Post on 24-Feb-2020
US/Japan Workshop onPower Plant Studies and Related Advanced Technologies
With EU ParticipationApril 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
3. Progress in R&D of Candidates1) Reduced Activation Ferritic steels2) Vanadium Alloys3) SiC/SiC Composites
4. Design Windows
5. Summary
Requirements for Fusion Materials (1)IAE, Kyoto University
● 14MeV Neutron Irradiation (DT reaction)1) Heavy displacement damage (200 dpa)2) High He and H concentration (0.5 at.%)
14Mev Neutron
Blanket
3) Low activation
Requirements for Fusion Materials (2)IAE, Kyoto University
● Blanket System Integration1) Compatibility with coolant and breeder2) Weld/Joint performance3) Complex and synergetic effects
14Mev Neutron
Plasma
Coolant
First Wall
Neutron Multiplier
T-Breeder
First WallBlanket
Requirements for Fusion Materials (3)IAE, Kyoto University
● Industrial Engineering Basis
0 2 4 6 8 10 (m)
1) Large-scale manufacturing2) Impurity control3) Cost
Reduced Activation MaterialsIAE, Kyoto University
● Safety and Environmental PerformanceReduction of radioactive waste
Some elements, such as Mo, Ag and Nb arestrictly limited to a level less than 5 wt.ppm.
● Low-activation elements: C, Cr, W, V, Ta, Ti, Mn, Si, B, Fe
Fe-9Cr-2W reduced activation ferritic steelsV-4Cr-4Ti reduced activation vanadium alloysSiC/SiC composites
Candidates for Fusion Structural Material (1)IAE, Kyoto University
500
700
900
1100
20 30 40 50
Coo
lant
Tem
pera
ture
/ °C
Thermal Efficiency (%)
F/M steel (Water or Gas Cooling Blanket)ODS steel (Gas Cooling Blanket)V –alloy (Self-cooling Liquid Blanket)SiC/SiC (Gas Cooling Blanket)
F/M steel
1. Fe-9Cr-2W reduced activation ferritic steels
V alloyODS steel
2. V-4Cr-4Ti reduced activation vanadium alloys
SiC/SiC
3. SiC/SiC composites
300
Achievements of the CandidatesIAE, Kyoto University
RadiationResistance
CostPerformance
EnvironmentalPerformance
OperationTemperature(Thermal
Efficiency)
Materials R&D Maturity(Data Base)
●F/M Steel ■V-Alloy○SiC/SiC
By A. KimuraApril, 2002Engineering
Maturity
Ferritic Steels R&D (1)IAE, Kyoto University
● Improvement of Impact Properties(1) JMTR/363K, 0.006dpa: RT(2) MOTA/663K, 22dpa: RT(3) MOTA/663K, 35dpa: RT(4) MOTA/733K, 24dpa: RT(5) MOTA/683K, 36dpa: RT(a) JMTR/493K, 0.15dpa: RT(b) EBR-II/663K, 26dpa: RT(c) MOTA/638K, 7dpa: 638K(d) HFIR/323K, 5dpa: RT(e) HFIR/673K, 40dpa: 673K
050
100
150
200
250
300
-200 -100 0 100 200 300 400
9Cr-1Mo
Shift
in D
BTT
/ K
Irradiation Hardening / MPa
(3)
(b)
(4)(5)
(a)
(c)
(d)
(e)
9Cr-2W(1)
(2)
(b)580appm He120appm He
9Cr-1Mo
0 10 20 30 40 50dpa
9Cr-2W
9Cr-2W:by Kimura9Cr-1Mo: by Klueh
1) Higher resistance to irradiation embrittlement.2) Saturation of irradiation embrittlement at 10 dpa
( above 100 dpa?)
Ferritic Steels R&D (2)IAE, Kyoto University
● Improvement of High Temperature Strength
40
6080
100
300
500
700
1 10 10 10 104
App
lied
Stre
ss /
MPa
F82H 650°C(SA)JLS-1 650°C(SA)JLS-2 650°C(SA)JLS-3 650°C(SA)
13Cr-ODS 700°C(SA)13Cr-ODS 700°C(PT)9Cr-ODS 700°C(SA)9Cr-ODS 700°C(PT)
200
JLS:by KohyamaODS:by Ukai
32
A significant increase in the creep strength has been obtained byincreasing W conc. and/or mechanical alloying method.
Irradiation study is inprogress.
