G. Federici

1
Materials Selection and Engineering Design of
ITER PFCs
G. Federici ITER JWS Garching

• Outline
• Highlights of PFC design material selection
• Erosion during transient heat loads
• Remaining open questions
• Summary

2nd SOL and Divertor Physics ITPA Meeting, Ioffe
Institute, St. Petersburg - July 14-17, 2003
2
ITER
Rationale for material selectionPresent choice
3 different materials
Advantages Drawbacks
Be Good O2 gettering capability tokamak practice (mainly JET) low Z repairability by plasma-spraying established joining technology low tritium inventory. Inadequate for the divertor (Tmlt1300C). Be dust on hot surfaces -gt H production due to reactivity with steam during accidents. Compatibility with Type I ELM loads disruptions or mitigated disruptions.
C Good power handling, thermal shock and thermal fatigue resistance does not melt (but sublimes) low radiative power losses with influx to plasma due to low Z well established joining technology broad tokamak operation experience. Chemical erosion --gt erosion lifetime (few thousand pulses). Tritium codeposition --gt needs adequate recovery methods.
W low physical sputtering yield and high sputtering threshold energy no chemical sputtering in H-plasma low tritium inventory repairability by plasma-spraying well established joining technology. Plasma compatibility--gtfavourable operation experience from ASDEX-Upgrade, dust radiological hazard. Near strike points, its operational lifetime, has still uncertainties due to melt layer loss during disruptions and large ELMs.
3
ITER
Divertor design loads/lifetime

• Total power Q ? 150 MW
• Nominal surface heat loads q 10 MW/m2, ?
  400-500 sec, N 3000 cycles (1st div.)
• Transient heat load q 20 MW/m2, ? 10 sec, N
  300 cycles (i.e., 10)
• Disruption heat load Q 10-100 MJ/m2, ?
  0.1?10 ms, N 300 event (i.e., 10)
• Ion flux parameters J 1021-1024 m-2s-1, E
  10-100 eV
• Neutron load J(Egt0.1MeV)?1018 m-2s-1, D 0.1
  dpa
• Other factors electromagnetic loads (P? 4 MPa),
  H environment, etc.

W/m2
Code simulation of surface heat flux on divertor
4
ITER
Design assembly/maintenance

• The PFCs and cassette body combine to provide
  adequate shielding to the vessel and the coils.
• The divertor comprises 54 cassettes installed in
  ITER via 3 equi-spaced handling ports.
• Each cassette consists of a cassette body onto
  which are mounted 3 PFCs.
• These PFCs can be exchanged in hot cell in order
  to refurbish or to change the geometry of the
  divertor.
• Several complete exchanges are foreseen during
  the life of ITER.

• The cassettes are accurately positioned in the
  vessel such that each PFCs is aligned within 2mm
  with respect to the PFCs on adjacent cassettes.

5
Initial operation strategy Current strategy is to
initially install CFC on the targets
ITER

• Tungsten has still uncertainties due to melt
  layer loss during disruptions and large ELMs.
• Maintain option to switch to an all-W divertor,
  prior or during T-operation. This option will be
  considered if
• we do not succeed in mitigating the effects of T
  co-deposition (need to be to be determined during
  D-phase)
• Design mitigation/temperature tailoring.
• Mixed-material effects (gtneed to use existing
  tokamaks).
• we fail in developing reliable and effective
  techniques of in-situ tritium removal, which need
  to be demonstrated/tested in tokamaks.
• there is substantial progress in mitigating
  heat loads during disruptions and ELMs.

W
CFC

• Urgent need for development PMI diagnostics to be
  tested validated in existing tokamaks and during
  early operation with D in ITER.

6
ITER
First-wall design loads/lifetime

• Total power Q ? 690 MW
• Surface heat loads q 0.25 MW/m2 (avg.) 0.5 max.
• ?
  400-500 sec, N 30000 cycles
• Disruption heat load None.
• VDEs q 60 MW/m2, ? 300 ms
• N300 cycles (1),
• Neutron load 0.56 MW/m2 (avg.) / 0.78 (max)
• D 1 dpa.

• 412 blanket modules attached to the vessel.
• 4.5 t/module RH constraint.
• steel shielding block
• separate first wall panels.

7
ITER
Design assembly/maintenanceSeparate first-wall
to minimise the operational waste

• Each blanket module is connected to a permanent
  water cooling manifold by two pipes.
• The FW part repairable and/or replaceable in
  hot-cell.
• The modules are maintained by a special remotely
  driven in-vessel transporter inserted through the
  equatorial port.

8
ITER
Separate first-wall conceptsThe FW has 4 or 6
separate panels depending on the option chosen
for FW attachment.

• Option A attachment with bolts and small shear
  ribs to support EM loads and to prevent sliding
  due to thermal expansion.
• Option B central beam attachment connected to
  the shield block on the rear side.

