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

● 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 Performance Reduction 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

300

500

700

900

1100

20 30 40 50

Co

ola

nt

Tem

per

atu

re /

°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

Achievements of the CandidatesIAE, Kyoto University

RadiationResistance

CostPerformance

Engineering Maturity

Materials R&D Maturity(Data Base)

EnvironmentalPerformance

OperationTemperature(Thermal Efficiency)

●F/M Steel ■V-Alloy○SiC/SiC

By A. KimuraApril, 2002

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

0

50

100

150

200

250

300

-200 -100 0 100 200 300 400

9Cr-1Mo

Sh

ift

in D

BT

T /

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 50

dpa

9Cr-2W

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

NIFS-HEAT-2

Ingot

Improved

V-4Cr-4Ti

V-4Cr-4Ti

V-4Cr-4Ti

by Nagasaka and Muroga ◎ Oxygen: 80 wt.ppm

Nitrogen: 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

Test temperature / C

NIFS-HEAT-1

USDOE-HEAT

C+N+O=475ppm

C+N+O=290ppm

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 50

Test Temperature, T/oC

Ab

sorb

ed E

ner

gy,

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-mechanical treatment

lowering DBTT

by Satoh and Abe

Vanadium Alloys R&D (4)IAE, Kyoto University

● Improvement of Oxidation Resistance

0

0.02

0.04

0.06

0.08

0.1

500 600 700 750

V-4Cr-4TiV-4Cr-4Ti-0.5SiV-4Cr-4Ti-0.5AlV-4Cr-4Ti-0.5Y

Oxidation temperature (˚C)

We

igh

t g

ain

(m

g/m

m2)

◎ Addition of Al and Y　　 appears to be effective to increase oxidation resistance at 700°C.

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 stresses

110MPa (2/3 of the fracture stress)

Void swelling (>2%)

Neutron wall loading MWa/m2

0 2 4 6 8 10 12 14 16 18 20

100

200300400500600

700800900

100011001200

13001400

1500

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 )

◎ Good performance at stoichiometric composition

by Kotani and Koyama

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 Response

◎ 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.

by Kohyama

SiC/SiC Composites R&D (4)IAE, Kyoto University

● Good Performance to 10 dpa

◎ No degradation of strength of stoichiometric Hi-Nicalon Type-S fibers.

by Katoh and Koyama

(?)

Design Windows of SiC/SiCIAE, Kyoto University

Point defect swelling (>2%)

Decrease in thermal conductivity

Irradiation creep (?) Transmutation effect (?)

Void swelling (>2%)

Thermal creep

Neutron wall loading MWa/m2

0 2 4 6 8 10 12 14 16 18 20

100

200300400500600

700800900

100011001200

13001400

1500

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

Ceram ics

com posits

Ca

nd

ida

tes

fo

r S

tru

ctu

ral M

ate

ria

ls

Austenitic

steels

2050

Reduced

activation

ferritic steels

ODS steels

Vanadium

alloys

2030 2035 2040 204520252005 2010 2015 2020

Com ertial

reactor

FU

SIO

N R

ea

cto

rs

Materials 2000

DEMO

Experim ental

reactor (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 purification O,N,C <10wt.ppm*coating tech.

*need processing R&D and 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<383Ky > 220MPa 105hr at =100MPa

DBTT<383K105hr at =110MPa

DBTT<383K105hr at =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

*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.

ConclusionIAE, Kyoto University

RadiationResistance

CostPerformance

Engineering Maturity

Materials R&D Maturity(Data Base)

Environmental Safety

OperationTemperature(Thermal Efficiency)

●F/M Steel ■V-Alloy○SiC/SiC

By A. KimuraJuly, 2001

The solutions will be discovered by you!!