r. lässer, 24 th soft, 11-15 sept 2006, warsaw 1 of 35 slides structural materials for demo:...

R. Lässer, 24th SOFT, 11-15 Sept 2006, Warsaw

1of 35 slides

Structural Materials for DEMO: Structural Materials for DEMO: Development, Testing and ModellingDevelopment, Testing and Modelling

R. LR. Läässersser11, N. Baluc, N. Baluc22, J.-L. Boutard, J.-L. Boutard11, E. Diegele, E. Diegele11, , S. DudarevS. Dudarev33, M. Gasparotto, M. Gasparotto11, A. Möslang, A. Möslang44, R. Pippan, R. Pippan55, B. , B.

RiccardiRiccardi11 and B. van der Schaaf and B. van der Schaaf66

1 EFDA Close Support Unit Garching, D-85748 Garching, Germany2 CRPP-EPFL, CH-5232 Villigen-PSI, Switzerland

3 Euratom/UKAEA Fusion Association, Culham Science Centre, OX14 3DB UK4 Forschungszentrum Karlsruhe, 76021 Karlsruhe, Germany

5 Erich Schmid-Institute, Leoben, Austria6 NRG, Petten, The Netherlands


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ContentContentIntroductionIntroduction

– Path to DEMO

– European breeder blanket strategy

Materials DevelopmentMaterials Development– Similarities and differences of fusion and fission neutrons

– Neutron damage

Portfolio of structural materials for DEMOPortfolio of structural materials for DEMO– RAFM steel EUROFER

– ODS steels

– Tungsten and tungsten alloys

– SiCf/SiC

European Modelling Programme for irradiation effectsEuropean 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 conditionsIssue of additional He and testing under fusion relevant conditions

ConclusionsConclusions


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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- Tech- nology nology

R&DR&D

PhysicsPhysicsR&DR&D

Path to DEMO

ITERITER

DEMODEMO

IFMIFIFMIF


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Comparison: ITER, DEMO and Power ReactorComparison: ITER, DEMO and Power Reactor

ITER DEMO Reactor

Fusion Power 0.5 GW 2 – 2.5 GW 3 - 4 GW

Heat flux (FW) 0.1-0.3 MW/m2 0.5 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 10 - 15 MW.year/m2

Displacement per atom (dpa)

< 3 dpa 50 - 80 dpa 100 - 150 dpa

The challenge

Transmutation product rates at first wall

~10 appm Helium / dpa

~45 appm H / dpa


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Back plate 4

Back plate 3 Back plate 2

Back plate 1 Pb-17Li internal pipes

Pb-17Li inlet pipe+manifold

He inlet pipe

He inlet pipe

Pb-17Li outlet pipe+manifold

Pb-17Li distribution box

Breeder unit Stiffening grid

Vertical key way

Horizontal key way

Stiffening rod bolt

First wall

Stiffening rod

Cover

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 SiCf/SiC for insulating flow inserts,

Even Longer Term:

D) Self-cooled LiPb blanket with SiCf/SiC as structural material.

The European Breeder Blanket Strategy The European Breeder Blanket Strategy

First wall

HCLL TBM


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EUROFER

LiPb

300 / 500

W alloy

~ 540/720

He

W

Increasing Attractiveness

Increasing Development „Risk“

~ 600/990~ 540/720~ 540/720140/167Coolant I/O T(°C)

LiPbHeHeH2OCoolant

WWWWArmour material

SiCf/SiCW alloyW alloyCuCrZrStructural material

1.121.151.121.06TBR

LiPbLiPbLi4SiO4LiPbBreeder

700 / 1100480 / 700 300 / 480

300 / 500285 / 325Coolant I/O T(°C)

LiPbLiPb/

He

HeH2OCoolant

SiCf/SiCEUROFEREUROFEREUROFERStructural material

Model D or Self-cooled

Model C or Dual-Coola.

Model B or HCPB

Model A or WCLL

bla

nke

td

ive

rto

r

Materials for Blanket and Divertor according PPCS

Net efficiency 0.31 0.35 0.36 0.42 0.60

Materials listed above require R&D.