9
ITER
Remote handling classificationbased on need for
scheduled and unscheduled maintenance and
modifications, likelihood for maintenance, and
impact on operations and availability

• Divertor is Class 1
• Requires scheduled maintenance or replacement.
• Component design and RH equipment and procedures,
  optimised to ensure task completion within a
  minimum time.
• Feasibility of maintenance tasks demonstrated
  with RD during EDA.
• Demonstration using real components during
  initial assembly prior to active phase of
  operation is highly desirable.
• Blanket is Class 2
• Do not require scheduled but likely unscheduled
  or very infrequent maintenance.
• Components are designed for full remote repair or
  replacement but minimisation of repair is
  subordinate to consideration on nuclear
  performance and reliability.
• Feasibility of maintenance tasks partially
  demonstrated with RD during EDA.
• Demonstration using real components during
  initial assembly prior to active phase of
  operation is desirable.

10
ITER
Maintenance time estimatesIn-vessel components
are removed from the VV by 4 equatorial ports, 3
divertor ports and (?) upper ports (EC and
diagnostics).

• Divertor cassette refurbishment
• 18 cassettes --gt 1 RH port gt2 months (7 working
  day/week, 2 (8 hrs) wk shifts/day, 1 cask
  transporter (8 hrs/shift/day).
• 3 RH ports in series gt 6 months
• Replacement of 1 single faulty cassette 2
  months!!
• Blanket maintenance
• Replacement of some shield modules is likely due
  to local damage.
• Replacement of the full first wall is presently
  not anticipated, but should be feasible.
• 1 blanket module 25 days
• 1 toroidal row 32-151 days
• All blanket modules 276-916 days
• Depends on of deployed IVT.

• Current design policy small number of spare
  parts.

11
Critical PFC design/operation issues
ITER

• Heat loads and erosion during type I ELMs
• Ongoing vigorous ITPA effort, EU PWI task force
• Heat loads and erosion during thermal quench
  disruptions and VDEs
• Ongoing vigorous ITPA effort, EU PWI task force
• Hydrocarbon transport/T codeposition in remote
  areas and removal
• Surveys in tokamaks (EU PWI task force)
• Identify source and sinks.
• Deposition patterns and dependence on operation
  parameters.
• Composition of exhaust.
• Laboratory simulations (EFDA, EU PWI task force)
• Sticking of radicals.
• Effects of temperature, H/C ratios, etc.
• Mixing of materials (C/Be)

12
Heat loads and erosion during type I
ELMs Tolerable ELMs in ITER set by materials
TplateELM lt physical limits (evap., melting).
ITER
Temp. excursions during ELMsgt no ratcheting!!!

• Critical parameters are
• (1) energy loss from pedestal,
• (2) fraction reaching the divertor,
• (3) wetted area,
• (4) duration/shape of ELM heat pulse.
• Knowledge of these quantities still uncertain.

Triang. ELMs, 0.3 ms, 1 MJ/m2 (1) BOL 20 mm CFC,
10 mm W. (2) EOL 2 mm CFC, 2 mm W.
Only near-surface 400µm gt Steep Temp. gradients
13
Tolerable ELM size A large number of ELMs gt1
MJ/m2 cannot be tolerated
ITER
ITER
Armour material For W assume 50 melt loss. CFC W
Baseline divertor design All ELMs w. same parameters Statistical evaluation 9 MJ 4 MJ 10 MJ 3 MJ
More inclined divertor target All ELMs w. same parameters Statistical evaluation 15 MJ 6 MJ 17 MJ 5 MJ

• In ITER Wped 100 MJ, nped 8x1019 m-3,
• Tped3.5 keV, dw10 cm gtlow collisionality
  (nped 0.04)
• Scaling to ITER - DWELM/Wped
• nped 15-20 --gt 15-20 MJ.
• t// 10-15 --gt 10-15MJ.
• nped/nGW 4-5 --gt 4 -5 MJ.
• 50-70 of energy is deposited in divertor
• Wetted area 4.5 -9 m2 with modest (50)
  broadening
• impact time gt t// (240 µs) tIR/ t// 1.5-3.1.

• CFC and W show similar ELM erosion lifetime.
  Lifetime for W depends on melt layer loss.
• Erosion lifetime shorter if one use a statistical
  evaluation of ELM parameters.
• More inclined divertor target performs better.
• Compatibility of inter-ELM plasmas with irregular
  W surface remains an issue.
• Additional macroscopic erosion mechanisms due to
  high frequency pulsing and high temperature
  excursions localised in the near-surface (lt400µm).

14
A non-negligible fraction of the ELM energy from
the main plasma reaches the main chamber wall
ITER
DWELMdiv 50 80 of DWELMdia
Despite narrow l_at_ELM DWELMdiv/DWELM 0.6

• Where does the Rest of DWELM go ?
• Toroidal Asymmetries (probably No)
• Main Chamber (probably Yes)
• Transiently enhanced PRADELM (probably No)

JET Type I ELM
15
Effects of type I ELMs at the main chamber wall
ITER
Expect interaction with protruding surfaces (1
m2) gt tolerable only 1-2 MJ depending on
duration (if 1 ms no melting of W)
16
Thermal Quench Disruptions Only a fraction of the
energy reaches the divertor and is distributed
rather uniformly.
ITER
Evolution of the surface temperature near and
far from strike points in a DWth5.6 MJ
G. Matthews et al., 19th IAEA FEC 2002, Lyon To
appear in Nucl. Fusion.

