ITC-12/APFA'01
December 10-14, 2001
Toki, Gifu, Japan

Fusion Technology R & D in Japan- Goes into the 21st Century under the Unified Structure -Akira Kohyama
Institute of Advanced Energy,
Kyoto University

Fusion Program in Japan
by N. Inoue (Chair of Fusion Council, Japan)
Institute of Advanced Energy,
Kyoto University

Japan intends to develop fusion as a viableenergy option for the future
Construction of experimental reactor has thehighest priority
Serious discussions have been and are being
made to make a confident decision on ITER
construction
In parallel, Japan studies various concept
improvements in plasma confinement, as well as
materials development and reactor technology

Fusion Development Strategy in Japan
Third Phase Basic Program

Second Phase

Fourth Pha se Program

Improved Confinement

AdvancedTokamak
Study

Advanced Divertor
Steady Sta teIgnition
Lon g-Burn(steady state)Core
ProgramExperime ntal
Rea ctor（ITER）(ITER)

JT-60

Selectionof Concept

Electricity
Gen eration
DEMO
Power
PlantsEquivalent Q > 1~
Safety Impro vemen t
Re actor RelevantTechnology R&D
ReactorTechnologies

SC, H eating,Tritium, Blanket
Remote Maintenance, High Heat Fluxetc.
Demo Blanket
IFMIF

High Field, High Temperature SC Coil

Low Activat ion Materials (Ferritic Steel etc.)

Advanced Materials (SiC etc.)

Alternative and /o r Innovative Concept Researches
Univers ities andNationa l Lab.

Helical System, RFP, Mirror, ICF, etc .

By N. Inoue (FESAC at Snowmass, 1999)

Material/Blanket R & D Strategy
Institute of Advanced Energy,
Kyoto University

Fusion Technology R & D in Japan- Goes into the 21st Century, under the Unified Structure -Institute of Advanced Energy,
Kyoto University

From April, 1, 2001:

MOE (Monbusho)

STA

Universities/NIFSJAERI/Nat. Labs.

MEXT
Ministry of Education,Culture, Sports, Scienceand Technology

Fusion Technology R & D in Japan- In the past, under the dual structure -Institute of Advanced Energy,
Kyoto University

MOE (Monbusho)

STA

Universities/NIFSJAERI/Nat. Labs.

Long-Term R & D Issues
Basic/Fundamental Research

Near-Term R & D Issues
Energy Emphasized Efforts

Fusion Engineering Network
MOE/US-DOE Collaboration

ITER-EDAJAERI/ORNL Collaboration

There have been Many efforts forInteraction/Collaboration

Fusion Engineering Network Activity
Institute of Advanced Energy,
Kyoto University

Structural Materials
Materials/FuelIn-reactor Component(PWI)
Blanket Technology

Magnetoelectric/MagnetTritium Science/Engineering
Tritium Bio-chemical effects

System/Safety

Thermo-mechanics
Reactor Design
System/Safety Design

ICF

Neutronics

A Good Example can be seen in Materials R & D- In the past, under the dual structure -Institute of Advanced Energy,Kyoto University

MOE (Monbusho)

STA

DOE/Monbusho
Collaboration

JAERI/ORNL
Collaboration

RTNS-II (14MeV effect)

Austenitic Stainless Steel (Fundamental)

FFTF/MOTA (High dpa effects)

Austenitic S. S. (Weldment/Component)

JUPITER (Dynamic/Varying/
Cumulative effects)

Reduced Activation Ferritic Steels( Fundamental)

JUPITER-2 (Integration forAdvanced Blanket)

Reduced Activation Ferritic Steels(weldment/He effects,,)

Another Good Example can be seen in ITER EDA- ITER/Japan Team with University participationsInstitute- of Advanced Energy,
Kyoto University

There are many extinguished accomplishments
In Fusion Engineering R & D( well known 7 accomplishments)
Many supporting activities in Japanese Universities

Examples of ITER EDA Accomplishments- Fusion Engineering -Institute of Advanced Energy,
Kyoto UniversityBlanket Technology
Be armour(1) Development of ITER Shielding Blanket
Beryllium-armored full-width First Wall panel(DSCu/SS) has successfully fabricated by HIP
technique first time.

