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Vladimir Barabash and the ITER International Team, A. Peacock2, S. Fabritsiev3, G. Kalinin4, S. Zinkle5, A. Rowcliffe5,
J.-W. Rensman6, A.A. Tavassoli7, P.J. Karditsas8, F. Gillemot9, M. Akiba10
1 ITER International Team, 85748 Garching, Germany2 EFDA Close Support Unit, 85748 Garching, Germany
3 D.V. Efremov Scientific Research Institute, 196641 St. Petersburg, Russia4 ENES, P.O. Box 788, 101000 Moscow, Russia
5 ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138, USA6 NRG, P.O. Box 25, 1755 ZG Petten, The Netherlands
7 CEA/Saclay, 91191 Gif sur Yvette Cedex, France8 EURATOM/UKAEA Fusion Association, Abingdon, Oxon, OX14 3DB, UK
9 AEKI Atomic Research Institute, 1121 Budapest, Hungary10 JAEA, Naka-machi, Naka-gun, Ibaraki-ken 311-0193, Japan
Materials Challenges for ITERMaterials Challenges for ITER
-- Current Status and Future ActivitiesCurrent Status and Future Activities
2
Outline
ITERITER-- Status of projectStatus of project-- Materials activity strategyMaterials activity strategy
Materials selection for different componentsMaterials selection for different components-- Vacuum vesselVacuum vessel-- InIn--vessel componentsvessel components
-- Stainless steel 316L(N)Stainless steel 316L(N)-- CuCrZr alloyCuCrZr alloy-- Alloy 718Alloy 718-- TiTi--6Al6Al--4V4V-- NiAl bronzeNiAl bronze-- Armour materials (Be, W, CFC)Armour materials (Be, W, CFC)
Test Blanket Modules ActivitiesTest Blanket Modules ActivitiesSummarySummary
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
3
Status of ITER
ITER is international fusion energy research project. ITER is international fusion energy research project.
ITER goals:
Show scientific and technological feasibility of fusion energy for peaceful purposes
Integrate and test all essential fusion power reactor technologies and components
Demonstrate safety and environmental acceptability of fusion
ITER activities were started in 1992, Final Design Report was issued in 2001.
Extensive negotiations about construction agreement, selection of site, sharing between Parties, etc. are being carried out during 2001-2005.
Finally, in June 2005, the ITER construction site was selected – Cadarache, France.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
4
ITER Design Parameters (FDR 2001)
Total Fusion Power 500 (700) MW
Q=Fusion Power/Heating power
10
Plasma major radius 6.2 mPlasma minor radius 2.0 mToroidal field, 6.2 m 5.3 TPlasma current 15 (17.4) MAPlasma Volume 837 m3
Plasma Surface 678 m2
Neutron flux Av. 0.57 MW/m2
Max. 0.8 MW/m2
Neutron fluence Av. 0.3MW*a/m2
Max 0.5MW*a/m2
Heat flux, First Wall
0.2 - 0.5 MW/m2
Heat flux, Divertor 10 (20) MW/m2
Number of pulses ~ 30.000Pulse length ~ 400 s
ITER is the first tokamak, with significant neutron fluence. ITER is the first tokamak, with significant neutron fluence. ITER is the first tokamak, with significant neutron fluence.
5
Selection of materials for the ITER components was a challenging task, due to unique operational conditions, e.g.:
Selection of Materials
• Magnet – operation at 4K, high stresses;
• In-vessel components – combined effect of n-irradiation, heat flux, hydrogen environment, mechanical loads, complicated design with different type of joints;
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
Experience from current tokamaks has been taken into account,(plasma facing, diagnostics materials, etc.).
Knowledge from fission neutron irradiation programs was used.
However, in many cases, the choice was not simple, because for anumber of materials the existing data base was not sufficient.
6
The material choice for ITER has been orientated toward industrially available materials while taking account of physical and mechanical properties, maintainability, reliability, corrosion performance and safety requirements at the ITER operational conditions.
Selection of Materials – General strategy
The ITER materials categories:The ITER materials categories:
• Standard materials with established manufacturing technologies, (e.g. steel 304, steel 316L(N), alloy 718, W).
