Design Windows of Candidate Fusion Structural MaterialsDesign Windows of Candidate Fusion Structural...

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


● 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


● 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?)


● 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


● 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


● 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


● 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


● 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

)


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


● Improvement of Fabrication Technique

Interfacial structure

Matrixdensity

Relatively large products

by Yang and Noda


● 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.


● 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

)


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

Design Windows of Candidate Fusion Structural MaterialsDesign Windows of Candidate Fusion Structural...

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