japan-us workshop on march 16-18, 2009 at the university...

O i f FFHR d i ti itO i f FFHR d i ti itO i f FFHR d i ti itO i f FFHR d i ti it

Japan-US Workshop onFusion Power Plants and Related Advanced Technologies

with participation of EU and ChinaMarch 16-18, 2009 at the University of Tokyo in Kashiwa, JAPAN

Overview of FFHR design activityOverview of FFHR design activityOverview of FFHR design activityOverview of FFHR design activityAkio Sagara

with FFHR design group : S. Imagawa1, Y. Kozaki1, O. Mitarai2, T. Goto1, T. Tanaka1, T. Watanabe1, N. Yanagi1,

H. Tamura1, K. Takahata1, S. Masuzaki1, M. Shoji1, M. Kobayashi1, K Ni hi 1 T M 1 T N k 1 M K d 1 A Ni hi 1K. Nishimura1, T. Muroga1, T. Nagasaka1, M. Kondo1, A.Nishimura1,

H. Igami1, H. Chikaraishi1, S. Yamada1, T. Mito1, N. Nakajima1, S. Fukada3, H. Hashizume4, Y. Wu5, Y. Igitkhanov6, O. Motojima1

11National Institute for Fusion Science, Toki, Gifu, JapanNational Institute for Fusion Science, Toki, Gifu, JapanNational Institute for Fusion Science, Toki, Gifu, JapanNational Institute for Fusion Science, Toki, Gifu, Japan22Tokai University, Kumamoto, JapanTokai University, Kumamoto, Japan33Kyushu University, Fukuoka, JapanKyushu University, Fukuoka, Japan

44Tohoku University, Sendai, JapanTohoku University, Sendai, Japan55Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China.Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China.

66MM Pl kPl k I tit tI tit t füfü Pl h ikPl h ik IPPIPP EURATOM A G if ld GEURATOM A G if ld G

FFHR2m2FFHR2m23GWth3GWthFFHR2m2FFHR2m2

3GWth3GWth

66MaxMax--PlanckPlanck--Institut Institut fürfür PlasmaphysikPlasmaphysik, IPP, IPP--EURATOM Ass., Greifswald, GermanyEURATOM Ass., Greifswald, Germany

1 / 25

3GWth3GWth5 Tesla5 Tesla30,000 ton30,000 ton

3GWth3GWth5 Tesla5 Tesla30,000 ton30,000 ton

FFHR design collaborations FFHR design collaborations

Ignition accessConfinement scaling LHD Fusion Eng.R.C.

Fusion Research Network

Structural MaterialsBl k

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

K.Yamazaki(NIFS)

FFHR design / Sytem Integration

/ Replacement A.Sagara (NIFS)

Neutronics T.Tanaka, A.Sagara (NIFS)

gH.Yamada, Miyazawa (NIFS)

Magnetic structure T Morisaki Yanagi

Blanket material systemT.Muroga, T.Nagasaka, M.Kondo (NIFS)

Mag. material system A.Nishimura,

S C m agn et

S elf - cool ed T b ree der TB R lo cal > 1. 2

R adi ati on s h ield r educ tio n > 5 or der s T h erm al s hiel d

2 0 °C

Vacuu m ves s el

T b oun dar y&

P r otect io n wa ll Ph = 0 .2 M W / m Pn = 1 .5 M W / m N d = 4 50 dp a/3 0y

22

Be

T-d ise n ga ge r

14MeV neutron Liquid / G as

Blanket

Safety

shie ld ing materia ls.

high te mp.surface heat fluxCore Plasma

Blanket system T.Terai (Univ of Tokyo)

SC magnet & support

T.Morisaki, Yanagi (NIFS)

SC magnet & supprt S.Imagawa T.Mito Takahata

A.Nishimura,Y.Hishinuma (NIFS) Int.

