Fusion Technology R & D in JapanITC-12/APFA'01 December 10-14, 2001 Toki, Gifu, Japan Fusion...

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

Fusion Technology R & D in JapanFusion Technology R & D in Japan

Akira KohyamaAkira KohyamaInstitute of Advanced Energy,Institute of Advanced Energy,

Kyoto UniversityKyoto University

-- Goes into the 21st Century under the Unified Structure Goes into the 21st Century under the Unified Structure --

Fusion Program in JapanFusion Program in Japan

Japan intends to develop fusion as a viable energy option for the futureConstruction of experimental reactor has the highest priority Serious discussions have been and are being made to make a confident decision on ITER constructionIn parallel, Japan studies various concept improvements in plasma confinement, as well asmaterials development and reactor technology

by N. Inoue (Chair of Fusion Council, Japan)by N. Inoue (Chair of Fusion Council, Japan)Institute of Advanced Energy

, Kyoto University

Fusion Development Strategy in Japan

Second Phase Third Phase Basic Program Fourth Phase Program

JT-60Experimental

Reactor （ITER）

DEMO

Steady State

Improved Confinement

Advanced Divertor

Ignition Long-Burn (steady state)

Electricity Generation

Equivalent Q > 1~

Advanced Tokamak Study

Reactor Technologies

Core Program

Universities and National Lab. Helical System, RFP, Mirror, ICF, etc.

Power Plants

Safety Improvement

Demo Blanket

Reactor Relevant Technology R&D

SC, Heating,Tritium, Blanket Remote Maintenance, High Heat Flux etc.

Alternative and/or Innovative Concept Researches

High Field, High Temperature SC Coil

Low Activat ion Materials (Ferritic Steel etc.) Advanced Materials (SiC etc.)

Selection of Concept

By N. Inoue (FESAC at Snowmass, 1999)

IFMIF

(ITER)

Material/Blanket R & D StrategyMaterial/Blanket R & D StrategyInstitute of Advanced Energy

, Kyoto University

Fusion Technology R & D in JapanFusion Technology R & D in Japan-- Goes into the 21Goes into the 21stst Century, under the Unified Structure Century, under the Unified Structure --

From April, 1, 2001:From April, 1, 2001:

MOE (MOE (MonbushoMonbusho))

Universities/NIFSUniversities/NIFS

STASTA

JAERI/Nat. Labs.JAERI/Nat. Labs.

MEXTMEXT

Ministry of Education, Ministry of Education, Culture, Sports, Science Culture, Sports, Science

and Technologyand Technology

Institute of Advanced Energy,

Kyoto University

Fusion Technology R & D in JapanFusion Technology R & D in Japan-- In the past, under the dual structure In the past, under the dual structure --


Universities/NIFSUniversities/NIFS

STASTA

JAERI/Nat. Labs.JAERI/Nat. Labs.

NearNear--Term R & D IssuesTerm R & D IssuesEnergy Emphasized EffortsEnergy Emphasized Efforts

LongLong--Term R & D IssuesTerm R & D IssuesBasic/Fundamental ResearchBasic/Fundamental Research

There have beenThere have been ManyMany efforts forefforts forInteraction/CollaborationInteraction/Collaboration


Kyoto University

Fusion Engineering NetworkMOE/US-DOE Collaboration

ITER-EDAJAERI/ORNL Collaboration

Fusion Engineering Network ActivityFusion Engineering Network ActivityInstitute of Advanced Energy

, Kyoto University

Materials/FuelMaterials/Fuel

MagnetoMagneto--electric/Magnetelectric/Magnet

System/SafetySystem/Safety

ICFICF

Structural MaterialsStructural Materials

InIn--reactor Componentreactor Component(PWI)(PWI)

Blanket TechnologyBlanket Technology

Tritium Science/EngineeringTritium Science/Engineering

Tritium BioTritium Bio--chemical effectschemical effects

ThermoThermo--mechanicsmechanics

Reactor DesignReactor Design

System/Safety DesignSystem/Safety Design

NeutronicsNeutronics

A Good Example can be seen in Materials R & D A Good Example can be seen in Materials R & D -- In the past, under the dual structure In the past, under the dual structure --


DOE/DOE/MonbushoMonbushoCollaborationCollaboration

STASTA

JAERI/ORNLJAERI/ORNLCollaborationCollaboration

RTNS-II (14MeV effect) Austenitic Stainless Steel (Fundamental)

FFTF/MOTA (High dpa effects) Austenitic S. S. (Weldment/Component)

JUPITER (Dynamic/Varying/ Reduced Activation Ferritic SteelsCumulative effects) ( Fundamental)

