1. Structural Materials for DEMO: Development, Testing and Modelling
R. Lässer1, N. Baluc2, J.-L. Boutard1, E. Diegele1, S. Dudarev3, M. Gasparotto1, A. Möslang4, R. Pippan5, B. Riccardi1 and B. van der Schaaf6
1 EFDA Close Support Unit Garching, D Garching, Germany
2 CRPP-EPFL, CH-5232 Villigen-PSI, Switzerland
3 Euratom/UKAEA Fusion Association, Culham Science Centre, OX14 3DB UK
4 Forschungszentrum Karlsruhe, Karlsruhe, Germany
5 Erich Schmid-Institute, Leoben, Austria
6 NRG, Petten, The Netherlands
R. Lässer, 24th SOFT, Sept 2006, Warsaw

2. Content Introduction Materials Development
Path to DEMO
European breeder blanket strategy
Materials Development
Similarities and differences of fusion and fission neutrons
Neutron damage
Portfolio of structural materials for DEMO
RAFM steel EUROFER
ODS steels
Tungsten and tungsten alloys
SiCf/SiC
European Modelling Programme for irradiation effects
Scales and tools for multi-scale modelling
He thermodynamics and desorption in Fe
Issue of additional He and testing under fusion relevant conditions
Conclusions
R. Lässer, 24th SOFT, Sept 2006, Warsaw

3. Path to DEMO IFMIF DEMO ITER Tech- nology R&D Physics R&D
Structural Materials
Test Blanket Modules
IFMIF
Tech- nology
R&D
Components
SC Magnets
Tritium Handling System
Plasma Facing Components
Remote Maintenance System
Heating System
Safety
Blanket tests
in ITER
Facilities
Confinement
Impurity Control
Plasma Stability
ITER/DEMO Physics Support
DEMO
Physics
R&D
ITER
R. Lässer, 24th SOFT, Sept 2006, Warsaw

4. Comparison: ITER, DEMO and Power Reactor
Fusion Power
0.5 GW
2 – 2.5 GW
3 - 4 GW
Heat flux (FW)
MW/m2
0.5 MW/m2
Neutron Wall Load (First Wall)
0.78 MW/m2
< 2 MW/m2
~ 2 MW/m2
Integrated wall load (First Wall)
0.07 MW.year/m2
(3 years Inductive operation I)
5 - 8 MW.year/m2
MW.year/m2
Displacement per atom (dpa)
< 3 dpa
dpa
dpa
The challenge
Transmutation product rates at first wall
~10 appm Helium / dpa
~45 appm H / dpa
R. Lässer, 24th SOFT, Sept 2006, Warsaw

5. The European Breeder Blanket Strategy
Near Term:
A) Helium-Cooled Lithium-Lead (HCLL) Blanket,
B) Helium-Cooled Pebble-Bed (HCPB) Blanket.
Their Test Blanket Modules (TBMs) will be tested in ITER. They use EUROFER as structural material.
Long Term:
C) Dual-coolant concept: LiPb blanket with He-cooled steel box and divertor, uses ferritic-martensitic steel (EUROFER) for structures and SiCf/SiC for insulating flow inserts,
Even Longer Term:
D) Self-cooled LiPb blanket with SiCf/SiC as structural material.
HCLL TBM
First wall
Long-term Issue: Future fusion reactors (starting with DEMO) will require tritium self-sufficiency.
Breeder Blanket Strategy + Materials Development are stongly coupled.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

6. Increasing Attractiveness Increasing Development „Risk“
Materials for Blanket and Divertor according PPCS
Model A or WCLL
Model AB
or HCLL
Model B or HCPB
Model C or Dual-Coola.
Model D or Self-cooled
Structural material
EUROFER
EUROFER
EUROFER
EUROFER
SiCf/SiC
Coolant
H2O He He
LiPb/ He
LiPb
Coolant I/O T(°C)
285 / 325
300 / 500
300 / 500
300 / 500
480 / / 480
700 / 1100
blanket
divertor
Breeder
LiPb
LiPb
Li4SiO4
LiPb
LiPb
TBR
1.06
1.13
1.12
1.15
1.12
Structural material
CuCrZr
W alloy
W alloy
W alloy
SiCf/SiC
Increasing Attractiveness
Increasing Development „Risk“
Armour material W W W W W
Coolant
H2O He He He
LiPb
Coolant I/O T(°C)
140/167
~ 540/720
~ 540/720
~ 540/720
~ 600/990
Net efficiency
0.31
0.35
0.36
0.42
0.60
Materials listed above require R&D.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