(2) Development of Breeding Blanket• Effects of thermal cycles on thepebble bed structure has been
investigated.

He gas• World’s first mass production
technology of Beryllium neutron
multiplier pebbles has successfully
been developed by the Rotating
Electrode Method.
Electric arc

Development of ITER Divertor
Development of New Cooling Structure- Save space and cost -Be rotating
electrode

Be Pebbles

L5 Divertor Project
Annular Flow
Cassette body; US

CFC Monoblock

Vertical target; JA~1 m

SS Sliding
SupportStructure
Integration Tests of JA and UScomponents were successfully completed.

The vertical target mock-up with annular
flow has successfully withstood a heat loadof 20 MW/m2, 10s for 1000 cycles.

New Project JUPITER-II (2001 - 2006)- Materials integration utilizing reactor irradiation
and related basic research for advanced blankets -Institute of Advanced Energy,
Kyoto University

Targets of Fusion PowerReactors- Attractive Options Coolant Temperature (C)1100

Institute of Advanced Energy,
Kyoto University

SiC/SiC/He

900
700

V-4Cr-4Ti(Si,Al,Y) /Li

500

RAF-ODS(JLF-OD, F82H-OD) /H2O
RAFs (JLF-1, F82H) /H2O

300316 SS /H2O
203040

Energy Conversion Efficiency to Electricity (%)

50

Blanket R&D in Japan

Institute of Advanced Energy,Kyoto University“Medium-term research plan for power generating breeding blanket”
August 2000, by Fusion Council- R&D for DEMO blanket. Blanket module test in ITER: important milestone.
- Three C&Rs and selections scheduled- JAERI: core institute for solid blanket development
Universities (NIFS): fundamental studies to obtain perspective on liquid blanket,
material development, various fundamental studies on solid and liquid blankets

Reference blanket JAERI: lithium ceramics cooled by supercritical water
NIFS: FFHR, flibe as breeder and coolant

Advanced blanket concepts with high coolant temperature, advanced safety,
high resistance for large neutron flounce
Flibe Blanket, Liquid Lithium with Vanadium Alloy, Solid Breeder and SiC/SiC

JUPITER-II: Japan-MEXT US-DOE collaborative project on advanced blankets
2001-2006, mainly using facilities at INEEL, UCLA, ORNL, ANL

Reduced Activation Ferritic Steel R&D in Japan
Institute of Advanced Energy,Kyoto University

RAF Database (F82H/JLF-1)
Since ’92 for a decade, Under the Japanese initiative,
RAF database has been constructed ( IEA RAF WG)
Creep Property(F82H)

Display Modes (Example)

Data Plot (Example)

Reduced Activation Ferritic Steel R&D in Japan
Institute of Advanced Energy,
Kyoto University

Improvement in DBTT
F82H

W steels
W1 660K 22.9dpa
W2 660K 22.9dpa
W3 660K 22.9dpa
W4 660K 22.9dpa

Unirradiated17.7dpa 663K
44.0dpa 700K

300
F82H

DBTT (K)250
DBTT trend band for JLF-1
JLF-1 680K

200

JLF-1

Unirradiated0.01dpa 570K JMTR
13.1dpa 646K22.9dpa 681K
34.5dpa 680K
43.8dpa 700K
59.3dpa 680K

W steels

150

100
020406080

Displacement Damage (dpa)100Radiation Effect on Stress Amplitude

Institute of Advanced Energy,Kyoto University

JMTR ~0.005dpa (3.1x1019cm-2 / Irr. Temp. ~90C)

Peak stress (MPa)

800

F82H
F82H IEA
IEA Heat
Heat

700

UnirradiatedUnirradiated□∆ε=2%, Nf:9.8x10 22□ ∆ε∆εaa=2%a=2%, Nf:9.8x10◆∆ε=1%, Nf:4.6x10 33◆ ∆ε∆εaa=1%a=1%, Nf:4.6x10600
500400-400

Post
Post irradiation
irradiation□∆ε=2%, Nf:1.3x10 22□ ∆ε∆εaa=2%a=2%, Nf:1.3x10◆∆ε=1%,◆ ∆ε∆εaaa=1%=1%, NNff:2.3x10:2.3x1033-500-600-700-800

110102103
Number of cycles

10 4• The increase of initial stress amplitude was 292MPa at ∆εa= 2%,and 116MPa at ∆εa= 1%.
• Number of cycles to failure of ∆εa= 2% case was reduced to 13%of unirradiated case.