• Standard materials, which require some modifications, such as more stringent limits on the alloying elements, etc. (e.g. steel EK1, JJ1, CuCrZr, etc. ).
• Recently developed materials (e.g. CFC).
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
Materials were selected based on the evaluation of the impact of manufacturing technologies on materials properties and assessment of their performance at the ITER operational conditions.
These selection and justification were supported by the results of the dedicated world-wide R&D program.
7
Materials documentation – FDR 2001 (2004)ITER Materials Properties Handbook (MPH-Magnet, Cryo, VV&IC):- A collection of design-relevant data on physical and mechanical properties of a large variety
of the ITER relevant materials.
ITER Materials Assessment Report (MAR):- A description of the rationales for the selection of the specific materials grades for VV and In-
vessel components and assessment of the materials performance.
Appendix A of the ITER Addendum to the RCC-MR:- Design data for materials for the Vacuum Vessel.
Appendix A of the ITER Structural Design Criteria (SDC-IC):- Design data for in-vessel components (including neutron irradiation).
Safety Report:- Collection of safety relevant properties.
ITER Materials Properties Database (MPDB):- Collection of raw experimental data - base for further recommendations
These documents justify the selection and materials performance.These documents justify the selection and materials performance.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
Codes, R&D reports,Open sources
Codes, R&D reports,Open sources
Database(MPDB)
Database(MPDB)
ITER MPH,Experts Activity
ITER MPH,Experts Activity
Design DocumentsDesign
Documents
8
Materials Activity in current stage
Multidisciplinary task:Multidisciplinary task:
• Preparing of the procurement specifications for materials for various components (in accordance with construction schedule); - Discussion with Industries for implementing the ITER requirements- Discussion of the possible deviations- Schedule for materials procurements
• Consolidation of the materials properties database and modifications of the ITER Materials Documents;- Assessment of the new data from R&D program- Preparing of recommendations based on Design Codes requirements- Completion of MPH and App. A for VV Code, In-vessel SDC and Magnet SDC
• Carrying out the R&D in some still critical areas:- Further optimisation and simplification of design and manufacturing route- Generating of new data needed for the completion of the design evaluation
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
9
No. WBS Element 1. TOKAMAK BASIC MACHINE 1.1 TOROIDAL FIELD (TF) COILS SYSTEM 1.2 POLOIDAL FIELD (PF) COILS SYSTEM 1.3 CENTRAL SOLENOID (CS) SYSTEM 1.5 VACUUM VESSEL 1.6 BLANKET SYSTEM 1.7 DIVERTOR 2. TOKAMAK ANCILLARIES AND CRYOSTAT 2.3 REMOTE HANDLING (RH) EQUIPMENT 2.4 CRYOSTAT 2.6 COOLING WATER SYSTEM 2.7 THERMAL SHIELDS 3. TOKAMAK FLUIDS 4. POWER SUPPLIES - COMMAND CONTROL 5. PORT INTERFACING SYSTEMS 5.1 ION CYCL. HEATING (IC H&CD) SYSTEM 5.2 ELECTRON CYCL. HEATING (EC H&CD) SYSTEM5.3 NEUTRAL BEAM HEATING (NB H&CD) SYSTEM 5.4 LOWER HYBRID HEATING (LH H&CD) SYSTEM 5.5 DIAGNOSTICS 5.6 TEST BLANKETS
Selection of materials for ITER components
Magnet Magnet –– T operationT operation-- 44--77 K77 K
Coating, 5 μm (emissivity)Ag coating
InsulationGlass epoxy G10
Plasma sprayed insulationAl2O3 coatings
BoltsAlloy 718
FastenersSteel grade 660
PlatesTi-6Al-4V
Main structural material: - plates, tubesSteel 304L
Thermal Shield Materials (77Thermal Shield Materials (77--300K)300K)
Vacuum vessel and In-vessel materials: - structural - plasma facing
1.5 VACUUM VESSEL
1.6 BLANKET SYSTEM
1.7 DIVERTOR
5.1 ION CYCL. HEATING (IC H&CD) SYSTEM
5.2 ELECTRON CYCL. HEATING (EC H&CD) SYSTEM
5.3 NEUTRAL BEAM HEATING (NB H&CD) SYSTEM
