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R. Lässer, 24 th SOFT, 11-15 Sept 2006, Warsaw 1 of 35 slides Structural Materials for DEMO: Development, Testing and Modelling R. Lässer 1, N. Baluc 2,

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Presentation on theme: "R. Lässer, 24 th SOFT, 11-15 Sept 2006, Warsaw 1 of 35 slides Structural Materials for DEMO: Development, Testing and Modelling R. Lässer 1, N. Baluc 2,"— Presentation transcript:

1 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 1 of 35 slides Structural Materials for DEMO: Development, Testing and Modelling R. Lässer 1, N. Baluc 2, J.-L. Boutard 1, E. Diegele 1, S. Dudarev 3, M. Gasparotto 1, A. Möslang 4, R. Pippan 5, B. Riccardi 1 and B. van der Schaaf 6 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

2 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 2 of 35 slides Content Introduction –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 –SiC f /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

3 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 3 of 35 slides Components SC Magnets Tritium Handling System Plasma Facing Components Remote Maintenance System Heating System Safety Test Blanket Modules Structural Materials Blanket tests in ITER Facilities Confinement Impurity Control Plasma Stability ITER/DEMO Physics Support Tech- nology R&D PhysicsR&D Path to DEMO ITER DEMO IFMIF

4 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 4 of 35 slides Comparison: ITER, DEMO and Power Reactor ITERDEMOReactor Fusion Power 0.5 GW2 – 2.5 GW3 - 4 GW Heat flux (FW) MW/m MW/m 2 Neutron Wall Load (First Wall) 0.78 MW/m 2 < 2 MW/m 2 ~ 2 MW/m 2 Integrated wall load (First Wall) 0.07 MW.year/m 2 (3 years Inductive operation I) MW.year/m MW.year/m 2 Displacement per atom (dpa) < 3 dpa dpa dpa The challenge Transmutation product rates at first wall ~10 appm Helium / dpa ~45 appm H / dpa

5 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 5 of 35 slides Long-term Issue: Future fusion reactors (starting with DEMO) will require tritium self-sufficiency. Breeder Blanket Strategy + Materials Development are stongly coupled. 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 SiC f /SiC for insulating flow inserts, Even Longer Term: D) Self-cooled LiPb blanket with SiC f /SiC as structural material. The European Breeder Blanket Strategy First wall HCLL TBM

6 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 6 of 35 slides EUROFER LiPb 300 / 500 W alloy ~ 540/720 He W Increasing Attractiveness Increasing Development Risk ~ 600/990~ 540/ /167Coolant I/O T(°C) LiPbHe H2OH2OCoolant WWWWArmour material SiC f /SiCW alloy CuCrZrStructural material TBR LiPb Li 4 SiO 4 LiPbBreeder 700 / / / / / 325 Coolant I/O T(°C) LiPbLiPb/ He H2OH2O Coolant SiC f /SiCEUROFER Structural material Model D or Self-cooled Model C or Dual-Coola. Model B or HCPB Model A or WCLL blanket divertor Materials for Blanket and Divertor according PPCS Net efficiency Materials listed above require R&D. Model AB or HCLL He 300 /

7 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 7 of 35 slides 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). 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. Materials Development: Mission driven

8 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 8 of 35 slides Fusion and Fission Neutrons in Materials: Similarities and Differences In fission most neutrons have energies below 2 MeV.In fission most neutrons have energies below 2 MeV. In fusion the D-T generated neutrons have 14.1 MeV and are usedIn 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.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.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.

9 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 9 of 35 slides 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) Neutron Damage in Materials (Primary Damage) 7 keV Cascade in Ni (fcc) 0,3 ps0,7 ps 1,5 ps 2,5 ps 10,3 ps PKA High energy PKAs in Fe (bcc) 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. Fission: E max (n) = 2 MeV, E max (PKA)=0.14 MeV. Fusion: E (n) = 14.1 MeV, E max (PKA)= 1 MeV.