Model ABor HCLL

He

300 / 500

1.13


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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 drivenMaterials Development: Mission driven


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Fusion and Fission Neutrons in Materials: Fusion and Fission Neutrons in Materials: SimilaritiesSimilarities and and DifferencesDifferences

• In In fissionfission most neutrons have most neutrons have energies below 2 MeV. energies below 2 MeV.

• In In fusionfusion the D-T generated the D-T generated neutronsneutrons have have 14.1 MeV 14.1 MeV and are used 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 causeThe more energetic fusion neutrons cause far larger damagefar larger damage (multiple(multiple cascades)cascades) andand more transmutationmore transmutation products, e.g. H and He, (due to many new products, e.g. H and He, (due to many new nuclear reaction channels) andnuclear reaction channels) and higher activation higher activation than fission neutronsthan fission neutrons..

As a consequence: As a consequence: Structural materials developed for conventional fission have to Structural materials developed for conventional fission have to be improved and modified for fusion applicationbe 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.


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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)7 keV Cascade in Ni (fcc)

0,3 ps 0,7 ps 1,5 ps

2,5 ps 10,3 ps 10,3 ps

PKA

High energy PKAs in Fe (bcc)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:

Emax (n) = 2 MeV,

Emax (PKA)=0.14 MeV.

Fusion:

E (n) = 14.1 MeV, Emax (PKA)= 1 MeV.


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The Portfolio of Structural Materials for DEMOThe Portfolio of Structural Materials for DEMO Reduced Activation Ferritic Martensitic (RAFM) steel EUROFERReduced 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-RAFMODS-RAFM steelssteels 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.

SiCSiCff/SiC/SiC ceramic composites for advanced Breeding Blanket concepts ceramic composites for advanced Breeding Blanket concepts

ConConsidered in the long term for their potential to increase thermal efficiency (model D of PPCS). Operational T-window 600-1200°C. First use likely as functional material (flow channel inserts for Dual Coolant BB Concept).

Tungsten alloys for structural application in gas cooled divertorTungsten alloys for structural application in gas cooled divertor

Candidate material in the high temperature region of the gas cooled divertor concepts.


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EUROFER Alloy DevelopmentEUROFER Alloy DevelopmentDevelopment 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 Highly activating impurityimpurity elements ( elements (Nb, Mo, Nb, Mo, Ni, Cu, Al, Si, Co,..)Ni, Cu, Al, Si, Co,..) are reduced to the are reduced to the “lowest” content, that is technically “lowest” content, that is technically achievable at reasonable cost.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).


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Issues and Objectives identified for RAFM SteelsIssues and Objectives identified for RAFM Steels

1.1. Critical issues Critical issues at lower operational temperatureat lower operational temperature limit limit (300°C):(300°C):• Radiation hardening and embrittlement,• Additional effect of superimposed helium (He/dpa),• Reduction of the uncertainties in the DBTT and in fracture

toughness. 2.2. Critical issues Critical issues at upper operational temperatureat upper operational temperature limit limit (550°C)(550°C)::

• Thermal creep,• Compatibility (corrosion by Pb17Li above 500°C)

3.3. Detailed analysis of the irradiation data is being performed for a better Detailed analysis of the irradiation data is being performed for a better understanding of the understanding of the role of the major elements (Cr, Ta, W)role of the major elements (Cr, Ta, W) in order to in order to improve possibly the composition of the EUROFER and to focus the improve possibly the composition of the EUROFER and to focus the R&D on the most critical area (i.e. temperature range for the R&D on the most critical area (i.e. temperature range for the irradiation) leading to irradiation) leading to EUROFER-2EUROFER-2..

4.4. Production of EUROFER heats with Production of EUROFER heats with controlled impuritycontrolled impurity contents to contents to confirm the reproducibility of the properties and the low activation confirm the reproducibility of the properties and the low activation potential (potential (EUROFER-3EUROFER-3).).

5.5. The development of sound welds and dissimilar connections with The development of sound welds and dissimilar connections with improved properties requires post-heat treatment at about 730°C. improved properties requires post-heat treatment at about 730°C. More generally, particular attention should be paid on the welds and More generally, particular attention should be paid on the welds and joints development. This is necessary for the production of the TBMs.joints development. This is necessary for the production of the TBMs.


