 pulses)
• But strong concern associated with trapping of
  tritium in the deposits.

Kirschner, PSI 2006
5
(unavoidable) Material Damage due to Transient
Thermal Loads
G. Federici, et al., Plasma Phys. and Contr. Fus.
45 (2003) 1523
Qgt 2.2 MJ/m2
1 pulse
2 pulse
5 pulse
W
C

• Need effective and reliable mitigation methods
  for type-I ELMs and disruptions.
• Implementation of RPM coils in ITER very
  difficult.

6
ITER Divertor High Heat Flux Components Design
CFC Monoblock

• Modification of the surface by LASER
• Casting of copper in the presence of silicion and
  titanium as activating elements

• CuCrZr tube with CFC or W blocks joined by
  hipping
• and hot radial pressing
• Critical areas
• Quality of CuCrZr
• CFC homogeneity of properties for production of
  large batches
• Diagnostics to prevent approaching of burn-out
  conditions
• Swirl tape attachment
• Defect detection and acceptance criteria
• Behavior under n-irradiation
• Minimise manufacturing costs
• Development of high heat flux W-clad targets
• High heat flux testing
• (100 of the first 10 then 10 of the remaining
  90 - correlate with high temp. water
  thermography)

• Best testing results achieved so far
• CFC and W 3000 cycles at 10 MW/m2
• CFC 2000 cycles at 20 MW/m2
• W 2000 cycles at 15 MW/m2

7
ITER Divertor High Heat Flux Components
Pre-qualification Achievements
Full monoblock version

• Prequalification process
• Each Party supply at least two medium-scale QPs
• The Party is considered qualified if
• two of the delivered QPs meet all the prescribed
  acceptance criteria
• At least one of the delivered QPs withstands the
  high heat flux qualification tests

Companies involved Plansee SE (A) and Ansaldo
(I) Options full monoblock and mono-flat tile
(superseded) CFC material SNECMA NB41
ANSALDO
Testing Plan for pre-qualification (fulfilled by
EU)
Mono-flat tile version

• W armoured part
• Thermal Mapping at 1 MW/m2
• 1000 cycles at 3 MW/m2
• Thermal Mapping at 1 MW/m2
• 1000 cycles at 5 MW/m2
• Thermal Mapping at 1 MW/m2

• CFC armoured part
• Thermal Mapping at 5 MW/m2
• 1000 cycles at 10 MW/m2
• Thermal Mapping at 5 MW/m2
• 1000 cycles at 20 MW/m2
• Thermal Mapping at 5 MW/m2

8
Still Unresolved ITER PFC Issues

• Resolving issues of power handling and plasma
  wall interactions is a prerequisite for ITER
  success and for demonstrating the viability of
  fusion as an energy source

• Despite remarkable advances there are areas that
  require further urgent work. They are
• Suppression/mitigation of disruptions and ELMs
• Confirm material options for divertor (C vs. W)
• Safety implications of some of erosion products,
  diagnostics and control
• Improve design and confirm remote maintainability
  of internal components, especially blanket and
  first wall
• Confirm fabricability and Improve reliability of
  adopted technologies

Courtesy G. Matthews, JET
9
Divertor Heat Removal in DEMO

• DEMO will need to make 2500 MW of fusion power
  in a device size of ITER. It will present
  unprecedented divertor challenges.
• It will almost certainly have several times
  higher upstream parallel heat fluxes than ITER.
• There will be extremely high heat flux at the
  divertor of DEMO.
• ITER has 120MW of heating power Ph and assumed
  core radiation fraction of 50.
• Ph/R which is 5 times higher (Ph 500MW, R 6m),
  so with a core radiation fraction of 50, the
  upstream parallel heat flux should be expected to
  be in the range of 5 times higher.
• To attain a // heat flux to ITER, the core
  radiation fraction would have to be 90.
• Analysis from the viewpoint of confinement, He
  exhaust, thermal stability, and disruptivity
  indicates that such high core radiation fractions
  are very unlikely to be tolerable with acceptable
  reactor performance.
• Innovative solutions may be needed. This may
  involve new magnetic geometries such as Super-X
  divertor and or innovative materials including
  liquid surfaces.