Rupture Time / hr
Ferritic Steels R&D (3)IAE, Kyoto University
● Improvement of High Temperature StrengthODS steel (JNC)
0
200
400
600
800
1000
1200
1400
0 200 400 600 800
Tens
ile S
tres
s (M
Pa)
PNC-FMS
PNC316 20%CW
0
2
4
6
8
10
12
0 200 400 600 800
Uni
form
Elo
ngat
ion(
%)
PNC-FMS
ODS steel
ODS steel
Temperature(℃) Temperature (℃)
M11: 9Cr-0.12C-2W-0.2Ti-0.35Y2O3
60
80
100
300
500
Hoo
p St
ress
(MPa
)
PNC-FMS(650℃)
PNC-FMS(700℃)
PNC-FMS(750℃)
650℃
750℃
700℃
PNC316(650℃)
(700℃)
(750℃)
Target
IAE, Kyoto University
● Improvement of Creep PropertyODS steel (JNC)Basic Chemical Composition
・9Cr-Martensitic ODS9Cr-0.12C-2W-0.2Ti-0.35Y2O3Advanced radiation resistance due to reduced Cr content
・12Cr-ferritic ODS12Cr-0.03C-2W-0.3Ti-0.25Y2O3Advanced 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’
Ferritic Steels R&D (4)
10 100 1000 10000Time to Rupture (hr)
Design Windows of RAFSIAE, Kyoto University
0 2 4 6 8 10 12 14 16 18 200100200300400500600700800900
100011001200130014001500
Tem
pera
ture
(°C
)
DBTT increase by irradiation
He effect (?)
Void swelling (>1%)
Thermal creep (1% creep strain at σy/3 )
He effect (?)
Neutron wall loading MWa/m2
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
0
2
4
6
8
10
12
0 100 200 300 400 500 600 700
UN
IFO
RM
ELO
NG
AT
ION
[%]
IRRADIATION TEMPERATURE [oC]
Test Temp. ~ Irrad. Temp.
ATR 0.7-4.7dpaFFTF 7.2-54dpa, EBR-II 11dpaB. A. Loomis et al., FFTF 13-33dpaS. J. Zinkle et al., EBR-II 4dpaH. Tsai et al., ATR 4.1-4.3dpaH. Tsai et al., BOR-60 17-19dpaL. L. Snead et al., HFBR 0.5dpaH. M. Chung et al., HFIR 10dpa
V-5Ti-5Cr-1Si-1Al-1Y
V-(4-5)Cr-(4-5)Ti
(900oC)
◎ Marked ductility loss by irradiation ◎ Internal oxidationtechnique
Al, Si, Y additionAl2O3, SiO2, Y2O3
Reduction of O in thematrix 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 p
pm
Nitrogen, CN / mass ppm
V metal
IngotPlate
V metal
PlateIngotUS-DOE-HEAT
NIFS-HEAT-1
0
NIFS-HEAT-2
Ingot
Improved
V-4Cr-4Ti
V-4Cr-4Ti
V-4Cr-4Ti
by Nagasaka and Muroga◎ Oxygen: 80 wt.ppmNitrogen: 130 wt.ppm
0
2
4
6
8
10
12
14
-200 -150 -100 -50 0 50 100 150
Abs
orbe
d en
ergy
/ J
NIFS-HEAT-1
USDOE-HEAT
C+N+O=475ppm
C+N+O=290ppm
Test temperature / C
Vanadium Alloys R&D (3)IAE, Kyoto University
● Ductility Improvement by Processing
0.1
0.2
0.3
0.4
0.5
0.6
0.7
-200 -150 -100 -50 0 50Test Temperature, T/oC
Abs
orbe
d En
ergy
, E/J・
m-3
Final annealing: 1000oC, 1h
(a) 50%CW+IA(1000oC, 1h)+50%CW(b) HIP(1300oC, 147MPa, 2h)+50%CW(c) 75%CW(d) 50%CW
(a) (b)
(c)
(d)
◎ Thermo-mechanicaltreatment
lowering DBTT
by Satoh and Abe
Vanadium Alloys R&D (4)IAE, Kyoto University
● Improvement of Oxidation Resistance
500 600 700 750
V-4Cr-4TiV-4Cr-4Ti-0.5SiV-4Cr-4Ti-0.5AlV-4Cr-4Ti-0.5Y
Oxidation temperature (˚C)
Wei
ght g
ain
(mg/
mm
2 )
0.1◎ Addition of Al and Y appears to be
effective to increase oxidation resistanceat 700°C.
0.08
0.06
0.04
0.02
0 by Fujiwara and Abe
Design Windows of V AlloysIAE, Kyoto University
DBTT increase by low temperature irradiation
He embrittlement (?)