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Yield stress of unirradiated EUROFERYield 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


14of 35 slides

~ 200 K

-30%

TBM design window

Irradiation effects

Operational window

Unirradiated

Irradiated

Degradation of Impact Properties under neutron irradiationDegradation 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 Ductile-Brittle Transition

C. Petersen, FZK


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Embrittlement behavior of irradiated RAFM steelsEmbrittlement behavior of irradiated RAFM steelsIrradiation conditions: 16 dpa, 250 - 450°C

Higher DBTT are observed at lower irradiation temperatures (Tirr ≤ 300°C).

DBTT shift is of less concern for Tirr > 350°C.

250 300 350 400 450

-50

0

50

100

DB

TT

(°C

)

Tirr (°C)

EUROFER 97 ANL EUROFER 97 WB F82H-mod OPTIFER-Ia GA3X

E. Gaganidze, FZK

F82HEUROFER

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).


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• 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 PropertiesA potential Strategy for Recovery of Impact Properties

EUROFER 97

Unirradiated

Irradiated 15 dpa, 330°C

Irradiated andannealed

EUROFER 97

C. Petersen, FZK


17of 35 slides

RAFM Steels – Fabrication ProcessesRAFM Steels – Fabrication ProcessesComparative study on Post Weld Heat Treatments (PWHTs)Comparative study on Post Weld Heat Treatments (PWHTs)

LASER5 mm

TIG5 mm

EB5 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.


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0

2

4

6

8

10

12

-200 -100 0 100 200 300

Testing Temperature [°C]

Imp

ac

t E

ne

rgy

[J

]

0.0

0.2

0.4

0.6

0.8

1.0

1.2

La

tera

l Ex

pa

ns

ion

[m

m]

Eurofer97-TIG25 CEA, unirr.Eurofer97-TIG25 CEA, irr. SUMO-03, avg. 2.13dpa@300°CF82H-TIG15, unirr.F82H-TIG15, irr. CHARIOT-05, avg. 2.28dpa@300°CF82H-TIG15, irr. CHARIOT-07, avg. 8.94dpa@300°C

0

2

4

6

8

10

12

-200 -100 0 100 200 300

Testing Temperature [°C]

Imp

ac

t E

ne

rgy

[J

]0.0

0.2

0.4

0.6

0.8

1.0

1.2

La

tera

l Ex

pa

ns

ion

[m

m]

Eurofer97-EB25 CEA, unirr.Eurofer97-EB25 CEA, irr. SUMO-03, avg. 2.64dpa@300°C

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 Tirr = 300°C

Results on Irradiated EUROFER (Welds)Results on Irradiated EUROFER (Welds)

J.W. Rensman, NRG-FOM


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ODS Steels ODS Steels

Composition: EUROFER powder plus 0.3 wt% Y2O3. 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.

2 1

1 2

1

2

1

2 Pb-17Li

2

1

Potential use of ODS-Layer plated to FW


20of 35 slides

A) Creep Behaviour of ODS-EUROFER compared to EUROFERA) 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


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B) Impact Energy Values and Comparison with EUROFER-97 B) Impact Energy Values and Comparison with EUROFER-97 KLST SpecimensKLST Specimens

2nd Generation ODS-EUROFER

1st Generation ODS-EUROFER

EUROFER 971m

M23C6

1m

(FeCr)23C6

M. Klimiankou et al., FZK

M. Rieth, FZK


22of 35 slides

Tungsten and Tungsten AlloysTungsten and Tungsten Alloys

Potential application:Potential application:

W and W-alloys are promising materials for W and W-alloys are promising materials for high temperature structural applications, e.g. high temperature structural applications, e.g. the hottest part in the “high heat flux, high the hottest part in the “high heat flux, high temperature heat removal units” of gas temperature heat removal units” of gas cooled divertors. cooled divertors.

Requirement for this application:Requirement for this application:

Advantages of W-alloys:Advantages of W-alloys: High melting High melting point, high thermal conductivity and good point, high thermal conductivity and good thermal shock resistance, (thermal shock resistance, (low vapour low vapour pressure, high erosion resistancepressure, high erosion resistance).).

