US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation April 6-7, 2002, Hotel Hyatt Islandia, San Diego, CA Design Windows of Candidate Fusion Structural Materials Akihiko Kimura Institute of Advanced Energy Kyoto University
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US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation
US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation April 6-7, 2002, Hotel Hyatt Islandia, San Diego, CA. Design Windows of Candidate Fusion Structural Materials. Akihiko Kimura Institute of Advanced Energy Kyoto University. Contents. - PowerPoint PPT Presentation
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US/Japan Workshop onPower Plant Studies and Related Advanced Technologies
With EU Participation April 6-7, 2002, Hotel Hyatt Islandia, San Diego, CA
Design Windows of Candidate Fusion Structural Materials
Akihiko KimuraInstitute of Advanced Energy
Kyoto University
ContentsIAE, Kyoto University
1. Requirements for Fusion Structural Materials
2. Candidates for FSM
4. Design Windows
3. Progress in R&D of Candidates1) Reduced Activation Ferritic steels2) Vanadium Alloys3) SiC/SiC Composites
5. Summary
Requirements for Fusion Materials (1)IAE, Kyoto University
● 14MeV Neutron Irradiation (DT reaction)
14Mev Neutron
Blanket
1) Heavy displacement damage (200 dpa)
2) High He and H concentration (0.5 at.%)3) Low activation
Requirements for Fusion Materials (2)IAE, Kyoto University
● Blanket System Integration
14Mev Neutron
First Wall
Neutron Multiplier
T-Breeder
Plasma
Coolant
BlanketFirst Wall
1) Compatibility with coolant and breeder2) Weld/Joint performance3) Complex and synergetic effects
Requirements for Fusion Materials (3)IAE, Kyoto University
1) Higher resistance to irradiation embrittlement.2) Saturation of irradiation embrittlement at 10 dpa ( above 100 dpa?)
9Cr-2W:by Kimura9Cr-1Mo: by Klueh
Ferritic Steels R&D (2)IAE, Kyoto University
● Improvement of High Temperature Strength
A significant increase in the creep strength has been obtained byincreasing W conc. and/or mechanical alloying method.
Irradiation study is inprogress.
40
60
80100
300
500
700
1 10 10 10 104
Ap
plie
d S
tres
s / M
Pa
Rupture Time / hr
F82H 650C(SA)JLS-1 650C(SA)
JLS-2 650C(SA)
JLS-3 650C(SA)
13Cr-ODS 700C(SA)
13Cr-ODS 700C(PT)
9Cr-ODS 700C(SA)
9Cr-ODS 700C(PT)
200
JLS:by KohyamaODS:by Ukai
32
Ferritic Steels R&D (3)IAE, Kyoto University
0
200
400
600
800
1000
1200
1400
0 200 400 600 800
Ten
sile
Str
ess
(MP
a)
Temperature (℃)
PNC-FMS
PNC316 20%CW
0
2
4
6
8
10
12
0 200 400 600 800
Un
ifo
rm E
lon
gat
ion
(%)
Temperature (℃ )
PNC-FMS
● Improvement of High Temperature Strength ODS steel (JNC)
ODS steel
ODS steel
M11: 9Cr-0.12C-2W-0.2Ti-0.35Y2O3
60
80
100
300
500
10 100 1000 10000
Ho
op
Str
ess
(M
Pa)
Time to Rupture (hr)
PNC-FMS(650 )℃
PNC-FMS(700 )℃
PNC-FMS(750 )℃
650℃
750℃
700℃
PNC316(650 )℃
(700 )℃
(750 )℃
Target
Ferritic Steels R&D (4)IAE, Kyoto University
● Improvement of Creep Property ODS steel (JNC)Basic Chemical Composition
・ 9Cr-Martensitic ODS 9Cr-0.12C-2W-0.2Ti-0.35Y2O3
Advanced radiation resistance due to reduced Cr content・ 12Cr-ferritic ODS 12Cr-0.03C-2W-0.3Ti-0.25Y2O3
Advanced corrosion resistance ・ Claddings manufactured by cold-rolling
Mechanical Properties・ Advantageous at higher temperature over 700 in creep rupture and ℃ tensile strength, comparing with PNC316・ Maintain good ductility
Irradiation・ Up to 15 dpa at 400 to 530℃ ℃・ Maintain strength and ductility without α’
Design Windows of RAFSIAE, Kyoto University
Neutron wall loading MWa/m2
0 2 4 6 8 10 12 14 16 18 200100
200300400500600
700800900
100011001200
13001400
1500
Tem
pera
ture
(°C
)
DBTT increase by irradiation
He effect (?)
Void swelling (>1%)
Thermal creep (1% creep strain at y/3 )
He effect (?)
And More for Ferritic Steels IAE, Kyoto University
◎Toward DEMO ●Data base●Helium and hydrogen effects●Compatibility issues
◎Toward Power Reactor● All the above● RAFS/ODS combination design
Vanadium Alloys R&D (1)IAE, Kyoto University
● Ductility Improvement by Alloying
◎ Marked ductility loss by irradiation ◎ Internal oxidation technique
Al, Si, Y addition Al2O3, SiO2, Y2O3
Reduction of O in the matrix of vanadium
Heat treatment condition is another concern.
by Satoh and Abe
Vanadium Alloys R&D (2)IAE, Kyoto University
● Ductility Improvement by Purification
100
200
300
400
100 200 300 400
Oxy
gen,
CO
/ m
ass
ppm
Nitrogen, CN
/ mass ppm
V metal
IngotPlate
V metal
PlateIngot
US-DOE-HEAT
NIFS-HEAT-1
0
* W. R. Johnson and J. P. Smith, J. Nucl. Mater., 258-263 (1998) 1425-1430
ITER 1st Plasma Shift to DEMO DEMO startup Power reactor
3rd Material Selection
IFMIF CDA &EDA Construction Testing
*Engineering bases
*Large components
Testing
SummaryIAE, Kyoto University
1. Ferritic steel is the first candidate of fusion blanket structural materials. The issue of high thermal efficiency can be solved by application of ODS steel. Combined utilization of ferritic steel and ODS steel is quite promising for power reactor.
2. Vanadium alloys and SiC/SiC composites are also desirable for fusion power reactors. Extensive improvements have been achieved so far. R&D of material processing for production of large-scale component is strongly demanded.