7. Materials Development: Mission driven
In ITER: Use of austenitic steels SS316 for shielding blankets and vacuum vessel is acceptable (low neutron fluence and low temperature application).
In Fusion Power Reactors (DEMO, PROTO): Other types of structural materials are required due to the effects of high energy fusion neutrons and higher operation temperatures. In particular, these materials have a different crystal structure (bcc) to avoid excessive volume changes under n-irradiation.
The materials design requirements for DEMO-relevant materials include:
Good physical (e.g. thermal conductivity, thermal expansion) and mechanical (tensile and fracture) properties, in particular also good creep strength and fatigue resistance.
Ductile to Brittle Transition Temperature (DBTT) well below 250°C at the end of life (irradiation dose up to 70 dpa at least).
Minimum embrittlement due to transmutation products (hydrogen-isotopes and Helium).
Good compatibility with lithium lead (corrosion resistance) and low hydrogen permeation.
Low residual activation under neutron irradiation.
Dimensional stability under fusion reactor relevant conditions (low swelling).
R. Lässer, 24th SOFT, Sept 2006, Warsaw

8. Fusion and Fission Neutrons in Materials: Similarities and Differences
In fission most neutrons have energies below 2 MeV.
In fusion the D-T generated neutrons have 14.1 MeV and are used
to transfer 80% of fusion energy from the plasma into the blanket for further power conversion,
to reproduce the tritium burnt in the fusion reaction or lost elsewhere to achieve tritium self-sufficiency.
Both, fission and fusion neutrons, cause activation and irradiation damage.
The more energetic fusion neutrons cause far larger damage (multiple cascades) and more transmutation products, e.g. H and He, (due to many new nuclear reaction channels) and higher activation than fission neutrons.
As a consequence: Structural materials developed for conventional fission have to be improved and modified for fusion application
to be resistant to the fusion environment and fusion loading conditions
to fulfill the requirement of low activation waste (recycling after 100 years).
In particular, some alloying elements acceptable in fission have to be avoided in fusion.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

9. Neutron Damage in Materials (Primary Damage)
7 keV Cascade in Ni (fcc)
High energy PKAs in Fe (bcc)
0,3 ps
0,7 ps
1,5 ps
2,5 ps
10,3 ps
Fission:
Emax (n) = 2 MeV,
Emax (PKA)=0.14 MeV.
Fusion:
E (n) = 14.1 MeV, Emax (PKA)= 1 MeV.
PKA
Neutron irradiation
destroys the crystal structure and affects the chemical bonds,
creates point defects, He and H, clusters, modifies the microstructure and leads to hardening/embrittlement.
causes degradation of physical and mechanical properties.
Yellow: Vacancies + V-clusters; Brown: Interstitials + I-clusters; BLUE: Atoms displaced but on regular positions (no effect in metals, but large one in ordered alloys)
R. Lässer, 24th SOFT, Sept 2006, Warsaw

10. The Portfolio of Structural Materials for DEMO
Reduced Activation Ferritic Martensitic (RAFM) steel EUROFER 9%Cr-W-V-Ta-steel (0.1% C)
The EU reference structural materials for breeding blankets in DEMO (and in the first Power Plant according to „Fast Track“)
Will be used in the EU Test Blanket Modules (TBMs) to be installed in ITER. Operational limits ~300º - 550°C.
ODS-RAFM steels
Developed to increase the upper temperature limit to 650°C, or even 700°C for nano-structured ferritic steels (12-14% Cr), in order to achieve higher thermal efficiency of the breeder blanket concepts.
In addition, ODS materials can be also used as backbone material of the He cooled divertor concept.
SiCf/SiC ceramic composites for advanced Breeding Blanket concepts
Considered in the long term for their potential to increase thermal efficiency (model D of PPCS). Operational T-window °C. First use likely as functional material (flow channel inserts for Dual Coolant BB Concept).
Tungsten alloys for structural application in gas cooled divertor
Candidate material in the high temperature region of the gas cooled divertor concepts.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