JUPITER-2 (Integration for Reduced Activation Ferritic SteelsAdvanced Blanket) (weldment/He effects,,)


Kyoto University

AnotherAnother Good Example can be seen in ITER EDA Good Example can be seen in ITER EDA -- ITER/Japan Team with University participations ITER/Japan Team with University participations --Institute of Advanced Energy

, Kyoto University

There are many extinguished accomplishmentsThere are many extinguished accomplishmentsIn Fusion Engineering R & D In Fusion Engineering R & D ( well known 7 accomplishments)( well known 7 accomplishments)

Many supporting activities in Japanese Universities Many supporting activities in Japanese Universities

Examples of ITER EDA AccomplishmentsExamples of ITER EDA Accomplishments-- Fusion Engineering Fusion Engineering --


Kyoto University

(1) Development of ITER Shielding Blanket

Beryllium-armored full-width First Wall panel (DSCu/SS) has successfully fabricated by HIP technique first time.

Blanket Technology

(2) Development of Breeding Blanket• Effects of thermal cycles on the

pebble bed structure has been investigated.

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

Be armour

He gas

Be rotatingelectrodeElectric arc

Be Pebbles

(1) Development of ITER Shielding Blanket

Beryllium-armored full-width First Wall panel (DSCu/SS) has successfully fabricated by HIP technique first time.

Blanket Technology

(2) Development of Breeding Blanket• Effects of thermal cycles on the

pebble bed structure has been investigated.

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

Be armour

He gas

Be rotatingelectrodeElectric arc

Be Pebbles

Development of New Cooling Structure- Save space and cost -

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

CFC MonoblockL5 Divertor Project

Cassette body; US Vertical target; JA

Integration Tests of JA and US components were successfully completed.

Annular Flow

SS Sliding Support Structure

~1 m

Development of ITER Divertor

New Project JUPITER-II (2001 - 2006)- Materials integration utilizing reactor irradiation and related basic research for advanced blankets -


Kyoto University

Co

ola

nt

Tem

per

atu

re (

C)

Energy Conversion Efficiency to Electricity (%)

RAFs (JLF-1, F82H) /H2O

20 30 40 50

300

500

700

900

1100

316 SS /H2O

Targets of Fusion Power Targets of Fusion Power ReactorsReactors

-- Attractive Options Attractive Options --

SiC/SiC/He

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

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


Kyoto University

Blanket R&D in JapanBlanket R&D in Japan

“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 developmentUniversities (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 waterNIFS: FFHR, flibe as breeder and coolant

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

JUPITER-II: Japan-MEXT US-DOE collaborative project on advanced blankets2001-2006, mainly using facilities at INEEL, UCLA, ORNL, ANL


Kyoto University

Since ’92 for a decade, Under the Japanese initiative, RAF database has been constructed ( IEA RAF WG)

Display Modes (Example)Display Modes (Example) Data Plot (Example)Data Plot (Example)

Creep Property(F82H)

RAF Database (F82H/JLF-1)


Kyoto University

Reduced Activation Ferritic Steel R&D in JapanReduced Activation Ferritic Steel R&D in Japan

Improvement in DBTT


Kyoto University

Reduced Activation Ferritic Steel R&D in JapanReduced Activation Ferritic Steel R&D in Japan

DBTT trend band for JLF-1

100

150

200

250

300

0 20 40 60 80 100

Displacement Damage (dpa)

DB

TT

(K)

Unirradiated0.01dpa 570K JMTR13.1dpa 646K22.9dpa 681K34.5dpa 680K43.8dpa 700K59.3dpa 680K

Unirradiated17.7dpa 663K44.0dpa 700K

F82H

Unirradiated17.7dpa 663K44.0dpa 700K

F82H

JLF-1

F82H

JLF-1 680K

W3 660K 22.9dpaW4 660K 22.9dpa

W1 660K 22.9dpaW2 660K 22.9dpa

W steels

W3 660K 22.9dpaW3 660K 22.9dpaW4 660K 22.9dpaW4 660K 22.9dpa

W1 660K 22.9dpaW1 660K 22.9dpaW2 660K 22.9dpaW2 660K 22.9dpa

W steels

W steels

Radiation Effect on Stress AmplitudeRadiation Effect on Stress Amplitude

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

• 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.