10
NEED a solution for gtgt higher upstream q
Problems at high divertor power and low density
The fusion gain in steady state maximizes at low
density for constant ?N. The limitation on
reducing the density is set by the impact on the
divertor.

• Divertor plate heat flux
• With the same core radiation fraction as ITER,
  and 5 times greater P/R, Reactor q is five
  times greater
• Technological limits of He- W targets 10 MW/m2
  , perhaps less at much higher n-fluence
• Helium pumping
• In simulations, degrades very rapidly with power
  and lower density
• Plasma erosion of the plate/plasma impurities
• High plasma plate Te/low ne greatly increases
  sputtering/reduces prompt redeposition
• Divertor survival of disruptions/ELMs
• Several times larger than ITER
• A divertor which can tolerate much larger
  ELM/disruptions is highly desirable
• Neutron damage to divertor
• Divertor in DEMO exposed to over an order of
  magnitude higher n-fluence than ITER
• Severe degradation of bulk material properties
  is expected ductility (hardening, He
  embrittlement, thermal conductivity, swelling)

11
He-cooled W Divertor Concepts (i) A coincise
review of status / achievements

• To achieve peak heat load of 10 MW/m²
• Reduce the thermal stresses (small modules)
• Short heat conduction paths from plasma-facing to
  cooling surface
• High heat transfer coefficients with as small
  flow rate and pumping power as possible
• To survive 100 1000 thermal cycles
• Cooling by helium jets (10MPa, 600?C) impinging
  onto the heated surface of the thimble.

HEMS (FZK) Backup (He-cooled modular divertor
with integrated slot array)
HEMJ Reference (FZK) (He-cooled modular
divertor with multiple Jet cooling)
HETS Concept (ENEA)
12
He-cooled W Divertor Concepts (ii) A coincise
review of status / achievements
Cooling technology JET impingement in HEMJ

• Small finger modules favourable for stress
  reduction.
• Small tile made of W brazed to a thimble made of
  W-1La2O3.
• The main finger units are connected to the main
  structure of ODS Eurofer Steel by means of Cu
  casting with mechanical interlock

He at 10 MPa, 600C, 5-15 g/s
Typically applied in gas turbine, airplane
engines.

• Work carried out to date consists
• design, analyses, demonstration of fabrication
  technology,
• and experimental design verification
• HHFT in a combined E-beam and He-loop facility at
  Efremov
• Three HHFT campaigns
• 2006 (6 mock-ups 5 HEMJ 1 Slot type)
• 2007 (optimised mock-up geometries)
• 2008 10 mock-ups different material, different
  brazing

13
Early HHF Tests (2006)

• 6 mock-ups (5 from HEMJ and one slot type) were
  tested
• 4 additional were fabricated but cracking
  occurred during brazing of the W tiles with the
  thimble
• Mockups tested 5-13 MW/m2 with He at Pin10 MPa,
  Tin 600C and g5-15 g/s
• Comparison of measured and predicted surface
  temperature within 5
• Detected damages
• thimble cracking with or without gas leakage,
• tile cracking with and without cracks propagation
  through the thimble,
• losing of thermal contact in tile/thimble
  interface
• Main reason of failure related to thimble cracking

14
Early HHF Tests (2007)

• 10 mock-ups were fabricated 6 of them where HHF
  testable.
• Various technical improvements were made which
  led to noticeable improvement in perfromance and
  resistance against thermal loadings.
• The most successfully tested mock-up achieved 10
  MW/m2 for more than 100 cycles without any
  damage.
• But high temeparture increase and gas leak during
  the heat load cycles were detected in many
  mock-ups, but no damage after experiment
  termination.
• It became also clearer that the high failure rate
  of mock-ups were
• Base material quality
• Manufacturing quality (W turning, JET holes
  drilling, EDM of W surfaces, etc.)
• Overheating of the tile/thimble brazed joint
  leading to detachment, and
• Induced high thermal stresses