5.4 LOWER HYBRID HEATING (LH H&CD) SYSTEM
5.5 DIAGNOSTICS
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
10
Materials for Vacuum Vessel:Austenitic steel 316L(N)-IG plates, forgings, rodsBorated steels 304B7 and 304B4 shielding platesFerritic steel 430 platesSteel 660 fasteners, forgingsSteel 304 plates, beamsAlloy 718 boltsWeld filler materials electrodes, wireXM-19 bolts for shieldingSteel 316 bolts (B8M)Pure Cu clad on VV
Materials for Vacuum Vessel support:Steel 304 plates, forgings, rodAlloy 718 boltsSteel 660 forgings, boltsNiAl bronze pinsPTFE for sliding elementsNeoprene rubber
Selection of materials – Vacuum Vessel- Safety Important Component => licensing in France- Code – RCC-MR + ITER addendum- Main structural material – 316L(N)-IG is the Code qualified and main specifications are available- Other structural materials: steel 304, steel grade 660, Alloy 718 are also in the Code- Only functional materials (borated steels, etc.) are not in the Code.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
11
Selection of materials – Vacuum Vessel
Element
316L
ASME A240
Z2 CND 17-12 controlled nitrogen
RCC-MR 2002
316L(N)-IG ITER grade
min max min max min max C 0.03 0.030 0.03
Mn 2.00 1.60 2.00 1.60 2.00 Si 0.045 0.50 0.50 P 0.030 0.035 0.025 S 0.75 0.025 0.01
Cr 16.00 18.00 17.00 18.00 17.00 18.00 Ni 10.00 14.00 12.00 12.50 12.00 12.50 Mo 2.00 3.00 2.30 2.70 2.30 2.70
Nb+Ta+Ti 0.15* Cu 1.00 0.30 B 0.0020 0.002
(0.001)** Co 0.25 0.05 N 0.1 0.060 0.080 0.060 0.080
Sm, RT
115
147
147 Nb < 0.1, Ta < 0.01
Materials for VV – ongoing activity:Complete App. A to ITER Addendum to RCC-MR
- Introduce required additional properties (e.g. fracture toughness) for structural materials- Introduce properties of functional materials in the ITER Addendum
Non-metallic materials – justification of the properties for licensing- Windows, etc.
Properties of welds- In case of use welds, which are not in Code, – provide the needed justification.
Preparation of the Procurement Specifications- Composition, minimum properties, NDE&QA
Chemical composition of steels
0
20
40
60
80
100
120
140
160
0 50 100 150 200 250 300 350 400 450
Temperature, °C
Des
ign
Stre
ss In
tens
ity S
m, M
Pa
316L(N)-IG
316L, Sec. II, Part D, 2001
316LN, Sec. II, Part D, 2001
316L(N)-IG, RCC-MR, Ed. 2002
316L, ASME Sec. II, Part D, 2004
316LN, ASME Sec. II, Part D, 2004
12
General considerations: • Design of structural elements – in accordance with the ITER Structural Design
Criteria for IC (SDC-IC)• Design of some elements (Be/Cu, CFC/Cu, Window/Cu etc.) – by experiment
ITER SDC-IC:• Needs reliable and traceable materials data (including neutron irradiation effect) • Properties of joints (welds, SS/Cu, etc.) have also to be assessed
ITER MPH includes the justification of recommendations.
Properties needed for design analyses:- Physical (incl. neutron irradiation) - average (+deviation)- Tensile properties for design (incl. neutron irradiation) - minimum and average- Design fatigue curves (incl. neutron irradiation) - minimum (∆ε/2; Nf/20)- Fracture toughness (incl. neutron irradiation) - minimum and average
Specifications for materials supply: - Chemical composition,- Minimum properties;- NDE&QA
Selection of materials – In-vessel components
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
13
Selection of materials – In-vessel components
Square waveguides Front mirrors
Windows~ 4 m
Material Material Grade Components • Armour Beryllium S-65C VHP or equivalent • Armour for first wall and
limiter Tungsten Pure W • Armour for divertor CFC SEP NB 31 or equivalent • Armour for divertor
• Structural 316L(N)-IG • Blanket shield modules
• Thin walled pipes • Cooling manifolds • Divertor body
Austenitic steels
304L, 316L • Cooling pipes XM-19 • Divertor support PH steel Steel grade 660 • Divertor support
CuCrZr • PFCs, heating systems, electrical strips, etc.