Kyoto University

1 10

400-400

500600

700

800

-500

-600

-700

-800102 103 104

Number of cycles

Pea

k st

ress

(M

Pa)

Unirradiated□　∆εa=2%, Nf:9.8x102

◆　∆εa=1%, Nf:4.6x103

Post irradiation□　∆εa=2%, Nf:1.3x102

◆　∆εa=1%, Nf:2.3x103


◆　∆εa=1%, Nf:4.6x103


◆　∆εa=1%, Nf:2.3x103

1 10

400-400

500600

700

800

-500

-600

-700

-800102 103 104

Number of cycles

Pea

k st

ress

(M

Pa)


◆　∆εa=1%, Nf:4.6x103


◆　∆εa=1%, Nf:2.3x103


◆　∆εa=1%, Nf:4.6x103


◆　∆εa=1%, Nf:2.3x103

F82H IEA HeatF82H IEA Heat

∆εa=2%∆εa=1%

∆εa=2%∆εa=1%

Data base development toward DEMO.

Ferromagnetic effects.

Development of high heat-resistant super steels and ODS steels.

Development of welding/joining technology and ODS-clad processing.

Compatibility with pressurized water and super critical water.

Technology/Engineering Integration for

Blanket/Reactor Components Fabrication

Performance Evaluationand Improvement under

Neutron EnvironmentBlanket Environment

SSTR or ASSTR

R & D of Ferritic Steels for FusionR & D of Ferritic Steels for Fusion-- fromfrom Fundamental Materials R & D tFundamental Materials R & D to Technology/Engineering Integration o Technology/Engineering Integration --


Kyoto University

Impurity level of NIFS- and US- HEATs

Fabrication of High Purity Large Products of VFabrication of High Purity Large Products of V--4Cr4Cr--4Ti4Ti(NIFS(NIFS--HEATsHEATs) )

Large V-4Cr-4Ti ingots with reduced impurity levels were produced in NIFS

Feasibility of recycling by quasi-remote (simply shielded) processing was verified

The resulting products were used for Round-robin test by international collaboration

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

NIFS-HEAT-2

Ingot

Improved

(Muroga, Nagasaka, Heo, NIFS)

US832665US832864NIFS-HEAT-1NIFS-HEAT-2

0

200

400

600

800

1000

1200

Al Nb Ag Mo(x10) (x100)

Impu

rity

Leve

l (w

ppm

)

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

less than*# not available

***#

Impurity level and recycling criteria

Improvement of Welding PropertyImprovement of Welding Property

Reduction of oxygen level in NIFS-HEAT resulted in significant enhancement of the mechanical property of the weld joint

Impact property of the TIG weld joint

Weld joint

0

2

4

6

8

10

12

14

-200 -150 -100 -50 0 50 100 150A

bsor

bed

ener

gy /

JTest temperature / C

NIFS-HEAT-1

USDOE-HEAT

(Muroga, Nagasaka, Heo, NIFS)

TIG weld joint of NIFS-HEAT-1

Improvement of Resistance to Radiation and Improvement of Resistance to Radiation and Oxidation by Addition of Si, Al and YOxidation 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 by exposure to airUniform elongation of irradiated alloys

0

0.02

0.04

0.06

0.08

0.1

773 873 973 1023

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

Oxidation temperature (K)

Wei

ght g

ain

(mg/

mm

2 )

Oxidation in air for 1 hr

(Abe, Satou, Fujiwara, Chuto, Tohoku University)

Melted


Kyoto University

Irradiation Effects Irradiation Effects on Mechanical Properties of SiC/SiC on Mechanical Properties of SiC/SiC

1: 500C, HFIR2: 400C, HFIR 3: 200-500C, HFIR4: 300-500C, JMTR5: 430-500C, EBR -II

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

6: 300C, HFIR7: 800C, HFIR8: 800C, JMTR9: 740C, HFIR10: 630, 1020C, ETR

0

1

0.1 1 10 100Neutron Dose [dpa-SiC]

S uIr

rad./S

uUni

rrad

.

3

2

2

7

6

7

7

4

5

5

5

5

2

2

3

1

1

1

888

82

2

6 7

710 9

7

7

6

Improvements inImprovements inFiber and processingFiber and processing

Improvement in Thermal Stress Figure of MeritImprovement in Thermal Stress Figure of Merit-- by LPSby LPS--SiC/SiC SiC/SiC --

0

5

10

15

20

25

30

35

40

45

50

200 700 1200 1700

Temperature [K]

M [k

W/m

]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)

LPS-SiC/SiCMass-production grade

( )E

KM

th

thUTS

ανσ −

=1


Kyoto University


Kyoto University

SiC/SiC R & D Goals and StatusSiC/SiC R & D Goals and Status-- 2000 2000 --

TAURO

ARIES-AT

D-SSTR

CERASEP

CREST-CVI

?

?

