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Key issues in LHD-type reactor FFHRKey issues in LHD-type reactor FFHRand present design activities and present design activities

Akio SAGARAAkio SAGARA

collaborating withcollaborating with S. Imagawa, T. Muroga, T. Uda, K. Yamazaki, K. Watanabe, S. Imagawa, T. Muroga, T. Uda, K. Yamazaki, K. Watanabe,

O. MitaraiO. Mitarai1)1), T. Kunugi, T. Kunugi2)2), S. Satake, S. Satake3)3), H. Matsui, H. Matsui4)4), , H. HasizumeH. Hasizume5)5), A. Shimizu, A. Shimizu6)6), S. Fukada, S. Fukada6)6), ,

S. TanakaS. Tanaka7)7), T. Terai, T. Terai7)7), Dai-Kai Sze, Dai-Kai Sze8)8), Y. Wu, Y. Wu9)9) and FFHR design group and FFHR design group

National Institute for Fusion Science, Toki 509-5292, JapanNational Institute for Fusion Science, Toki 509-5292, Japan1)Kyushu Tokai University, Kumamoto 862, Japan1)Kyushu Tokai University, Kumamoto 862, Japan

2) Kyoto University, Kyoto 6060-8501, Japan2) Kyoto University, Kyoto 6060-8501, Japan3) Toyama University, Toyama 930-8555, Japan3) Toyama University, Toyama 930-8555, Japan

4) Institute for Materials Research, Tohoku Univ., Sendai 980, Japan4) Institute for Materials Research, Tohoku Univ., Sendai 980, Japan5) Tohoku University, Sendai 980, Japan5) Tohoku University, Sendai 980, Japan

6) Kyushu University, Kasuga , Fukuoka 816, Japan6) Kyushu University, Kasuga , Fukuoka 816, Japan7) University of Tokyo, Tokyo 113, Japan7) University of Tokyo, Tokyo 113, Japan

8) Argonne National Laboratory, IL 60439, USA8) Argonne National Laboratory, IL 60439, USA9) IPP, Chinese Academy of Sciences, Hefei, Anhui, 230031, China9) IPP, Chinese Academy of Sciences, Hefei, Anhui, 230031, China

Japan-US Workshop onJapan-US Workshop onFusion Power Plants and Related Advanced Technologies with participation of EUFusion Power Plants and Related Advanced Technologies with participation of EU

October 9-10, 2003 , UCSD,October 9-10, 2003 , UCSD,

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Heliotron-H T-1 UETOR-M MSR CT6 HSR FFHR-1 MHR-S SPPSMain field coils 2 3 (modular) 3 (modular) 2(modular) 2 50 (modular) 3 2(modular) 32(modular)

Toroidal field periods 15 20 6 6 6 5 18 10 4Major radius 21m 29.2m 24.1m 20.2m 6.57m 20m 20m 16.5m 13.95m

Average plasma radius 1.8m 2.3m 1.7m 1.8m 1.74m 1.6m 2m 2.35m 1.6mToroidal field on axis 4T 5T 5.5T 6.4T 6T 5T 12T 5T 4.95T

Average beta 6% 3.54% 5% 4% 4.70% 4-5% 0.70% 5% 5%Fusion output 3.4GW 4.3GW 5.5GW 4GW 1.8GW 2-3GW 3GW 3.8GW 2.29GW

T-breeder Li2O Li Li17Pb83 6Li17Pb83 Flibe LiStructure material SUS HT-9 HT-9 JLF-1 V

References 7, 8 9 10 11 12 13 15, 16 17 181974-1982 1978 1981 1981 1989 - 1992 - 1995 - 1996 - 1997

Kyoto-U MIT Wisconsin-U Los Alamos ORNL IPP-Garching NIFS NIFS UCSD

Princeton

A.Sagara, J.Plasma Fusion Research, 74(1998) pp.947-951.

