Title: Gianfranco Federici
1Technology Development Status and Prospects for
Divertor PFCs in DEMO
- Gianfranco Federici
- F4E, Barcelona, Spain
Meeting on DEMO Garching, Germany September
29-30, 2009
2Outline
- 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
3Power 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)
4Divertor 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.
6ITER 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
7ITER 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
8Still 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
9Divertor 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.
10NEED 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)
11He-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)
12He-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
13Early 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
14Early 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
15Results 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.
16DEMO 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 15Â 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)
17Innovation 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
18Hot Tungsten and Liquid Metals introduce New
Issues of Impurity Influx
19Prospects/ 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 21ITER 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