11. EUROFER Alloy Development
Development Strategy
EUROFER is a RAFM steel developed on the basis of conventional 9%Cr-1%Mo steels used in fission, where
Highly activating alloying elements (Mo, Nb) were replaced by those (W, Ta) offering lower activation.
8-10% Cr: optimized concentration for good fracture properties and corrosion resistance.
1-2% W: optimized for mechanical properties (ductility, strength, fracture properties).
0.07% Ta: stabilizes grain size and improves strength
Highly activating impurity elements (Nb, Mo, Ni, Cu, Al, Si, Co,..) are reduced to the “lowest” content, that is technically achievable at reasonable cost.
As, Be, H, Sb, P, O, S, Sn should be avoided because they degrade mechanical properties, same holds for P and B (transmute and generate He).
R. Lässer, 24th SOFT, Sept 2006, Warsaw

12. Issues and Objectives identified for RAFM Steels
Critical issues at lower operational temperature limit (300°C):
Radiation hardening and embrittlement,
Additional effect of superimposed helium (He/dpa),
Reduction of the uncertainties in the DBTT and in fracture toughness.
Critical issues at upper operational temperature limit (550°C):
Thermal creep,
Compatibility (corrosion by Pb17Li above 500°C)
Detailed analysis of the irradiation data is being performed for a better understanding of the role of the major elements (Cr, Ta, W) in order to improve possibly the composition of the EUROFER and to focus the R&D on the most critical area (i.e. temperature range for the irradiation) leading to EUROFER-2.
Production of EUROFER heats with controlled impurity contents to confirm the reproducibility of the properties and the low activation potential (EUROFER-3).
The development of sound welds and dissimilar connections with improved properties requires post-heat treatment at about 730°C. More generally, particular attention should be paid on the welds and joints development. This is necessary for the production of the TBMs.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

13. Yield stress of unirradiated EUROFER
The properties of unirradiated EUROFER are today well known.
The data of EUROFER achieved in the past are stored in relevant Data base.
Further high temper-ature (HT) rules and Structural Design Criteria (SDC-IC) are still needed.
Upper temperature limit: 550ºC
R&D for highly irradiated EUROFER is ongoing and needed in future.
F. Tavassoli, CEA
R. Lässer, 24th SOFT, Sept 2006, Warsaw

14. Degradation of Impact Properties under neutron irradiation
~32 dpa, 332°C, ARBOR 1 irradiation
Irradiation effects
Ductile-Brittle Transition
-30%
Unirradiated
~ 200 K
Irradiated
EUROFER 97
C. Petersen, FZK
TBM design window
Concerns: i) ΔDBTT > 200 K ii) Effect of Helium?
Operational window
R. Lässer, 24th SOFT, Sept 2006, Warsaw

15. Embrittlement behavior of irradiated RAFM steels
Irradiation conditions: 16 dpa, °C
Higher DBTT are observed at lower irradiation temperatures (Tirr ≤ 300°C).
DBTT shift is of less concern for Tirr > 350°C.
EUROFER
F82H
E. Gaganidze, FZK
Irradiation is highest and thus most critical at FW, but only small volumes around the cooling channels are at T ~ 300°C to 370°C during steady state (TBM: plasma heat flux MWm-2 and NWL = 0.78 MWm-2).
R. Lässer, 24th SOFT, Sept 2006, Warsaw

16. A potential Strategy for Recovery of Impact Properties
Annealing at ~ 500°C
Recovery of properties
Unirradiated
Irradiated and
annealed
Irradiated
15 dpa, 330°C
EUROFER 97
EUROFER 97
C. Petersen, FZK
How often can this recovery be achieved? What about memory effects?
How is the degradation and recovery under subsequent irradiation?
Can such an annealing step at 500°C also be done with BBs?
What happens if large concentrations of He are present?
R. Lässer, 24th SOFT, Sept 2006, Warsaw

17. Various PWHT performed
RAFM Steels – Fabrication Processes Comparative study on Post Weld Heat Treatments (PWHTs)
Notches of charpy specimens in the weld centre.
M. Rieth, FZK
EB 5mm
LASER
5 mm
Laser 5mm
TIG
5 mm
TIG 5mm
Base material
TIG 10mm EB
5 mm
Various PWHT performed
Heat Treatment at 700°C for 2 h,
considered as „limiting“ case
R. Lässer, 24th SOFT, Sept 2006, Warsaw