Critical issues of W-alloys: Critical issues of W-alloys: Creep rate and Creep rate and strength (700 to 1300°C), fracture toughness strength (700 to 1300°C), fracture toughness (RT to 1300°C), DBTT usually well above (RT to 1300°C), DBTT usually well above RT, ductility, recristallisation, low and high RT, ductility, recristallisation, low and high cycle fatique, low oxidation resistance above cycle fatique, low oxidation resistance above 490°C, behaviour under irradiation,490°C, behaviour under irradiation,

W tileW tile:: max. allow temp. 2500°C max. calc. temp. 1711°C

DBTT (irr.): 700°C

Thimble:Thimble: max. allow. temp. 1300°Cmax. 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 HEMJ 10 MW / m10 MW / m22

P. Norajitra, FZK

He-cooled Modular divertor with multiple jet Cooling


23of 35 slides

Strategy for Tungsten-alloysStrategy for Tungsten-alloysThis 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 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.

30 MPa√m


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Thermal Stability - Microstructure

HPT-W, HPT-W, , Before Before and after annealing at 1200°C for 1 h and after annealing at 1200°C for 1 h

1µm 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


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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 (1000- 1200°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).


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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), 2.65 g/cm3 (3D).

Specified properties for 2D and 3D SiCf/SiC composites were achieved.

3D composites (MT Aerospace)

fibrefibre

fibre

pore

fibre matrix


27of 35 slides

European Modelling Programme of Irradiation European Modelling Programme of Irradiation Effects in RAFM SteelEffects in RAFM Steel

Objectives:Objectives:

Study of the radiation effects in EUROFER under fusion relevant Study of the radiation effects in EUROFER under fusion relevant conditions from RT to 550°C and in the presence of high conditions from RT to 550°C and in the presence of high concentrations of nuclear transmutation products (i.e. H, He).concentrations of nuclear transmutation products (i.e. H, He).

Development of tools:Development of tools:o to correlate results from MTRs, fast reactors, spallation

sources, accelerators, fusion neutron sources, etc., o 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 Experimental validation of the models and the derived tools.the models and the derived tools.


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Radiation Modified MicrostructureScale and tools for multi-scale modelling

Rate Theory

Monte Carlo on Objects

Monte Carlo on Events

Ballistic Phase ThermalisationShort Term Recovery

DiffusionMicro--

structure

Primary Damage

Displacement Cascades

Molecular Dynamics

Short Term prediction

Molecular Dynamics

Atomic Kinetic Monte Carlo

10 -15 s 10 -11 s 10-8 s 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


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Radiation-induced HardeningScale and Tools for Multi-scale Modelling

1. Molecular Dynamics: void-dislocation interaction

40 n

m

2. Discrete Dislocation DynamicsOne dislocation Collective Behavior

5 µm

Increasing strain

TensileS

urf

ace

stra

in d

istr

ibu

tio

nAverage Tensile strain 5 %

Experiment

Computation

Average Tensile strain 9 %Experiment

Computation

3. Crystalline plasticity Finite Elements: low-alloy steel

Strain εS. Sekfali, PhD, 2003


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He Thermodynamics & DesorptionHe Thermodynamics & Desorption

He on substitu- He on substitutional sitetional site

He on interstitialHe on interstitial

(tetrahedral) site(tetrahedral) site

He on interstitialHe on interstitial

(octahedral) site(octahedral) site

E E Sol Sol = 4.22 eV= 4.22 eV E E SolSol = 4.39 eV = 4.39 eV E E Sol Sol = 4.57 eV= 4.57 eV

• Ab initio solution energies: substitutional He: the stable configuration

10-3

10-1

10 103

10-4

10-3

10-2

10-1

De

so

rbe

d f

rac

tio

n

Annealing Time (s)

559 K

577 K

667 K

Model

Experimental Ef (V) & Em (V) of Fe-C1: Ef (V) = 1.6 eV & Em (V) = 1.1 eV

• Ab initio energies: 1) Em(Hei) = 0.06eV. 2) Eb of Hei with point defects.

