THE EUROPEAN R&D PROGRAMME ON DIVERTOR ARMOR MATERIALS …€¦ · Develop models for neutron radiation effects, specifically microstructural evolution and embrittlement, in iron

THE EUROPEAN R&D PROGRAMME ON DIVERTORARMOR MATERIALS AND TECHNOLOGY – STATUS ANDSTRATEGY

M. Rieth, S. Antusch, J. Hoffmann, M. Klimenkov, J. Reiser, S. Brezinsek, W. Biel, J. Coenen, J. Linke, Ch. Linsmeier, A. Litnovsky, Th. Loewenhoff, G. Pintsuk, B. Unterberg, M. Wirtz, H. Greuner, A. Kallenbach, R. Neu, J. Riesch, J.H. You, T. Barrett, F. Domptail, S. Dudarev, M. Fursdon, M. Gilbert, A. Galatanu, L.M. Garrison, Y. Katoh, L.L. Snead, D. Armstrong, S. Roberts

Budget (w/o overheads)

2 1st IAEA Technical Meeting on Divertor Concepts | IAEA HQ | VIC | Vienna | 29.09.-2.10.2015

M. Rieth et al.: EUROfusion WPMAT HHFM - Status & Strategy

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1 2 3 4 5

Hardware (EC/k€)Total Allocated2014‐18

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2700

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Manpower (EC/k€)Total Allocated2014‐18

Total Allocated Resources

273 lab ppy30 ind ppy8.74 M€ hw

today

Overall Objectives 2014-2018



Fill gaps in the database and develop design codes for the baseline materials

Development of new materials to mitigate requirements of advanced DEMO component designs

Demonstration of the production of such materials in processes scalable to industrial standards

Characterization of the properties of such materials

Develop models for neutron radiation effects, specifically microstructural evolution and embrittlement, in iron alloys, steels, tungsten, and degradation of functional materials

Perform neutron irradiation experiments

focus of this presentation: armour

Overview of HHF Materials



W (PIM) W PLANSEE (rolled)

Ductile @ 200 °CFracture

W‐2Y2O3W‐1TiC W‐2La2O3

W Alloys (by PIM)




W

CuCrZr

CuCuCuCu

W

WW

Cu/CuCrZr‐W Pipes/Laminates

Composites Cu/CuCrZr‐W‐fiber Pipes W‐W‐fiber Blocks

CuCrZr




W‐Cu/CuCrZr Composite




Material Characterisation• microstructure, chemical analysis• physics: heat conductivity (diffusivity & heat capacity)• strength: tensile, bending (3 pt., 4 pt.)• toughness: bending (fracture mechanics, DBTT, strain rate effect)• HHF tests

mockup (JUDITH & GLADIS) thermal shock (several FZJ facilities)

4 mm

4 mm

Divertor assessment methodology


• A suitable method of divertor PFC assessment is a significant issue to be resolved

– Current nuclear design codes neither consider multi-layer/multi-material structures nor stress due to manufacturing

– Instead of “design by analysis”, ITER have used “design by experiments” including intensive HHF testing

• Bespoke criteria based on elasto-plastic analyses are currently under development in EUROfusion

• The immediate need to facilitate PFC design optimisation was a standardised analysis procedure using linear elastic code rules

Typical Mock‐Up

Thermal analysis

Stress analysis

1st IAEA Technical Meeting on Divertor Concepts | IAEA HQ | VIC | Vienna | 29.09.-2.10.20158

Standard thermo-mechanical analysis


• “Monoblock Elastic Analysis Procedure” - “MEAP”• Coherent/consistent analyses across EU DIV project• Reserve factors (margin to failure) to 5 Rules are

calculated, enabling ranking of design concepts• 2 structural rules, which are valid despite the

considerable residual stress field, and 3 thermal rules• 10 MW/m2 surface heat flux is used for analyses• NOT a method for “absolute” failure assessment!

MEAP rule Rule details

Rule #1Ratchetting (3Sm) first check, runaway ratchetting is not expected in reality*. Requires material Sm data.

Rule #2 Fatigue – following IC3132.3.1*. Requires material cyclic ‐ and ‐n data.

Rule #3For a CuCrZr pipe, maximum temperature <300°C to avoid creep/softeningunder irradiation.Minimum temperature > 150°C to limit embrittlement.