10 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 10 of 35 slides The Portfolio of Structural Materials for DEMO Reduced Activation Ferritic Martensitic (RAFM) steel EUROFER 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. SiC f /SiC ceramic composites for advanced Breeding Blanket concepts Con 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.

11 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 11 of 35 slides 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.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).

12 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 12 of 35 slides Issues and Objectives identified for RAFM Steels 1.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. 2.Critical issues at upper operational temperature limit (550°C): Thermal creep, Compatibility (corrosion by Pb17Li above 500°C) 3.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. 4.Production of EUROFER heats with controlled impurity contents to confirm the reproducibility of the properties and the low activation potential (EUROFER-3). 5.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.

13 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 13 of 35 slides 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

14 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 14 of 35 slides ~ 200 K -30% TBM design window Irradiation effects Operational window Unirradiated Irradiated Degradation of Impact Properties under neutron irradiation ~32 dpa, 332°C, ARBOR 1 irradiation Concerns: i) ΔDBTT > 200 K ii) Effect of Helium? EUROFER 97 Ductile-Brittle Transition C. Petersen, FZK

15 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 15 of 35 slides Embrittlement behavior of irradiated RAFM steels Irradiation conditions: 16 dpa, °C Higher DBTT are observed at lower irradiation temperatures (T irr 300°C). DBTT shift is of less concern for T irr > 350°C. E. Gaganidze, FZK F82H EUROFER 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 0.25 MWm -2 and NWL = 0.78 MWm -2 ).

16 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 16 of 35 slides Annealing at ~ 500°CAnnealing at ~ 500°C Recovery of propertiesRecovery of properties 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? A potential Strategy for Recovery of Impact Properties EUROFER 97 Unirradiated Irradiated 15 dpa, 330°C Irradiated and annealed EUROFER 97 C. Petersen, FZK

17 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 17 of 35 slides RAFM Steels – Fabrication Processes Comparative study on Post Weld Heat Treatments (PWHTs) LASER 5 mm TIG 5 mm EB 5 mm Heat Treatment at 700°C for 2 h, considered as limiting case Various PWHT performed TIG 10mm TIG 5mm EB 5mm Base material Laser 5mm M. Rieth, FZK Notches of charpy specimens in the weld centre.

18 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 18 of 35 slides TIG EB 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 T irr = 300°C Results on Irradiated EUROFER (Welds) J.W. Rensman, NRG-FOM

19 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 19 of 35 slides ODS Steels Composition: EUROFER powder plus 0.3 wt% Y 2 O 3. 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

20 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 20 of 35 slides A) Creep Behaviour of ODS-EUROFER compared to EUROFER ODS-EUROFER EUROFER ODS-EUROFER: The temperature- window increased by ~100 K to 650ºC Gain in creep strength R. Lindau, FZK

21 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 21 of 35 slides B) Impact Energy Values and Comparison with EUROFER-97 KLST Specimens 2 nd Generation ODS-EUROFER 1 st Generation ODS-EUROFER EUROFER 97 1 m M 23 C 6 1 m (FeCr) 23 C 6 M. Klimiankou et al., FZK M. Rieth, FZK

22 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 22 of 35 slides 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, W tile: W tile: max. allow temp. 2500°C max. calc. temp. 1711°C DBTT (irr.): 700°C Thimble: Thimble: max. allow. temp. 1300°C max. calc. temp. 1170°C DBTT (irr.): 600°C ODS-Eurofer: ODS-Eurofer: He-out temp. 700°C He-in temp. 600°C DBTT (irr.): 300°C HEMJ 10 MW / m 2 P. Norajitra, FZK He-cooled Modular divertor with multiple jet Cooling

23 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 23 of 35 slides 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. Findings: Fracture toughness at RT increases to ~ 30 MPam (increase by a factor of 5, still low compared to other materials like steel). DBTT is shifted towards lower temperatures, WVM is promising. 30 MPam

24 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 24 of 35 slides Thermal Stability - Microstructure HPT-W, HPT-W,, Before and after annealing at 1200°C for 1 h 1µm Thermal treatment in vacuum furnace at 1200°C for 1 hour (requirement: thermal stability for 1000 h at 1200°C) F F Fixed Rotating Pistons Sample Severe Plastic Deformation (SPD): e.g. High Pressure Torsion (HPT) BSE investigation

25 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 25 of 35 slides SiC f /SiC Ceramic Composites SiC f /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).