R & D of Ferritic Steels for Fusion- from Fundamental Materials R & D to Technology/Engineering Integration Institute of Advanced Energy,Kyoto University

Data base development toward DEMO.
Ferromagnetic effects.
Development of high heat-resistant supersteels and ODS steels.
Development of welding/joining
technology and ODS-clad processing.
Compatibility with pressurized water andsuper critical water.

Performance Evaluationand Improvement
underNeutron Environment
Blanket Environment

Technology/Engineering
Integration forBlanket/Reactor
Components Fabrication
SSTR or ASSTR

Fabrication of High Purity Large Products of V-4Cr-4Ti(NIFS-HEATs)
Large V-4Cr-4Ti ingots with reduced impurity levels were produced in NIFSFeasibility of recycling by quasi-remote (simply shielded) processing was verified
The resulting products were used for Round-robin test by international collaboration
1200

400

1000200
Ingot

Ingot
NIFS-HEAT-1
Plate

800
Recycling Limit
Single remote recycling
Single quasi-remote recycling
Single hands-on recycling
Multiple remote recycling

600
400

Improved
V metal

200

100
200
300
Nitrogen, C N / mass ppm

0NIFS-HEAT-2
0

Impurity Level (wppm)

V metal100

US832665
US832864
NIFS-HEAT-1
NIFS-HEAT-2

US-DOE-HEAT300

O

Oxygen, C / mass ppm

Plate
Ingot

than*# lessnot available# *400

Impurity level of NIFS- and US- HEATs

AlNb(x10)*Ag(x100)*Mo

Impurity level and recycling criteria(Muroga, Nagasaka, Heo, NIFS)

Improvement of Welding Property
Reduction of oxygen level in NIFS-HEAT resulted in significant
enhancement of the mechanical property of the weld joint
14
Weld joint

Absorbed energy / J

12NIFS-HEAT-1

108642USDOE-HEAT

0-200 -150 -100 -50

050100 150

Test temperature / C

TIG weld joint of NIFS-HEAT-1

Impact property of the TIG weld joint(Muroga, Nagasaka, Heo, NIFS)

Improvement of Resistance to Radiation andOxidation by Addition of Si, Al and Y
Ductility after irradiation at 300~400C was significantly enhanced
Oxidation during exposure to air was strongly suppressed to 973K

Weight gain (mg/mm2)

0.1
0.08

V-4Cr-4Ti
V-4Cr-4Ti-0.5Si
V-4Cr-4Ti-0.5Al
V-4Cr-4Ti-0.5Y

Melted

0.06
0.04

0.02
0Oxidation inair for 1 hr

773

873

973

1023

Oxidation temperature (K)Uniform elongation of irradiated alloys

Weight gain by exposure to air(Abe, Satou, Fujiwara, Chuto, Tohoku University)

Irradiation Effects
on Mechanical Properties of SiC/SiC

Institute of Advanced Energy,Kyoto University

SuIrrad./SuUnirrad.

◆ Hi-Nicalon Type-S/PyC/FCVI-SiC■ Hi-Nicalon/PyC/FCVI-SiC▲ Nicalon/PyC/FCVI-SiC◇ Tyranno-SA/PyC/FCVI-SiC○ Monolithic CVD-SiC

12888 81

1: 500C, HFIR
2: 400C, HFIR
3: 200-500C, HFIR
4: 300-500C, JMTR
5: 430-500C, EBR -II

26 7 7667 7422 122110975

6: 300C, HFIR
7: 800C, HFIR
8: 800C, JMTR9: 740C, HFIR
10: 630, 1020C, ETR

Improvements in
Fiber and processing

757533500.1

110
Neutron Dose [dpa-SiC]

100

Improvement in Thermal Stress Figure of Merit- by LPS-SiC/SiC -Institute of Advanced Energy,Kyoto University