• If the JET results extrapolate to ITER then
  disruptions would not damage a W target.
• However, it is not known at the moment where and
  by what processes the missing and thermal an
  magnetic energies are deposited in the main
  chamber.
• If this energy deposition is not sufficiently
  uniform, then additional damage to main chamber
  components might be expected.

17
ITER thermal quench specifications Need to be
revisited on the basis of the new findings
(ongoing)
ITER
Case 1
Case 2

• If energy deposited in the divertor during
  disruptions is lt 40 of the thermal energy with a
  broadening of the order 20-30 times, energy
  density at the target remains below the melting
  threshold for W.
• Some shallow melting can nevertheless take place
  some times.

• Concern remains on whether generation during
  ELMs/disruptions of surface irregularities in
  tungsten due to melting, and in CFC due to
  brittle destruction, might form hot spots during
  normal operation.

18
Effects of mitigated disruptionsAllowed time
scale for energy dissipation (see talk of D.
Whyte)
ITER

• The thermal energy density in ITER (plasma stored
  thermal energy/ wall surface area) is 350
  MJ/800m2 0.45 MJ/m2.
• This energy density sets the minimum time in
  which the plasma thermal energy can be
  radiatively dissipated to the wall before
  melting/ablation occurs.

• The allowed time scales for ITER w. Be first
  wall is closed to the limits set by the MHD time
  scale.
• Assuming uniform (spatial and temporal)
  dissipation tlim0.5 ms.
• Assuming a PF of 1.5-2 tlim2 ms.

Melting as a function of deposited energy
G. Federici, G. Strohmayer 3/2003
19
ITER
Parametric analyses of VDE effectsVDE
parameters 60 MJm-2, 300 ms, 1
UncertaintiesMelting - comparison Be and W
10 mm 2 mm

lt10 MJ/m2 t500 ms avoid melting
lt10 MJ/m2 tgt500 ms avoid melting
Be
lt10 MJ/m2 tgt100 ms avoid melting
lt10 MJ/m2 tgt100 ms avoid melting
W
If t500 ms no melting up to 70 MJ/m2
If t500 ms no melting up to 30 MJ/m2
20
ITER
Parametric analyses of VDE effectsTemp.
excursions at Cu interface - comparison Be and W
10 mm 2 mm

TCu melting
Be
TCu melting
W
Problems!! We need at least 5 mm W
21
ITER
Future Work/ Open Questions3-5 years before
starting PFC material procurement for ITER
Current efforts EU task force on PWI / ITPA
work

• Type I ELM and disruption heat loads - energy
  loss, energy to divertor and wall, duration,
  broadening, impurities, etc. (ITPA, JET TF, EU
  PWI TF)
• Mitigation of disruptions (ITPA, JET TF, EU PWI
  TF)
• Effect of mitigated disruptions on Be wall
  (ITPA)
• Material damage due to ELMs/disruptions (Russia,
  EU PWI TF)
• Erosion - mixed material effects (PISCES-B/EFDA)
• Explain co-deposition in existing machines (ITPA,
  JET EU PWI TFs)
• Mitigation of co-deposition (temperature
  tailoring, reactive species (N2))
• Monitor CxHy deposits on cryopumps (JET, AUG)
• Carbon removal - e.g. baking with oxygen -
  temperature, hold time (???)
• Dust generation - volume, location, BET,
  mobilisation.
• Plasma interaction with irregular and/or molten
  W surfaces.
• Measurements/diagnostics

22
ITER
SummaryOpportunities and challenges

• According to current ITER construction plans, 3-5
  years are available for further RD and physics
  input to divertor/first-wall design and PFM
  choice.
• We take a big risk if we construct ITER without
  first testing/validating proposed material mix in
  an existing tokamak.
• We should reduce to a minimum the extent that
  ITER has to be a PMI experiment and explore all
  these effects before hand in existing machines.
• In case we retain CFC armour and use it during T
  operation a reliable method to remove the
  codeposted layers and control T-uptake is
  required.
• Type-I ELMs are still challenging for the ITER
  divertor and design optimisation is ongoing.
  Power deposition on the first-wall is still
  uncertain.
• IF JET divertor results extrapolate to ITER, the
  large majority of disruptions would not lead to
  melting a W target. Compatibility of plasma with
  irregular W surfaces and macroscopic C erosion
  and impact on operation/performance require
  investigations.
• ITER divertor design has enough flexibility.
  Feasibility of maintenance has been demonstrated.
  For the first-wall only infrequent maintenance is
  anticipated and feasibility demonstration is
  needed.