NiAl bronze • Divertor attachment CuNiBe • Support system
Cu alloys
DS Cu • heating systems,
• Functional Austenitic steel
316 • Fastening components
PH steel Steel grade 660 • Fastening components Ni alloys Alloy 718 • Bolts
• Divertor connections Ti alloy Ti-6Al-4V • Blanket attachment Ceramic Al2O3 or MgAl2O4 • Electrical insulators Special materials
• Pure Cu • Pure Ni • Low C iron • Brazes • Weld fillers
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
14
Stainless steel 316L(N)-IG
ITER recommendation for the reweldability limits, provided by the allowable levels for He production in weld:
– < 1 appm for thick plate welding,– < 3 appm for thin plate or pipe welding.
Rewelding
K. Asano, J. van der Laan, MAR, 2001V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
Different types of steels (different treatment, etc.):- Wrought- HIP- Powder HIP - have to be qualified!- Cast- Welds
Welding and joining
100
200
300
400
500
600
700
800
0
10
20
30
40
50
60
0 5 10 15 20
Weld Metal (19-12-2)Irr and test - 400°C
Elon
gatio
n, %
Dose, dpa
U.T.S
Y.S.
T.E.
U.E.
A.A. Tavassoli, this Conf.
• Un-irradiated data for welds are Code qualified;• Irradiated data are being reviewed, the behavior
is similar to base metal; the same safety margin can be used;
• The properties of joints after HIP (solid and powder) joining technique are adequate.
15
0
5
10
15
20
25
30
35
40
45
0 50 100 150 200 250 300 350 400
T test = T irr, oC
Spe
cific
atio
n U
EL
min
, %
unirradiated
2 dpa
5 dpa9 dpa
irradiation effect
Stainless steel 316L(N)-IG
Neutron effect: Tensile properties:-Strengthening and loss of ductility (loss of strain hardening)
Fatigue:– no effect on fatigue
Fracture toughness:– reduction, but material still ductile
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
0
100
200
300
400
500
600
700
800
0 1 2 3 4 5
D, dpa
Spe
cific
atio
n S
y, m
in, M
Pa
100oC200oC300oC
0.1
1.0
10.0
1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06 1.00E+07
Fatigue life Nf
Tota
l str
ain
rang
e %
RCC-MR ed. 2002 20°C design curveUnirradiated data Ttest ≤ 250°C[2] Källstrom, Tirr = 35°C, Ttest = 25+250°C, 0.3 dpa[3] Horsten, Tirr = Ttest = 227+327°C, 7.5-11.2 dpa[4] Horsten, Tirr = Ttest = 77°C, 0.58-0.83 dpa[5] Lind, Tirr = Ttest = 290°C, 0.73+2.5 dpa[6] Puzzolante, Tirr = 42°C, Ttest = 25°C, 5.1-5.4 dpa[7] Bergenlid, axial, Tirr = 250°C, Ttest = 200°C, 2.8+9.2 dpa[8] Vandermeulen, Tirr = Ttest = 250°C, 6.2-10 dpa Best fit to unirradiated data (aNf^b)Best fit to irradiated data (aNf^b)Fit irr data displaced by Nf/20Fit irr data displaced by eps/2
0
100
200
300
400
500
600
700
0 100 200 300 400 500
Temperature, °C
Min
imum
toug
hnes
s va
lues
Lower Bound JIC (kJ/m2)
3 dpa = *.76
5 dpa = *.61
J.W. Rensman, MPH 2004&2005
J.W. Rensman, MPH 2005
G. Kalinin, MPH 2005
16
Alloy Designation Cu Cr Zr O Other elements
CuCrZr, RF base 0.4-1.0 0.03-0.08 0.1 max CuCrZr, C18150 base 0.5-1.5 0.05-0.25 CEN/TS 13388 base 0.3-1.2 0.03-0.3 0.2 CuCrZr ITER Grade base 0.60-0.90 0.07-0.15 0.002 max 0.01 max
0.05 -Cd
Chemical compositionChemical composition
Reasons for modification of the chemical composition:- smaller scatter of properties, - less coarser Cr particles – better toughness, - better radiation resistance is expected,- welding improved.