18. Results on Irradiated EUROFER (Welds)
TIG EB
J.W. Rensman, NRG-FOM
Preliminary Results
----- unirradiated EUROFER base material
----- irradiated EUROFER base material ~2 dpa
Irradiation of TIG welds might be of concern even at low dose (~2 dpa) and Tirr = 300°C
R. Lässer, 24th SOFT, Sept 2006, Warsaw

19. ODS Steels Potential use of ODS-Layer plated to FW
Composition: EUROFER powder plus 0.3 wt% Y2O Nb, Mo, Ni, Cu, Al and Co < ppm values.
Advantage:
EUROFER based RAFM-ODS material exhibits higher thermal creep resistance than conventional RAFM steels due to the oxide particles and can be used at temperature up to 650°C – 700°C .
Critical issues:
The main problem is the embrittlement at low temperature (higher DBTT) and reduced fracture toughness compared to conventional (EUROFER) steels.
The oxide dispersion has to be stable under irradiation to keep the high initial thermal creep resistance.
The fabrication processes are still to be optimized, e.g. with respect to mechanical and thermal treatment.
Potential use of ODS-Layer plated to FW
R. Lässer, 24th SOFT, Sept 2006, Warsaw

20. A) Creep Behaviour of ODS-EUROFER compared to EUROFER
Gain in creep strength
ODS-EUROFER
EUROFER
ODS-EUROFER:
The temperature-window increased by ~100 K to 650ºC
R. Lindau, FZK
R. Lässer, 24th SOFT, Sept 2006, Warsaw

21. B) Impact Energy Values and Comparison with EUROFER-97 KLST Specimens
M23C6
2nd Generation ODS-EUROFER
1st Generation ODS-EUROFER
EUROFER 97
M. Klimiankou et al., FZK
1m
(FeCr)23C6
M. Rieth, FZK
R. Lässer, 24th SOFT, Sept 2006, Warsaw

22. Tungsten and Tungsten Alloys
Potential application:
W and W-alloys are promising materials for high temperature structural applications, e.g. the hottest part in the “high heat flux, high temperature heat removal units” of gas cooled divertors.
Requirement for this application:
Advantages of W-alloys: High melting point, high thermal conductivity and good thermal shock resistance, (low vapour pressure, high erosion resistance).
Critical issues of W-alloys: Creep rate and strength (700 to 1300°C), fracture toughness (RT to 1300°C), DBTT usually well above RT, ductility, recristallisation, low and high cycle fatique, low oxidation resistance above 490°C, behaviour under irradiation,
He-cooled Modular divertor
with multiple jet Cooling
HEMJ 10 MW / m2
P. Norajitra, FZK
W tile: max. allow temp °C max. calc. temp. 1711°C DBTT (irr.): 700°C
Thimble: max. allow. temp °C max. calc. temp. 1170°C DBTT (irr.): °C
ODS-Eurofer: He-out temp °C He-in temp °C DBTT (irr.): 300°C
R. Lässer, 24th SOFT, Sept 2006, Warsaw

23. Strategy for Tungsten-alloys
This W program is worldwide a unique effort. Scientific understanding and database are very limited. No obvious development path exists. Alloying elements should not affect good thermal properties and low activation.
First screening tests started to improve most critical properties (fracture toughness) and a limited irradiation program.
Main development route: Refinement of microstructure by production of ultra fine grained (UFG) materials.
Material investigated: W,
W 1%La2O3 (WL10),
W potassium doped (WVM).
W-Re no longer pursued.
30 MPa√m
Findings:
Fracture toughness at RT increases to ~ 30 MPa√m (increase by a factor of 5, still low compared to other materials like steel).
DBTT is shifted towards lower temperatures,
WVM is promising.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

24. Thermal Stability - MicrostructureF
Fixed
Rotating
Pistons
Sample
Thermal Stability - Microstructure
Severe Plastic Deformation (SPD): e.g. High Pressure Torsion (HPT)
Thermal treatment in vacuum furnace at 1200°C for 1 hour (requirement: thermal stability for 1000 h at 1200°C)
1µm
1µm
BSE investigation
HPT-W, e = 64
HPT-W, e = 64,
Before and after annealing at 1200°C for 1 h
R. Lässer, 24th SOFT, Sept 2006, Warsaw