FFHR-2 (1998 - ): LHD-type, 10m, 10T, 1.8%, 1GWIssues on compactness

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EnergyMultiplication

Need High Temp.Energy Extraction

We need High Power Density

thfusion MPMOiC

COE

Availability&replacement cost

資本費 燃料費 運転費

Need LowFailure Rate

年間経費年間発電量

=

発電容量 設備利用量

After M. Abdou (in APEX)

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Importance of Weight Power Importance of Weight Power DensityDensityJAERI

18.2m

VECTOR-OPT

Weight of Reactor Machine (ton)

We

igh

t P

ow

er D

ens

. (k

WF/t

on

)

0 300000

ARIES-ST

ARIES-RSA-SSTR2

SSTRARIES-I

DREAM

CREST

ITER

VECTOR-OPT

VECTOR-A2

VECTOR-ZERO

JT-60

PF Coil Midplane Positioning

2000010000

100

200

300

30 m

ITER

26m

ARIES-RS

20 m

ARIES-ST

By S.NishioBy S.Nishio

FFHR-2FFHR-2

11GWthGWth

33GWthGWth

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Main issues for Helical Demo

1. Enhancement of energy confinement 2. Data base at reduced , outer shift,etc. 3. Reduction of heat load on the wall( loss) 4. Divertor pumping systems (for ash) 5. Fueling methods (to the center) 6. Heating devices (selection of method) 7. High performance blanket systems 8. Methods of maintenance & replacement 9. Advanced SC magnet systems 10. Structure materials and divertor materials11. Reduction of construction cost12. Reduction of COE

compactcompact

Blanket spaceBlanket space

Key issue

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by K.Watanabe [ref. Nuclear Fusion, 41(2001) 63.]

Outward shift of the magnetic axis

0

0.2

0.4

0.6

0.8

-10

0

10

20

30

3.4 3.5 3.6 3.7 3.8 3.9 4 4.1

cc-p

(

m) IB

SC (kA

)

Rax

(m)

Calculation @ R0=3.9m, B

0=1.5T, ~0.4%

cc-p

-inner

cc-p

-outer

IBSC

~ R0B

0

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FFHRDesign Concept& Optimization

Molten-salt blanket with low MHD effect

Enhancing inherent safety & Thermal efficiency

Plasma

100 cm0

Self-cooled T breeder

Radiation shield Vacuum vessel

SC magnet

LCFS

First wall

550°C coolant out

SOL

Thermal shield 450°C coolant in

20°C

7 53

T boundary&

Flibe

217 5

JLF-1(30vol. %)

1 1

&

Carbon 10 ~ 20

JLF-1 + B4C (30 vol.%)

Be (60 vol.%)

Simplification of supporting structure

-0.3

-0.2

-0.1

0

0.1

0.2

0.3

0.4

0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3

< f

a >

/ (B

0I H

)

ac/H=4

2

LHD

FFHR-142

W

H

ac

= 3

@W/H=2

= 2

coil cross-section

NIFS-960812 / S.I.

Coil pitch parameter c =(m/ )( ac /R)

FFHR-2

a c/H W/H < f a > (MN/m)

LHD 3.4 1.74 10.7FFHR-1 2.3 2.0 96.7FFHR-2 3.0 2.0 124.5

Reduction of magnetic force

case C: Ti(0)=29keV, n(0)=1.5x1020m-3, < > =4.5 %, hH=3.5

case B: Ti(0)=24keV, n(0)=1.9x1020m-3, < > =2.2 %, hH=2.25

Ignition conditions for the 3GW fusion outputcase A: Ti(0)=22keV, n(0)=2.0x1020m-3, < > =0.7 %, h

H=1.5

0

5

10

15

20

0 5 10 15 20 25 30Mag

net

ic f

ield

on

axis

B 0

(

T)

C

10T

Major radius R (m)

B

12T

8T

=0.5m 1m 1.5m

=3 m =18R/a

p=10

R/ac= 6

NIFS-930901 / K.W & O.Mi

j=27A/mm2

A

Bmax

= 15T

Bmax (IHJ)1/2

High B design

-2

-1

0

1

2

2 3 4 5 6

Z (

m)

R (m)