CuCrZr alloy
Effect of treatment on properties.
Manufacturing treatments:(SA – 980-1000°C + water quench, Ageing – 450-500°C• SAA (different cooling rate, etc.)• SA + cold work + A• SAA + cold work• cast CuCrZr + SAA
0
100
200
300
400
0 100 200 300 400 500
Temperature, °C
Su m
in a
nd S
m, M
Pa
Su, min, SAA
Su, min SA+560C/2hSm, = 1/3 Su, min (SAA)
Sm=1/3Su min (SA+560C/2h)
Barabash, Zinkle, MPH 2005
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
17
CuCrZr alloy
Fatigue: - Seems that fatigue performance is similar for CuCrZr with different treatment
Creep – fatigue interaction:– very complicated performance:
=> Generally there is reduction of lifetime with hold time; => Higher Sy – higher fatigue lifetime
(see presentation of B. Singh, this Conference)
1.0E-02
1.0E-01
1.0E+00
1.0E+01
1.E+02 1.E+03 1.E+04 1.E+05 1.E+06
Number of cycles, Nf
Tota
l str
ain
rang
e, %
Design fatigue curve, SAA, 20 - 350°CExperimental data, CuCrZr SAA, 20 - 350°CSingh, SA+underageing, 250-350C°, unirradiated
Average limited data, FDR 2001Design curve for pure annealed Cu
Design fatigue curve:- min of average of delta et/2 and Nf/20
M. Li, P. Marmy, B. Singh, MPH 2005
I. Bretherton, 2001, MPH 2005
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
18
CuCrZr alloy
Neutron effect: Tensile properties:- loss of ductility (loss of strain hardening):
saturation at ~ 0.1-0.5 dpa at T ~ 100-300°C. - Some improvements of ductility is observed
after baking treatment (see Fabritsiev, this Conference)
Fracture toughness and fatigue:– effect is small, but data are only up to 0.3 dpa
There is no correlation between loss of ductility at tensile tests and fracture toughness.
0
100
200
300
400
500
0 50 100 150 200 250 300 350 400
Temperature (oC)
valu
e of
J in
itiat
ion
J Q (k
Jm-2
)
SEN(B) B=3mm W=4mm [4-6]
SEN(B) B=3mm W=4mm [7]
Polyn. fit (all)
Polyn. fit (unirrad.)
0.3 dpa
S. Tähtinen, MPH 2005
0
100
200
300
400
500
0.0 1.0 2.0 3.0
Damage dose, dpa
Su ir
r and
Sy
irr,
MPa
Su, irr, T = 150°C
Sy, irr, T = 150°C
Su, irr, T = 300°C
Sy, irr, T = 300°C
V.Barabash & S.Zinkle, MPH 2005
0
5
10
15
20
0.0 1.0 2.0 3.0
Damage dose, dpa
Uni
form
elo
ngat
ion,
%
UA, T = 150°C
UA, T = 300°C
ISDC limit 2 %
Summary for CuCrZr:• Final characterization after selected
heat treatment (s);• There are some missing data:
irradiation creep, fracture toughness and fatigues at high doses;
• Importance of creep-fatigue interaction shall be understood.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
19
Alloy 718
Key issues:
Stress relaxation -bolts:- ITER goal ~ 0.1 dpa
Loss of strength at low damage dose:– effect is ~ 15%.
J.W. Rensman, MPH 2005T.S. Byun, ORNL
y = e-0.694*x
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
J.W. Rensman, MPH 2005
20
Reasons: low elastic modulus and high strength provide high flexibility.
Ti-6Al-4V alloy
Key issues: Neutron irradiation embrittlement:- ITER goal ~ 0.1 dpa
Hydrogen environment:– it is well known that H affects properties.