26 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 26 of 35 slides SiC f /SiC Ceramic Composites Composites have been fabricated industrially as plates (area 20x20 cm 2 ) in the EU by Chemical Vapour Infiltration (CVI) (MT Aerospace AG, Germany). Properties of SiC f /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/cm 3 (2D), 2.65 g/cm 3 (3D). Specified properties for 2D and 3D SiC f /SiC composites were achieved. 3D composites (MT Aerospace) fibre pore fibrematrix

27 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 27 of 35 slides 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). 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: Development of tools: oto correlate results from MTRs, fast reactors, spallation sources, accelerators, fusion neutron sources, etc., oto 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. Experimental validation of the models and the derived tools.

28 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 28 of 35 slides Radiation Modified Microstructure Scale and tools for multi-scale modelling Rate Theory Monte Carlo on Objects Monte Carlo on Events Ballistic PhaseThermalisation Short Term Recovery Diffusion Micro -- structure Primary Damage Displacement Cascades Molecular Dynamics Short Term prediction Molecular Dynamics Atomic Kinetic Monte Carlo s s s Lifetime of Components Long Term Prediction no long range strain Only effect of dislocations is their bias and action as sink s s s Lifetime of components

29 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 29 of 35 slides Radiation-induced Hardening Scale and Tools for Multi-scale Modelling 1.Molecular Dynamics: void-dislocation interaction 40 nm 2. Discrete Dislocation Dynamics One dislocationCollective Behavior 5 µm Increasing strain Tensile Surface strain distribution Average Tensile strain 5 % Experiment Computation Average Tensile strain 9 % Experiment Computation 3. Crystalline plasticity Finite Elements: low-alloy steel Strain ε S. Sekfali, PhD, 2003

30 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 30 of 35 slides He Thermodynamics & Desorption He on substitu- tional site He on interstitial (tetrahedral) site He on interstitial (octahedral) site E Sol = 4.22 eV E Sol = 4.39 eV E Sol = 4.57 eV Ab initio solution energies: substitutional He: the stable configuration Experimental E f (V) & E m (V) of Fe-C 1 : E f (V) = 1.6 eV & E m (V) = 1.1 eV Ab initio energies: 1) E m (He i ) = 0.06eV. 2) E b of He i with point defects. Ab initio for pure Fe: E f (V) = 2.0 eV, E m (V) = 0.67 eV He-desorption: presence of C in He-impl. Fe needs to be considered C.C. Fu, F. Willaime, CEA M.J. Caturla, Uni. Alicante 1 C. Moser et al.: Atomic defects in metals

31 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 31 of 35 slides 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 Fe 54 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: ~10 2 appm He/dpa,Mixed spallation-neutron spectrum in a spallation target: ~10 2 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.Energetic ( MeV) alpha particle implantation: ~ appm He/dpa. Dual/triple beam irradiation (JANNuS) (Ion E ~a few MeV): up to 10 4 appm He/dpa.Dual/triple beam irradiation (JANNuS) (Ion E ~a few MeV): up to 10 4 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 The Issue of additional Helium and of Testing structural Materials under Fusion relevant Conditions The Issue of additional Helium and of Testing structural Materials under Fusion relevant Conditions

32 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 32 of 35 slides 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.

33 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 33 of 35 slides Conclusions (2/2) Development of materials for high-temperature application: –SiC f /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 SiC f /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.

34 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 34 of 35 slides END

35 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 35 of 35 slides RAFM ODS Steel: HFR Neutron Irradiation 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

36 R. Lässer, 24 th SOFT, Sept 2006, Warsaw 36 of 35 slides Budget Sharing within the Materials Development Area 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%) 0.4 M/y Budget: ~ 3 M /year CEC


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