LHD Rax=3.6m, Bq=100%,

Expansion of blanket area

m

l

ac

R

decrease

Continuos winding helical coil

issue

SC materials &Current density

issue

MHD stabilityissue

Heat transfer & structure materials

issue

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self-cooling Flibe

10 m9876

He cooling

helical coils

radiation shield & reflector

thermal insulation

R

T-breeder

cylindrical support

poloidal coils

vacuum chamberdivertor target

core plasma

FFHR-2FFHR-2

A.Sagara et al., 17th IAEA _FTP-03 (1998) and ISFNT-5(1999, Rome) Qp~30, th~38%, availability~0.8

. Parameters LHD FFHR-1 FFHR-2 . major radius : R 3.9 20 10 m av. plasma radius : <ap> < 0.65 2 1.2 m fusion power : Pf (GW) - 3 1 GW external heating power : Pex < 20 100 100 MW neutron wall loading : Pn - 1.5 1.5 MW/m2 toroidal field on axis : B0 4 12 10 T average beta : < > > 5 0.7 1.8 % enhancement factor of LHD 1.5 2.5 plasma density : ne(0) 1.E2 2.E20 2.8E20 m-3 plasma temperature : Te(0) > 10 22 27 keV effective ion charge : Zeff 1.5 1.5 alpha heating efficiency : h - 0.7 0.7 alpha density fraction : f - 0.05 0.05 synchrotron reflectivity : Reff - 0.9 0.9 hole fraction : fh - 0.1 0.1 av. heat load on divertor < 10 1.6 1.5 MW/m2 . number of pole : : 2 3 2 toroidal pitch number : m 10 18 10 pitch parameter : 1.12<1.25 <1.37 1 1.15 coil modulation : + 0.1 0 + 0.1 av. helical coil radius : <ac> 0.975 3.33 2.30 m coil to plasma clearance : L 0.03 1.1 0.70 ~ 1.25 m coil current : IH 7.8 66.6 50 MA/coil coil current density : J (53) 27 25 A/mm2 max. field on coils : Bmax (9.2) 16 13 T stored magnetic energy 1.64 1290 147 GJ . . construction cost 50 Byen .

Neutron wall loading is reduced to n = 1.5MW/m2 (120dpa/10y)

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Dose effect in SC coil under fast neutron irradiationDose effect in SC coil under fast neutron irradiation

H. Kodama et al., ICFRM-1, JNM 133&134 (1985) 819.

=10year x 3E+13 n/m2s

Should be reduced 5 orders from~ 5E+18 n/m2s

in case of 1.5MW/m2 ( for En > 0.1 MeV at the first wall)

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Dose effect in insulator due to gamma-rayDose effect in insulator due to gamma-ray

P.Komarek et al., ICFRM-3, JNM 155-157 (1988) 207.

1~2 order sever than SC

~3E+9 rad

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Self-cooled Tritium Breeding Blanket in FFHRSelf-cooled Tritium Breeding Blanket in FFHR

Plasma

100 cm0

Self-cooled T breeder

Radiation shield Vacuum vessel

SC magnet

LCFS

First wall

550°C coolant out

SOL

Thermal shield 450°C coolant in

20°C

7 53

T boundary&

Flibe

217 5

JLF-1(30vol. %)

1 1

&

Carbon 10 ~ 20

JLF-1 + B4C (30 vol.%)

Be (60 vol.%)

A.Sagara et al., Fusion Technol.,39 (2001) 753.

1.1

1.2

1.3

1.4

1.5

0 20 40 60 80 100

Loca

l TB

R

Li-6 Enrichment (%)

17cm (Be+Flibe) + 7cm (Flibe) + 10cm (Carbon)

TBR (total)

TBR (Li-6)

● ● Local TBRLocal TBR 〜〜 1.4.1.4.

● ● However, shielding However, shielding efficiencyefficiency for SC for SC magnet should be magnet should be improved more than improved more than one order.one order. 1.0

1.1

1.2

1.3

1.4

103

104

105

106

107

0 10 20 30 40 50

Loca

l TB

R

Shielding efficiency

Thickness of graphite reflector (cm)

-ray

fast neutron

TBR

17cm (Be+Flibe)+7cm (Flibe)natural Li

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Another approach using HTcSC for magnet coilsAnother approach using HTcSC for magnet coils

Even the shielding efficiency is poor,● temperature margin is wide,● controllability becomes high at high T operation, (because specific heat ~ T3)

T.Horiuchi et al., Fusion Technol. 8 (1985) 1654.