50
Steel (Fe-8Cr-2W)
V-alloy (V-4Cr-4Ti)
SiC/SiC (SNECMA Cerasep N3-1)
SiC/SiC (MER CVR)
SiC/SiC (CREST-ACE LPS Lab-grade)4540

M [kW/m]3530

M=252015

σ UTS K th (1 − ν )α th E

LPS-SiC/SiC
Massproduction
grade

10502007001200Temperature [K]1700

SiC/SiC R & D Goals and Status- 2000 -Institute of Advanced Energy,
Kyoto University

TAURO

IrradiatedTensileStrength(500MPa/FS)

IrradiatedUpper
Temp. Limit

Mass Density

ARIES-AT(3.0Mg/m3/FS)

D-SSTR

UnirradiatedTensileStrength?

Irradiated
ThermalConductivity(20W/m-K/FS)

CREST-CVI(500MPa/FS)?

(1500℃/FS)

CERASEP

UnirradiatedUpper
Temp. Limit Additional Issues?

(1500℃/FS)

Young’smodulus(300GPa/FS)

Unirradiated
ThermalConductivity(20W/m-K/FS)

Fatigue lifetime
Creep strength
JoiningHermetic sealing
Chemical compatibility
Tritium retention
Fabrication cost
Impurity control

PWI and Plasma Facing Materials
Institute of Advanced Energy,
Kyoto University

Major Subjects(1) High Z plasma facing materials and interaction with plasma
Nagoa Univ., Doshisha Univ., Fukuoka Univ. of Education, Kyoto Univ.,
NIFS, etc. (IEA-TEXTOR Collaboration)(2) Measurement of tritium in the plasma facing materials of fusion
experimental devices
Toyama Univ., Nagoya Univ., NIFS (IEA-TEXTOR Collaboration)(3) Developments and evaluation of high-Z plasma facing materials
Tohoku Univ., Kyushu Univ., NIFS, Kagoshima Univ., etc.
(LHD Joint Projects)(4) H and He irradiation experiments of W-coated materials with plasmasimulators
Kyushu Uinv., NIFS (J-US Collaboration)(5) Analysis of the first wall of TRIAM-1M and LHD
Hokkaido Univ., NIFS (LHD Joint Projects, NIFS Joint Projects)

Deposit of TRIAM-1M TokamakFusion Group RIAM Kyushu University

TRIAM-1M deposit

TDS of D2
D injection：RT, 3x1021 D/m2

D2 desorption[D 2/m 2s]

2 101.5 101 105 1018Mo+O on SUS
218Mo BulkSUS1817Fine fcc crystals, 1nm
0300 400 500 600 700 800 900 1000 1100
Temperature[K]Due to the co-deposition of Mo with residual Oxygen, structure
and properties of the wall surface change completely.
High D retention change hydrogen recycling.
Critical issue for the control of steady state plasma

FUEL CYCLE!, SAFETY!

ITER
Loss

Test Blanket(Heat, Tritium)

FusionPlasmaLoss
Emergency
RecoveryLoss

Tritium

Main Fuel Cycle
Deuterium

Ash

Loss

Tritium Flow in Fusion Reactor

(I)( II )

Preparatory Plasma
Experiment( III )

D-T Burning Plasma
ExperimentControlled Plasma
Operationsafety confinement
preparatory safety
confinement (remote control)
tritium
study
biologicalenvironmental
assessment
assessment(preparation, settlement, measurement, control,regulation, waste treatment)

fueling systemeachunit operation

fueling system
withone throughmode

fueling system
remote control
and continuous
operation

self-sustained
fuel cycle
improvedfueling system

Where are we struggling ?
-Radiation Effects-

0.04 psec

0.1 psec

0.2 psec

0.26 psec

Institute of Advanced Energy,
Kyoto University

50 keV
PKA(ave.
fusion)10 keV PKA(ave. fission)

5 nm

Peak damage state iniron cascades at 100K
(S.Zinkle et al.)Conclusion

Fusion Engineering Activities in Japan are quite active
and efficient under the newly unified structure, MEXT.
Near term issues, for ITER, and long term issues, forDEMO and Power reactor are simultaneously carriedout, well balanced and well managed condition.
Fusion Engineering Activities in Japan will bestrengthened and accelerated with the decision of theITER invitation to Japan, in the near future.