Rule #4 Maximumwall heat flux < device Critical Heat Flux (burnout)Rule #5 Maximum tungsten armour temperature <1800°C , to limit recrystallisation* Reference: ITER Structural Design Criteria for In‐vessel Components, ITER G‐74‐MA‐8‐01‐05‐28‐W‐0.2, 2012.


Outline



How to develop divertor materials without knowing the critical limits, which are closely connected to the design, which in turn is under development, too?

What are the relevant properties for divertor armour materials?

Do we perform appropriate assessment tests?

Load Analysis, Basic Properties



Thermal off‐normal 20 MW/m2

#1 #2 #100002 fpy

normal 10 MW/m2

t

HF 0.2‐1 GW/m2 ELM

15 MW/m2 (30s), small monoblock23 mm x 22 mm x 4 mm, D15 mm

water cooling, 200 °C M. Li, IPP

10 MW/m2 + 0.4 GW/m2 (1ms on, 50ms off)ITER type monoblock

28 mm x 28 mm x 12 mm, D17 mm

0

1800

Tempe

rature

(°C)



Mechanical, elasticnormal

t

HF

10 MW/m2

10s

path

‐1200

+1200

path (mm)

Temperature (°C)

Stress (MPa)

Yield Limit (MPa)

Sxx plastic deformationunder compression


+740 MPa

‐670 MPa

Stress




Mechanical, plastic normal

t

HF

10 MW/m2

10s 20sPlastic Deformation (10 s)

3.1E‐3

0

Pl. Strain

1.1E‐3

2.1E‐3

Secondary Stresses (20 s)

365 MPa

‐110 MPa

Stress

0

200 MPa




Mechanical, plastic normal

t

HF

10 MW/m2

10s 20s

Max. Stress during cooling

• Strain rate: 10‐3/s – 10‐2/s• 150 °C < DBTT <250 °C• T > 300 °C ductile regime no brittle fratcure

Mechanical, dynamic




Thermal: Tsurf = 1790‐2112 °C Recrystallisation (Rxx) off‐normal

t

HF

20 MW/m2

10s 20s

Dynamical: DBTT 250‐350 °C (W‐Rxx)ductile‐brittle regime

Fracture Mechanicso KIc = 5‐8 MPa m1/2 (W‐Rxx)o critical crack length = 16‐40 µmo grain size: min. 50 µm, max. >300 µmductile/brittle crack formation likely

Mechanicalo plastic surface deformation > 1%

during heatingo secondary surface tensile stresses

after cooling down: > 710 MPa

710 MPa

‐973 MPa

Stress

‐500 MPa

0




Thermal

normal

t

HF

10 MW/m2

1ms 50ms

0.2 GW/m2

time

300

1500

path (mm)

Tempe

rature

(°C)

Mechanical

strain rates > 1/s

‐200

350




normal

t

HF

10 MW/m2

1ms 50ms

0.2‐1.0 GW/m2

210 MW/m2, 1 ms 10 MW/m2, 51 ms

Mechanical Surface Area

only a thin layer (500 µm) is loaded with high strain rates (> 1/s) at 1000‐1400 °C plastic surface deformation under tensile and compression without immediate

damage Tmax: 0.2 GW/m2 – 1400°C, 0.4 GW/m2 – 1750°C, 0.6 GW/m2 – 2100°C, 0.8 GW/m2 –

2500°C, 1 GW/m2 – 2900°C (1 mm surface layer with 1‐5% deformation) Recrystallisation

785 MPa

‐385 MPa

Stress

0

760 MPa

‐350 MPa

Stress

0




10 MW pulse + 1 GW/1 ms stress after cooling down to 300 °C

off‐normal events could be moredemanding compared to ELM discharges

20 MW/m2 pulse on and off

710 MPa

‐973 MPa

Stress

‐500 MPa

0

704 MPa

‐313 MPa

Stress

0

First Conclusions



Comparison with experiment: ITER mockup, 300 x 20 MW/m2 pulses

lower operating temperature (pulse off period) is decisive

HHF mockup tests modification (coolingtemperature, in‐situ crack detection, ...)

relevant material properties for thisparticulare case: Recryst.‐Temp. & static DBTT (W‐Rxx) & initial cracks