Thermal creep life at an applied stresses110MPa (2/3 of the fracture stress)
Void swelling (>2%)
0 2 4 6 8 10 12 14 16 18 20
100200300400500600700800900
100011001200130014001500
Tem
pera
ture
(°C
)
Neutron wall loading MWa/m2
And More for Vanadium Alloy IAE, Kyoto University
◎Toward DEMO ●Data base●Helium and hydrogen effects●Coating technique
◎Toward Power Reactor● All the above● Industry engineering basis
(Spin-off to non-nuclear)
SiC/SiC Composites R&D (1)IAE, Kyoto University
● Improvement of Tensile Stress(PIP:Polymer Impregnation and Pyrolysis )
by Kotani and Koyama
◎ Good performance at stoichiometric composition
SiC/SiC Composites R&D (2)IAE, Kyoto University
● Improvement of Fabrication Technique
Interfacial structure
Matrixdensity
Relatively large products
by Yang and Noda
SiC/SiC Composites R&D (3)IAE, Kyoto University
● Improvement of Irradiation Responseby Kohyama
◎ Development of new fiber (Type-S Nicalon) improved response of mechanical properties to neutron irradiation.
The advanced fiber has an almost stichiometric composition.
R&D of stoichiometric matrix SiC and appropriate interfacial structure.
SiC/SiC Composites R&D (4)IAE, Kyoto University
● Good Performance to 10 dpaby Katoh and Koyama
(?)
◎ No degradation of strength of stoichiometric Hi-Nicalon Type-S fibers.
Design Windows of SiC/SiCIAE, Kyoto University
Point defect swelling (>2%)
Decrease in thermal conductivity
Irradiation creep (?) Transmutation effect (?)
Void swelling (>2%)
Thermal creep
0 2 4 6 8 10 12 14 16 18 20
100200300400500600700800900
100011001200130014001500
Tem
pera
ture
(°C
)
Neutron wall loading MWa/m2
And More for SiC/SiC IAE, Kyoto University
◎Toward DEMO ●Data base●Helium and hydrogen effects●Fracture toughness
(Interfacial structure)
◎Toward Power Reactor● All the above● Radiation resistance●Production of large components
R&D RoadmapsIAE, Kyoto University
400 dpa200 dpa100 dpa
Ceramicscomposits
Cand
idates
for S
tructu
ral M
ateria
ls
Austeniticsteels
2050
Reducedactivationferritic steelsODS steels
Vanadiumalloys
2030 2035 2040 204520252005 2010 2015 2020
Comertialreactor
FUSIO
N Rea
ctors
Materials 2000
DEMO
Experimentalreactor (ITER)
CDA &EDA
*fracture toughness, *matrix coating, *matrix/fiber *processing R&D, *leak rate, *weld/joint R&D
DEMO H. Load: 10MWa/m2
Temp.: 550°C
*Irradiation effects, *matrix*fiber, *complex, *boundary issues
*need purificationO,N,C <10wt.ppm
*coating tech.
*need processing R&Dand weld/joint tech.
*ready for ITER test blanket module
Construction and Operation
Reconstruction and Operation
Construction and Operation
1st Plasma
CDA &EDA
System startupTesting
Generation of power
ITER H. Load: 0.3MWa/m2
Temp.: <150°CCoolant: H2O
Reduced Activation ODS steels
V-4Cr-4Ti Alloys
SiC/SiC Composites
*ready for ITERSUS316L(N)-ITER Grade (0.06-0.08%N)
SUS30467(2%B) for shielding
F82H, JLF-1 (Martensitic steels)
92-ODS steels
HP-V-4Cr-4Ti Alloys
LS-SiC/SiC
JLF92H/92ODS Composites
JLF92H
1st Material Selection
POWER REACTOR H. Load: 20-40MWa/m2
Temp.: 600-800°C
DBTT<383Kσy > 220MPa 105hr at σ=100MPa
DBTT<383K105hr at σ=110MPa
DBTT<383K105hr at σ=110MPa
2nd Material Selection
ODS-V-4Cr-4Ti-Y,Al
LS-3D-SiC/SiC
Requirements *Heat loading: 20-40MWa/m2
*Transmutation He: 3000-6000atppm*Heat efficiency: >50%
ITER 1st Plasma Shift to DEMO DEMO startup Power reactor
3rd Material Selection
IFMIF CDA &EDA Construction Testing
*Engineering bases
*Largecomponents
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.
ConclusionIAE, Kyoto University
RadiationResistance
CostPerformance
Environmental Safety
OperationTemperature(Thermal
Efficiency)
The solutions will be discovered by you!!
Materials R&D Maturity(Data Base)
●F/M Steel ■V-Alloy○SiC/SiC
By A. KimuraJuly, 2001Engineering
Maturity