Conventional SC joint design

Re-mountable coilsusing HTcS.C. joints

H.Hashizume et al., J.Plasma Fusion Res. SERIES 5 (2001) 532.

● SC joint is promising and innovative.　　（ Ohmic heat on cryo. < a few ％ of Pf ）

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1750 2000150012501000

250

250

200

Helical coil

PF coil

FFHR-2a FFHR-2b

(modified)

Major radius 10 15 m

Average coil radius 2.30 3.45 m

Average plasma radius 1.2 1.8 m

Number of field periods 10 10

Field on axis 10 6 T

Max. field on coils 13 7.8 T

Electron Density(0) 3.0x1020 3.0x1020 [m-3]

Electron Temp(0) 15.0 15.0 [keV]

Av. Beta 0.96 2.66 [%]

Fusion Power 0.62 2.09 [GW]

Heating Power 94 333 [MW]

Energy Conf. Time 1.72 1.64 [s]

Increase of the FFHR2 configuration by 50%Reduction of the magnetic field to 6 T on axis • to provide more space for blanket and shield• to use NbTi in the helical windings.

Jc reduced

Modified FFHR-2Modified FFHR-2 pressented in 12th Intenational Toki Conference (2002)

IPP Garhing : H. Wobig, J. Kisslinger

One way for optimization.But 3 times of the weight

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Blanket in HSR 〜 1.5m

• He Cooled Li4SiO4 (KfK design)• He Cooled Pb-17Li (FZK design)• Water cooled Pb-17Li (ITER design)

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Tritium confinement boundary in FFHR

NBI

Fueling

Fuel Storage

TRS

※

Turbine

EVT

Stack

VPS

EGS

EGS

HePS

Blanket

SC Magnet

V V

TB1TB4

HX

BTRS

※

TB3TB2

BTRS : Blanket Trium Recovery system EVT : Expansion Volume Tank HePS : Helium Purification System

Divertor

Maintenance hall

Maintenance hall

Maintenance hall

VPS : Vacumm Pumping System TRS : Tritiume Recorvery & Storage System EGS : Exhaust Gas treatment System ACS : Air Cleanup System

970321 A.S

ACS

※

Tank

Pump

Availability to use the center space in a large R helical rectorAvailability to use the center space in a large R helical rector

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SC magnet

Self-cooled T breeder TBRlocal > 1.2

Radiation shield reduction > 5 orders Thermal shield

20°C

Vacuum vessel

T boundary&

Protection wall Ph =0.2 MW/m Pn =1.5 MW/m Nd =450 dpa/30y

22

Be

T storage

Pump

Pump

Tank

HXPurifier

T-disengager

14MeV neutron

Turbine

Liquid / Gas

Structural Materials

In-vessel Components

Blanket

Thermo- fluid

Tritium

June 6. 2003, A.Sagara

Safety & Cost

Chemistry

shielding materials .

high temp.surface heat flux

Core Plasma

Ignition access & heat flux O.Mitarai (Kyusyu Tokai Univ.)

Helical core plasma K.Yamazaki (NIFS)

Blanket system S,Tanaka (Univ. of Tokyo)

Thermo-mechanical analysis H.Matsui (Tohoku Univ.)

Helical reactor design / Sytem Integration A.Sagara (NIFS)

Thermofluid system K.Yuki (Tohoku Univ.)

Advanced thermofluid T.Kunugi (Kyoto Univ.)

Heat exchanger & gas turbine system A.Shimizu (Kyusyu Univ.)

T-disengager system S.Fukada, M.Nishikawa (Kyusyu Univ.)

Device system code H.Hashizume (Tohoku Univ.)

Thermofluid MHD S.Satake (Tokyo Univ. Sci.)

Advanced first wall T.Norimatsu (Osaka Univ.)

Reactor design activity in NIFS collaborationReactor design activity in NIFS collaboration

• on going cf. NIFS annual repo.

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TTohokuohoku-N-NIFSIFS T Thermofluidhermofluid loop loop for molten saltfor molten salt

Now HTS (simulant for Flibe) is used.