0
200
400
600
800
1000
1200
0.0 0.1 0.2 0.3 0.4 0.5
Damage dose, dpa
Stre
ngth
, M
Pa
Su, 50°C Sy, 50°C
Su, 260°C Sy, 260°C
Su, 350°C Sy, 350°C
0
5
10
15
20
25
0.0 0.1 0.2 0.3 0.4 0.5
Damage dose, dpa
TE a
nd U
E, % TE, 50°C UE, 50°C
TE, 260°C UE, 3260°C
TE, 350°C UE, 350°C
MPH 2005: S. Tahtinen, B. Rodchenkov
Hydrogen levels up to 150 wppm have moderate effect on toughness, at a test temperature of 150°C. P. Marmy, 2004
0.12 – 0.15 dpa
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
However, remaining issues:
• Role of ionizing radiation on possible enhanced hydrogen uptake (MIT launcher, May 2005)
• Fracture toughness and fatigue of irradiated Ti alloy.
21
NiAl bronze (NiAl bronze (CuAl10Ni5Fe4)CuAl10Ni5Fe4)::ASTM B150, C63200DIN EN 12163/1216
High strength, low friction and spark resistance are the key properties.
NiAl bronze
0
5
10
15
20
0.0001 0.001 0.01 0.1
Damage dose, dpa
Elon
gatio
n, %
Total elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 150°CUniform elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 150°CTotal elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 300°CUniform elongation, Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 300°C
Un-irradiated
Trend at 150 C
0
Trend at 300C
0
200
400
600
800
1000
0.0001 0.001 0.01 0.1
Damage dose, dpa
Tens
ile s
tren
gth,
MPa
Ref. 1, Fabritsiev (Cu127), Ttest = Tirr = 150°CRef. 1, Fabritsiev (Cu127), Ttest = Tirr = 300°C
Un-irradiated
Trend
0
V. Barabash, S. Fabritisiev, this Conference
Neutron effect: First data are presented at this Conference.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
22
Selection of armour materials
Compromise: Plasma Performance Materials Lifetime T retention
~ 680m2 Be first wall low Z compatibility with wide operating range and low T retention
~ 50 m2 CFC Divertor Target No melting under transients (ELMs and Disruptions)Low Z compatibility with wide range of plasma regimes (Te,div ~ 1 – 100 eV)Large T retention (co-deposition)
~ 100m2 Tungsten Baffle/Dome Low Erosion, long Lifetime and low T retention
For PSI issues see:G. Federici, “Plasma Wall Interactions in ITER”, to appear in Physica Scripta.G. Federici, et al., “Plasma material interactions in current tokama and their implications for next step fusion reactors”. Nucl. Fusion 41 (2001 )1967.
23
0
50
100
150
200
250
300
350
400
450
0 500 1000 1500 2000 2500 3000 3500 4000
Temperature, CTh
erm
al c
ondu
ctiv
ity, W
/mK
C42, C32
C30
C27
C11
C61
C62
C59
Carbon Fibre Composites
SEP NB31, CX-2002U, or equivalent
0
100
200
300
400
500
600
Ther
mal
con
duct
ivity
, W/m
K
0 500 1000 1500Temperature, °C
CX 2002U
SEP NS 31SEP NB 31
Dunlop, C 1NIC-01
Key issues: Properties variation at mass production (ITER needs ~ 10 tons):- variation of thermal conductivity (SDRT < 10 W/mK);- stable mechanical properties
Neutron effect:– it is well known neutron irradiation affect thermal conductivity, nevertheless the thermal performance is adequate (V.Barabash, ICFRM-11).
Thermal erosion at transient loads:– is being assessed (EFDA+ITER+TRINITI, RF)
Improving of manufacturing technology is ongoing.
24
Beryllium
S-65C, Brush Wellman USA, or equivalent grade(s)
Key issues: Thermal fatigue/thermal shock resistance- S-65C VHP has the better performance, but others may be also acceptable
Neutron effect:– The behavior of all Be grades with Be0 < 1% seems very similar.- However, the performance of irradiated Be in the irradiated components seems good, based on the results of in-pile tests and test of FW mock-ups after irradiation (see J. Linke, A. Gervash).