15
Results of Recent HHF Tests

• Clear progress was achieved in the latest 3rd HHF
  test series. .
• 11 1-finger mock-ups were tested, 2 mock-ups were
  tested for the second time and 9 mock-ups were
  tested once.
• Sensitivity tests to investigate

• Improved machining mechanical vs.
  electrochemical grinding.
• Orientation of material tile structure
  horizontal vs. vertical.
• Thermal loading ramp soft vs. sharp.
• No major difference observed.
• Temperature of brazing low (1050oC, high
  1300oC)
• First tests with high temp brazing showed
  delamination

Next HHF 1-finger tests will focus on
demonstrating 10MW/m2 and 1000 cycles.
16
DEMO Watercooled Divertor
Objectives Use of ITER consolidated solution (W
monobl.) Low activation material for heat
sink Water cooled concept is based on study
analysis Required heat flux 15 MW/m2 Coolant
p40bars,T150C ,v12 m/s
Enhanced concept (CEA)

• Compliance layer for differential dilatation
  between W and EUROFER (Papyex)
• Thermal barrier as flux repartition in the
  structural material (Pyrolithic graphite)

Basic concept

• High temperature water cooled divertor (325C
  coolant fluid), designed to sustain 15 MW/m2
• Maximal flux in structural material without
  thermal barrier is limited to 13 MW/m2 instead of
  18 MW/m2
• Structural integrity is demonstrated by a Finite
  Element thermo-mechanical analysis i.e. no HHF
  testing
• The choice of graphite have to be re-analysed
  considering the material behaviour under
  irradiation

Enhanced concept (CEA)
17
Innovation is Needed

• Compatibility of radiative divertors with
  optimised core confinement, but ..
• Precision divertor alignment to maximise use of
  flux expansion, but not enough
• Innovative divertor configurations for increased
  flux expansion, Super X
• Key idea q gt 10 limit gt only knob is
  increased Rdiv
• Liquid metal target may be attractive but
  experience is very limited

18
Hot Tungsten and Liquid Metals introduce New
Issues of Impurity Influx
19
Prospects/ Summary

• ITER
• During the last two decades Europe has made
  significant contributions to the development of
  technologies for ITER divertor PFCs.
• Underlying technologies have been validated
  mainly by manufacturing and HHF testing at or
  above ITER requirements a large of small-
  medium- and large-scale mock-ups.
• A complementary, vigorous RD program is also
  being launched in Europe to address all remaining
  physics and technology issues associated with the
  use of W at the strike points, in view of
  allowing a decision to start with a full W
  divertor in ITER from day one.
• DEMO
• In addition, technical solutions are being
  explored for devices beyond ITER (e.g., DEMO).
• Most of ongoing effort in Europe described here,
  centres to the development of He-cooled tungsten
  divertor concepts that (at FZK) can withstand
  only up to 10 MW/m2.
• Conventional understanding and simulations
  indicate that without relying on unrealistic
  assumptions of core and divertor radiation
  fractions (gt90), which are very unlikely to be
  tolerable with acceptable reactor performance,
  such parameters could be outside the envelope of
  the conventional divertor solutions presented
  here.
• We need to be seriously innovative
• A re-orientation of the programme and innovative
  divertor solutions that can handle several times
  higher upstream parallel heat flux than ITER
  might be needed.

20

• Extra slides

21
ITER Plasma Facing MaterialsArguably One of Most
Contentious Issues

• CFC
• Better able to handle transient heat loads (ELMs)
  and disruptions
• Demonstrated tritium retention problem
• Procurement (long and risky) high costs
• PFC manufacturing difficulties leading to
  significant rejection rate
• PFC qualification (complex acceptance criteria)
• W
• Very high melting temperature
• Low erosion in a reactor
• Still need to improve thermal fatigue performance
  
• Melt layer stability in tokamak conditions may
  be very poor
• Melting/resolidification induced irregularities
  with consequent melting/evaporation increase due
  to particle glancing incidence

• Be
• Good oxygen getter
• Low Z lowest plasma radiation
• May melt even as a result of disruption
  mitigation

• Use CFC in divertor for H/D operation,
• Then W for DT operations

Mix-material behavior in tokamaks is uncertain