10-3

10-1

10 103

10-4

10-3

10-2

10-1

De

so

rbe

d f

rac

tio

n

Annealing Time (s)

559 K

577 K

667 K

Model

Ab initio for pure Fe:Ef (V) = 2.0 eV, Em (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


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Degradation of mechanical properties under irradiation strongly depends on Degradation of mechanical properties under irradiation strongly depends on irradiation temperature and the amount of transmutation products produced. 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, Fission neutrons produce about a factor of 40 less He than 14.1 MeV ones. Hence, irradiation in fission reactors gives only irradiation in fission reactors gives only non-conservativenon-conservative results. results.

Various tricks or methods were used to produce higher He and H to dpa ratios in the Various tricks or methods were used to produce higher He and H to dpa ratios in the absence of an intense fusion neutron source (absence of an intense fusion neutron source (B and Ni-doped steelsB and Ni-doped steels or or FeFe5454 enriched steelsenriched steels). Any of these methods is still short by about a factor of 10 to 5.). 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.Other facilities can be used, but also have their own shortcomings.• Mixed spallation-neutron spectrum in a spallation target: ~10Mixed spallation-neutron spectrum in a spallation target: ~1022 appm He/dpa, appm He/dpa,

– But due to many other transmutation products and inhomogeniety of irradiation conditions it is difficult to draw conclusions.

• Energetic (20-100 MeV) alpha particle implantation: ~10Energetic (20-100 MeV) alpha particle implantation: ~1033 -10 -1044 appm He/dpa. appm He/dpa.• Dual/triple beam irradiation (JANNuS) (Ion E ~a few MeV): up to 10Dual/triple beam irradiation (JANNuS) (Ion E ~a few MeV): up to 1044 appm He/dpa. appm He/dpa.

– But only few microns depth, so no mechanical testing, only microstructure.

Consequences:Consequences:(1 ) (1 ) Modelling and understanding of irradiation results obtained under various Modelling and understanding of irradiation results obtained under various

conditions are clearly needed.conditions are clearly needed.

(2) A fusion relevant neutron source is mandatory for a “correct” (2) A fusion relevant neutron source is mandatory for a “correct” characterisation of the materials in the sense that they become characterisation of the materials in the sense that they become licensable: licensable: IFMIFIFMIF

The Issue of additional Helium and of Testing The Issue of additional Helium and of Testing structural Materials under Fusion relevant Conditionsstructural Materials under Fusion relevant Conditions


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Conclusions (1/2) Conclusions (1/2)

Development of RAFM steels: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.


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Conclusions (2/2)Conclusions (2/2)

Development of materials for high-temperature application: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: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.


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ENDEND


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RAFM ODS Steel: HFR Neutron IrradiationRAFM 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

0 5 10 15 20 250

200

400

600

800

1000

1200

Eurofer 97 (FZK)T

test = T

irr = 250°C

irradiated, 15 dpa unirradiated

Str

ess

[MP

a]

Strain [%]

E. Materna-Morris and R. Lindau, FZK


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Budget Sharing within the Materials Development AreaBudget Sharing within the Materials Development Area

Materials Development R&D 2002-2005

Nuclear Data10%

SiC/SiC 8%

EUROFER47%

Modelling7%

RAFM ODS13%

Tungsten8%

200518 %

2005: Peak 23 % to start irradiations

Type of ContractType of Contract Materials DevelopmentMaterials Development

[Average][Average]

Breeding BlanketsBreeding Blankets

[Average] (for comparison)[Average] (for comparison)

R&D (20-40%)R&D (20-40%) 9.0 M€/y9.0 M€/y 8.5 M€/y (25 M€ in 2002)8.5 M€/y (25 M€ in 2002)

Studies (40%)Studies (40%) 0.4 M€/y0.4 M€/y 0.2 M€/y0.2 M€/y

Industrial contracts (100%)Industrial contracts (100%) 0.4 M€/y0.4 M€/y 0.4 M€/y0.4 M€/y

Budget: ~ 3 M€ /year CEC

r. lässer, 24 th soft, 11-15 sept 2006, warsaw 1 of 35 slides structural materials for demo:...

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