Max.20L/min @ 600°CMax.20L/min @ 600°C~ 0.1m~ 0.1m3 3 @ 0.7MPa.@ 0.7MPa.

40 kW

(Fabricated by IHI Co., Ltd.)

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Heat transfer enhancer at Heat transfer enhancer at low flow ratelow flow rate… … Packed-bed tube in TNT loopPacked-bed tube in TNT loop

Fig.Re vs heat transfer characteristics (at 300 mm)

Fig.Pump frequency vs heat transfer coefficient (at 300 mm)

About About 33 times higher than times higher thanturbulent heat transferturbulent heat transfer

300 mm

HERE

Flow

Fig. Flow structure near wall(Result of 2D numerical simulation)

19.5mm

5mm

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MHD effectsMHD effects Evaluation methodEvaluation method for degradation of the insulator, for degradation of the insulator,

induced pressure increase.induced pressure increase. Proposing new methodProposing new method to reduce the MHD pressure to reduce the MHD pressure

increase.increase. Clarification of feasibilityClarification of feasibility for the liquid blanket system.for the liquid blanket system.

2003 2003 NIFS Collaboration ProgramNIFS Collaboration ProgramH.Hashizume (Tohoku Univ.)H.Hashizume (Tohoku Univ.)

K.Muroga (NIFS)K.Muroga (NIFS)A.Sagara (NIFS)A.Sagara (NIFS)

S.Tanaka (Tokyo Univ.)S.Tanaka (Tokyo Univ.)H.Horiike (Osaka Univ.)H.Horiike (Osaka Univ.)T.Kunugi (Kyoto Univ.)T.Kunugi (Kyoto Univ.)

Other 8 reserchersOther 8 reserchers

JUPITER-II projectJUPITER-II projectTask 1-1-ATask 1-1-BTask 1-2-ATask 3-1

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Velocity profiles (Velocity profiles (BB=10[T])=10[T])

(a) Without Insulator(b) Coated with Insulator

[1] Flibe

) (

M-shape Flat-shape

9 9HT

i

By H. HashizumeBy H. Hashizume

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(a) Without Insulator

) (

[2] Lithium

M-shape Flat-shape

9 9HT

i

(b) Coated with Insulator

H.Hashizume et al., submitted to ISEM 2003: The 11th International H.Hashizume et al., submitted to ISEM 2003: The 11th International Symposium on Applied Electromagnetics and Mechanics, Versailles, FranceSymposium on Applied Electromagnetics and Mechanics, Versailles, France

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Parameter Requirements for D-Parameter Requirements for D-33He Helical ReactorsHe Helical ReactorsO.Mitarai O.Mitarai 1)1), ,

A. Sagara, S. Imagawa, Y.Tomita, K.Y.Watanabe and T.Watanabe A. Sagara, S. Imagawa, Y.Tomita, K.Y.Watanabe and T.Watanabe 2)2)

1) Kyushu Tokai University, 9-1-1 Toroku, Kumamoto, 862-8652 Japan, 2) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, 509-5292 Japan

Submitted to ITC-13, Dec. 2003 @Toki, JapanSubmitted to ITC-13, Dec. 2003 @Toki, JapanD-3He ignition is possible in the helical reactor with

R=14.47m, <ap>=2.585 m, Bo=6 T, and <>~18 %

for the confinement enhancement factor of HH = 2.4

with the external heating power of PEXT= 300 MW

for D:3He fuel ratio of 2:1.

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Summary

Within present-day technology, medium-size Within present-day technology, medium-size designs are one way for helical demo-reactors.designs are one way for helical demo-reactors.

However, innovative concepts such as However, innovative concepts such as re-mountable coils using HTcS.C. jointsre-mountable coils using HTcS.C. joints are very are very attractive for compactness of LHD-type reactors.attractive for compactness of LHD-type reactors.

Design activities in NIFS collaborations are expanding Design activities in NIFS collaborations are expanding intointo Thermofluid exp., MHD effects, D-Thermofluid exp., MHD effects, D-33He, (& IFE)He, (& IFE)..