25. SiCf/SiC Ceramic Composites
SiCf/SiC composites are a promising structural material in the advanced LiPb self-cooled breeding blanket concepts offering high thermal efficiency.
Advantages:
 Low activation characteristics (at short and medium term) and low afterheat,
 Engineerability to cover wide range of properties,
Good creep strength and life time properties up to high temperatures ( °C),
 High corrosion resistance in LiPb
Critical issues:
Primary basic issues
 Nuclear transmutation products: H, He production due to (n,p), (n,α) reactions,
 Radiation stability of physical (thermal conductivity) and mechanical properties.
Technological issues
High porosity and high permeability requiring coatings,
Fabrication and joining (brazing) of large components,
Development of guidelines for designing components (inherent brittleness, anisotropy).
R. Lässer, 24th SOFT, Sept 2006, Warsaw

26. SiCf/SiC Ceramic Composites
Composites have been fabricated industrially as plates (area 20x20 cm2) in the EU by Chemical Vapour Infiltration (CVI) (MT Aerospace AG, Germany).
Properties of SiCf/SiC composites depend on
Fibres: Tyranno SA-3 fiber (UBE Industries, Japan),
Fibre architecture: 2D and 3D fabrics,
Interphases: single layer of pyrolithic carbon of 80 nm thickness,
Matrix: CVI SiC.
Density: 2.70 g/cm3 (2D), g/cm3 (3D).
Specified properties for 2D and 3D SiCf/SiC composites were achieved.
3D composites (MT Aerospace)
fibre
pore
matrix
R. Lässer, 24th SOFT, Sept 2006, Warsaw

27. European Modelling Programme of Irradiation Effects in RAFM Steel
Objectives:
Study of the radiation effects in EUROFER under fusion relevant conditions from RT to 550°C and in the presence of high concentrations of nuclear transmutation products (i.e. H, He).
Development of tools:
to correlate results from MTRs, fast reactors, spallation sources, accelerators, fusion neutron sources, etc.,
to extend the understanding of the effects of irradiation damage to the high fluence and high He & H concentrations relevant for DEMO and fusion power reactors.
Experimental validation of the models and the derived tools.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

28. Only effect of dislocations is their bias and action as sink.
Radiation Modified Microstructure Scale and tools for multi-scale modelling
Rate Theory
Monte Carlo on Objects
Monte Carlo on Events
Ballistic Phase
Thermalisation
Short Term
Recovery
Diffusion
Micro --
structure
Primary Damage
Displacement Cascades
Molecular Dynamics
Short Term prediction
Atomic Kinetic Monte Carlo 10 - 15 s s 8
Lifetime of Components
Long Term Prediction
no long range strain
Only effect of dislocations is their bias and action as sink.
10-15 s
10-13 s
10-10 s
Lifetime of components
R. Lässer, 24th SOFT, Sept 2006, Warsaw

29. Radiation-induced Hardening Scale and Tools for Multi-scale Modelling
Molecular Dynamics:
void-dislocation interaction
2. Discrete Dislocation Dynamics
One dislocation
Collective Behavior
40 nm
Increasing strain
5 µm
Surface strain distribution
3. Crystalline plasticity Finite Elements: low-alloy steel
Tensile
Average Tensile strain 5 %
Experiment
Computation
Average Tensile strain 9 %
Experiment
Computation
S. Sekfali, PhD, 2003
Strain ε
R. Lässer, 24th SOFT, Sept 2006, Warsaw

30. He Thermodynamics & Desorption
Ab initio solution energies: substitutional He: the stable configuration
He on substitu- tional site
He on interstitial
(tetrahedral) site
(octahedral) site
E Sol = 4.22 eV
E Sol = 4.39 eV
E Sol = 4.57 eV
C.C. Fu, F. Willaime, CEA
Ab initio energies: 1) Em(Hei) = 0.06eV. 2) Eb of Hei with point defects.
He-desorption: presence of C in He-impl. Fe needs to be considered
M.J. Caturla,
Uni. Alicante
Ab initio for pure Fe:
Ef (V) = 2.0 eV, Em (V) = 0.67 eV
Experimental Ef (V) & Em (V) of Fe-C1:
Ef (V) = 1.6 eV & Em (V) = 1.1 eV
1 C. Moser et al.: Atomic defects in metals
R. Lässer, 24th SOFT, Sept 2006, Warsaw