UnirradiatedTensile Strength

(500MPa/FS)

Young’smodulus

(300GPa/FS)

Mass Density(3.0Mg/m3/FS)

UnirradiatedThermal

Conductivity(20W/m-K/FS)

UnirradiatedUpper

Temp. Limit(1500℃/FS)

Additional IssuesFatigue lifetimeCreep strength

JoiningHermetic sealing

Chemical compatibilityTritium retentionFabrication costImpurity control

IrradiatedTensile

Strength (500MPa/FS)

IrradiatedUpper

Temp. Limit (1500℃/FS)

IrradiatedThermal

Conductivity (20W/m-K/FS)

?

TAURO

ARIES-AT

D-SSTR

CERASEP

CREST-CVI

?

?

UnirradiatedTensile Strength

(500MPa/FS)

Young’smodulus

(300GPa/FS)

Mass Density(3.0Mg/m3/FS)

UnirradiatedThermal

Conductivity(20W/m-K/FS)

UnirradiatedUpper

Temp. Limit(1500℃/FS)

Additional IssuesFatigue lifetimeCreep strength

JoiningHermetic sealing

Chemical compatibilityTritium retentionFabrication costImpurity control

IrradiatedTensile

Strength (500MPa/FS)

IrradiatedUpper

Temp. Limit (1500℃/FS)

IrradiatedThermal

Conductivity (20W/m-K/FS)

?

Major Subjects

(1)(1) High Z plasma facing materialsHigh Z plasma facing materials and interaction with plasmaNagoa 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 materialsTohoku Univ., Kyushu Univ., NIFS, Kagoshima Univ., etc.

(LHD Joint Projects)

(4) H and He irradiation experiments of W-coated materials with plasma simulators

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)

PWI and Plasma Facing Materials PWI and Plasma Facing Materials Institute of Advanced Energy

, Kyoto University

Deposit of TRIAMDeposit of TRIAM--1M 1M TokamakTokamakFusion Group RIAM Kyushu University

TRIAM-1M deposit

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

Fine fcc crystals, 1nm0

5 1017

1 1018

1.5 1018

2 1018 Mo+O2

on SUS

Mo BulkSUS

D2

des

orp

tio

n[D

2/m2 s

]

D injection：RT, 3x1021 D/m2

300 400 500 600 700 800 900 1000 1100Temperature[K]

TDS of D2

0

5 1017

1 1018

1.5 1018

2 1018 Mo+O2

on SUS

Mo BulkSUS

D2

des

orp

tio

n[D

2/m2 s

]

D injection：RT, 3x1021 D/m2

300 400 500 600 700 800 900 1000 1100Temperature[K]

300 400 500 600 700 800 900 1000 1100Temperature[K]

TDS of D2

Loss

Fusion PlasmaTest Blanket

Main Fuel CycleDeuterium

Tritium Flow in Fusion Reactor

Ash

EmergencyRecovery

(Heat, Tritium)

Loss

Loss

Loss

FUEL CYCLE!, SAFETY!

Tritium

ITER

( I ) ( II ) ( III )Preparatory Plasma D-T Burning Plasma Controlled Plasma

Experiment Experiment Operation

preparatorytritiumstudy

safetyconfinement

biologicalassessment

safety confinement(remote control)

environmentalassessment

fueling systemeach

unit operation

fueling systemwith

one throughmode

fueling systemremote controland continuous

operation

self-sustainedfuel cycle

improved fueling system

(preparation, settlement, measurement, control, regulation, waste treatment)

Where are we struggling ?Where are we struggling ?--Radiation EffectsRadiation Effects--

5 nm

Peak damage state iniron cascades at 100K

50 keVPKA(ave.

fusion)

10 keV PKA(ave. fission)

0.04 psec 0.1 psec

0.2 psec 0.26 psec


Kyoto University

(S.Zinkle et al.)

ConclusionConclusion

Fusion Engineering Activities in Japan are quite active Fusion Engineering Activities in Japan are quite active and efficient under the newly unified structure, MEXT.and efficient under the newly unified structure, MEXT.

Near term issues, for ITER, and long term issues, for Near term issues, for ITER, and long term issues, for DEMO and Power reactor are simultaneously carried DEMO and Power reactor are simultaneously carried out, well balanced and well managed condition.out, well balanced and well managed condition.

Fusion Engineering Activities in Japan will be Fusion Engineering Activities in Japan will be strengthened and accelerated with the decision of the strengthened and accelerated with the decision of the ITER invitation to Japan, in the near future. ITER invitation to Japan, in the near future.

Fusion Technology R & D in JapanITC-12/APFA'01 December 10-14, 2001 Toki, Gifu, Japan Fusion...

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Transcript of Fusion Technology R & D in JapanITC-12/APFA'01 December 10-14, 2001 Toki, Gifu, Japan Fusion...