6‐7.5 mm

Critical operating point for cracking

>1700°C

t

Surface Temperature20 MW/m2

compression

10s 20s

<100°CDBTT

0 MW/m2

tensile stress

Additional Material Properties



Fatigue#1 #2 #10000

2 fpy

normal 10 MW/m2

t

HF 0.2‐1 GW/m2 ELM

0

t

stress t

T

o Normal pulse operation would correspond to asymetric Low Cycle Thermal Fatigue conditions

o but ELM discharges correspond to high strain rate, high temperature, High Cycle Thermal (Shock) Fatigue

this is not covered by usual material tests (LCF, LCTF, HCF, etc.) current thermal shock tests need (slight) modification (?)

Changing Conditions During Operation



Microstructure o Recrystallisationo Cyclic plastic deformation Strength, DBTT, toughness, ductility and fatigue properties will vary during operation

Geometry o Erosion of armour surface due to PSI (max. 2‐3 mm/fpy ?, depending on T)

Temperature and load will vary with time

Irradiation o Continuous increase of damage (2 dpa/fpy in striking point, 4 dpa/fpy elsewhere)

o Damage strongly depends on temperature and neutron spectrum

Inhomogeneous variation of material properties Strength, DBTT, toughness, ductility, thermal conductivity, fatigue properties, microstructure, density (swelling), erosion rate?, others?

Irradiation Defects



Void Formation(Swelling)

Transmutation into Rhenium

o About 0.5‐1% Rhenium/fpy Embrittlement, loss of conductivity, … Extend unknown (no experiments with appropriate n‐spectrum available)

o Max. at 600‐900°C Embrittlement, loss of conductivity, …

Outlook: Preliminary Answers



How to develop divertor materials without knowing the critical limits, which are closely connected to the design, which in turn is under development, too?

What are the relevant properties for divertor armour materials?

Do we perform appropriate assessment tests?

Close cooperation with divertor community with the goal to minimize number of concepts, geometries, operating and boundary conditions, …

??? There is no simple answer. It depends on too many issues …

In principle, yes ‐ but …(1) Variation of monoblock tests would be interesting: cooling temperature, crack detection, …(2) Thermal shock tests should be re‐assessed for possible extension improved material assessment

However, the relevance of tests is closely connected to the design. Different concepts might require different tests.

Finally ...



Tungsten (armour) is not the only material involved

There are interfaces (e.g. W‐Cu, CuCrZr‐Cu) Composites

There are structural parts

There are additional boundary conditions …

Each additional material requires separate assessment for microstructural change, geometry, and irradiation damage !!!

WPMAT strategy: iterative approach (interaction with divertor & plasma community)

Development of a broad variety of different HHF materials& composites WITHOUT predefined target properties

Thank you !


M. Rieth et al.: EUROfusion WPMAT HHFM - Status & Strategy25

Partners:

and joined universities

Disclaimer: „This work has beencarried out within the frameworkof the EUROfusion Consortiumand has received funding from theEuratom research and trainingprogramme 2014‐2018 undergrant agreement No 633053. The views and opinions expressedtherein do not necessarily reflectthose of the European Commission.“

Additional Slides



Mass production of W partsand W alloys development



Achievements in HHFM

Monoblocks with various shapes Samples for ASDEX Upgrade Langmuir probes for WEST

HHF Tests in JUDITH: PLANSEE pure tungsten according to ITER specifications (‟IGP”) compared to PIM W alloys

PIM: W-2La2O3 PIM: WPIM: W-2Y2O3

longitudinal transversal recrystallized

# T [°C] Pabs [GW/m2] Δt [ms] Eabs [MJ/m2] FHF [MW/m2*s1/2] # shots

°C 1000 0.38 1 0.38 12 1000

100 µm 100 µm100 µm

100 µm 100 µm 100 µm



Achievements in HHFM

„Self-Castellation in ITER“



off‐normal

t

HF

20 MW/m2

10s 20suncracked monoblock cracked monoblock

crack from top to interlayer(free surface)

Irradiation Defects



Tungsten

Tungsten‐5% Rhenium(in solid solution)

J. Linke, FZJ

Documents

THE EUROPEAN R&D PROGRAMME ON DIVERTOR ARMOR MATERIALS …€¦ · Develop models for neutron radiation effects, specifically microstructural evolution and embrittlement, in iron