31. The Issue of additional Helium and of Testing structural Materials under Fusion relevant Conditions
Degradation of mechanical properties under irradiation strongly depends on irradiation temperature and the amount of transmutation products produced.
Fission neutrons produce about a factor of 40 less He than 14.1 MeV ones. Hence, irradiation in fission reactors gives only non-conservative results.
Various tricks or methods were used to produce higher He and H to dpa ratios in the absence of an intense fusion neutron source (B and Ni-doped steels or Fe54 enriched steels). Any of these methods is still short by about a factor of 10 to 5.
Other facilities can be used, but also have their own shortcomings.
Mixed spallation-neutron spectrum in a spallation target: ~102 appm He/dpa,
But due to many other transmutation products and inhomogeniety of irradiation conditions it is difficult to draw conclusions.
Energetic ( MeV) alpha particle implantation: ~ appm He/dpa.
Dual/triple beam irradiation (JANNuS) (Ion E ~a few MeV): up to 104 appm He/dpa.
But only few microns depth, so no mechanical testing, only microstructure.
Consequences:
(1 ) Modelling and understanding of irradiation results obtained under various conditions are clearly needed.
(2) A fusion relevant neutron source is mandatory for a “correct” characterisation of the materials in the sense that they become licensable: IFMIF
R. Lässer, 24th SOFT, Sept 2006, Warsaw

32. Conclusions (1/2) Development of RAFM steels:
Properties of unirradiated EUROFER are now well known.
Short-term (ITER): For the use of EUROFER in the fabrication of TBMs some technology issues (welding, joining, HIPping, PWHTs) are to still be solved.
Long-term (DEMO): Effects of high He concentrations on degradation of mechanical properties are to be studied. IFMIF and modelling are complementary. Both are mandatory.
ODS steels (9%Cr EUROFER-type and 12-14%Cr ferritic): Potential of higher upper operational temperature limit. Improvement of production processes ongoing, irradiation campaigns to address fundamental issues (on oxide stability and He trapping by oxides) will provide first answers on the time frame and the amount of further R&D needed before their application.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

33. Conclusions (2/2)
Development of materials for high-temperature application:
SiCf/SiC: Industrial production of larger samples with acceptable properties. Knowledge and data base on fundamental issues (He-effects and radiation stability) to be increased. First production run of SiCf/SiC for flow channel inserts in dual-coolant concept (application requires less strength, issue is the radiation stability of the low electrical conductivity).
Tungsten-alloys: R&D started in 2003, still in the phase where (a) the required properties are far from being achieved collectively, (b) the understanding of irradiation behaviour is very limited. It first needs a broader science driven basic programme.
Modelling of irradiation effects:
Effort increased to understand better the irradiation effects on microstructure in bcc Fe-Cr-C steels. Important intermediate results were achieved. Modelling will help in optimization of irradiation campaigns and understanding of physical properties under DEMO-relevant conditions.
Experimental validation of models and computational tools to be enforced using MTRs, fast reactors, ion beam facilities (JANNUS) and IFMIF.
R. Lässer, 24th SOFT, Sept 2006, Warsaw

34. END
R. Lässer, 24th SOFT, Sept 2006, Warsaw

35. RAFM ODS Steel: HFR Neutron Irradiation
Medium dose irradiation - outstanding results:
EUROFER ODS shows substantial work hardening (equivalent to significant elongation improvement)
EUROFER ODS shows almost no loss of uniform elongation
E. Materna-Morris and R. Lindau, FZK
R. Lässer, 24th SOFT, Sept 2006, Warsaw

36. Budget Sharing within the Materials Development Area
Budget: ~ 3 M€ /year CEC
Type of Contract
Materials Development
[Average]
Breeding Blankets
[Average] (for comparison)
R&D (20-40%)
9.0 M€/y
8.5 M€/y (25 M€ in 2002)
Studies (40%)
0.4 M€/y
0.2 M€/y
Industrial contracts (100%)
R. Lässer, 24th SOFT, Sept 2006, Warsaw