0
500
1000
1500
2000
2500
Cyc
les t
o C
rack
Initi
atio
n
0 0.5 1 1.5 2
Side Crack Propagation Depth (mm)
S-65C (L)
DShG-200
S-65C (T)
TShGT TShG-56S-65-H
S-200F (L)98% S-65
Extruded(T) TGP-56
Extruded (L)SR-200
S-200F-H94% S-65 I-400 (T)
Be/60%BeO Be/30%BeO
Grades with best fatigueperformance
S-200F (T)
B. Watson, SNLA
Metallography of S-65C and TR-30after disruption heat load
M. Roedig, FZJ
0
10
20
30
40
50
0
100
200
300
400
500
Yie
ld T
ensi
le S
treng
th, M
Pa
0 100 200 300 400 500 600 700
Temperature, °C
S-65
C/U
TS
Yield Strength, irr, L.Snead, 0.34 dpa~ 0.6 dpa Yield Strength, irr ,R.Chaouadi, ~ 1 -2.5 dpaYield strength, unirr.
Ttest = Tirr
0
5
10
15
20
25
30
35
40
45
50
0
10
20
30
40
50
Tota
l Elo
ngat
ion,
%
0 200 400 600 800Temperature, °C
M.Roedig, 0.35 dpa
L.Snead, 0.34 dpaR. Chaouadi, 1-2.5 dpaAverage, unirradiated
Ttest=Tirrad
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
25
Delamination cracks
Pure sintered W is selectedData base for pure W is sufficient.
Key issues: Thermal fatigue/thermal shock resistance- Performance of W significantly depends on orientation of grainsand production history (plate, rod, recrystallization, etc.)
0
250
500
750
1000
Duc
tile
to B
rittle
Tra
nsiti
on T
empe
ratu
re, °
C
0 25 50 75 100Fluence, 1024 n/m 2 (E>0.1 MeV), 1 dpa ~ 2 1025 n/m 2
Densimet, bending,Tirr = 250-300°C
W, tensile test, Tirr = 371-382°C
W, tensile test, Tirr = 100°C
W-10Re, bending,Tirr = 250-300°C
W, bending, Tirr = 250-300°C
0.42 1024 n/m 2
ITER Goal
Taniguchi, JAERI, 2000
Neutron effect:– Due to low irradiation temperature (~ 150-500°C, all W grades will be brittle at low dose irradiation- However, the performance of ‘brittle’ W in the irradiated components seems good (see M. Roedig, FZJ).
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
Tungsten
26
Remaining R&D issues
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
These issues have to be resolved before the starting of the procurements.
ITER is ready to be build, however, there is still a list of materials issues to be solved.
• Site adaptation: for safety important components a complete information have to be presented for the licensing needs in France;
• In some areas R&D is on-going with goal to simplify design, reduce cost, increase reliability – materials data for supporting of these designs are needed, e.g.:
• US propose cast steel for blanket modules – need neutron irr. data;• RF and China – cheaper Be grades – need to qualify them;• Different technologies used for FW – need qualify CuCrZr properties;• Etc.
• Need to have all missing data, which could impact the performance of materials in ITER
• Finally we need to provide the recommended data for Ultimate assessment and acceptance of the design.
27
Test Blanket Module Program in ITER:
ITER is considered as a most suitable test-bed for the breeding blanket of next step.
3 equatorial ports are allocated for TBM. Size of port ~1310 x 1760 mm.
Each port could include 2 types of TBM
The first TBMs must be installed in ITER since the start of the operation:
• Finalisation of the design and supporting R&D must be completed soon.
• All modules before installation in ITER must be qualified for licensing in France:
=> sufficient information about materials should be provided.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
28
Summary
The design and technical preparation for the construction of ITER are ready for implementation and inter-government negotiations are nearing to completion.
The design of the ITER components is supported by the proper and justified materials selection. This is a results of extensive and successful world-wide ITER materials and technological R&D program.
During construction phase the materials activity will be focused on:Resolving of the urgent remaining issues before starting the procurement of materials, Materials procurement and evaluation of the acceptance of the materials in the final components,Further consolidation of the data, which are needed for the licensing and for justification of the safe and reliable performance of materials during the ITER operation.
V.Barabash et al., ITER Materials, ICFRM-12, December 4 – 9, 2005, Santa Barbara, USA
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