CollaborationTITAN : TITAN : USA USA (ORNL, INL, (ORNL, INL,

T s to rag e

Pu m p

Pu mp

Tan k

H XPu rif ie r

Turbine

In-vessel Components

Thermo- fluid

Tritium

Safety& Cost

Chemistry

lasma (Univ. of Tokyo)

Virtual Reality tool N.Mizuguchi

Heating FuelingDivertor pumping

Takahata,Yamada, Tamura, (NIFS)

Cost evaluation Y. Kozaki(NIFS)

Safety T.Uda(NIFS)

((UCLA, UCSD)UCLA, UCSD)

CUP :CUP : China China (SWIP, ASIPP)(SWIP, ASIPP)

Int NetworkInt Networkfluid

Thermofluid system K.Yuki, H.Hashizume

(Tohoku Univ.)Advanced thermofluid

Heat exchanger & gas turbine system

T-disengager system S.Fukada, M.Nishikawa

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

Advanced first wall

g(NIFS)

Divertor pumping S Mas aki

Fueling Sakamoto (NIFS)

External heating O Kaneko Igami

Power supply H.Chikaraishi (NIFS)

Y. Kozaki(NIFS) Int. Network Int. Network Program : Program : EU EU (IPP), Russia, (IPP), Russia, Ukraine, Korea Ukraine, Korea (KAERI)(KAERI)

2 / 25April 10, 2008, A.Sagara

Advanced thermofluidT.Kunugi (Kyoto Univ.)

g yA.Shimizu (Kyusyu Univ.)

Advanced first wallT.Norimatsu (Osaka Univ.)

S.Masuzaki, Kobayashi (NIFS)

O.Kaneko, Igami(NIFS)

Role of Design Study to Helical DemoRole of Design Study to Helical Demo--Reactor Reactor based on LHD Projectbased on LHD Project

Role of Design Study to Helical DemoRole of Design Study to Helical Demo--Reactor Reactor based on LHD Projectbased on LHD Project

FFHR designFFHR designith R&D’ith R&D’with R&D’swith R&D’s

3 / 25

Two main features in FFHRTwo main features in FFHR(1)(1) QuasiQuasi--force free force free γγ optimizationoptimization

on continuous helical windingon continuous helical windingJ x Bsurf = 0, when ϑ = 45 deg.J x Bsurf = 0, when ϑ = 45 deg.

on continuous helical windingon continuous helical windingto reduce the magnetic hoop force to reduce the magnetic hoop force

(Force Free Helical Reactor: FFHR)(Force Free Helical Reactor: FFHR)large large maintenancemaintenance portsports

t d th bl k tt d th bl k tto expand the blanket spaceto expand the blanket space

γ =ml

a c

R

γ=tanθ1

2

γ=1.18γ=1.33

LHD Rax=3.6m, B q=100%, Φ = 0°

-1

0

Z (m

)

(2) Self(2) Self--cooled liquid Flibe cooled liquid Flibe (BeF(BeF22--LiF) blanketLiF) blanket

low MHD pressure losslow MHD pressure loss

4 / 254 / 214

-22 3 4 5 6

R (m)

low MHD pressure losslow MHD pressure losslow reactivity with airlow reactivity with airlow pressure operationlow pressure operationlow tritium solubilitylow tritium solubility

Presentation outline Presentation outline

1. 1. Blanket and divertor spaceBlanket and divertor space

Design windows and costDesign windows and cost

15

19981995FFHR1(l=3)

gg

Nuclear shield on SC coilsNuclear shield on SC coils

Reactor size optimizationReactor size optimization10

ld B

0 (T) FFHR2

Ap ~ 8γ =1.15

( 3)Ap ~ 10

γ =1

Reactor size optimizationReactor size optimization

New ignition regime New ignition regime

5

gnet

ic fi

el 2004 FFHR2m1

Ap ~ 8

FFHR2 2γ =1.15

outer shift

optimum ?ITERAp~3.1

ARIESARIES--CSCSAp~4.5Ap~4.5Design parametersDesign parameters

2. SC magnet and supports 2. SC magnet and supports

00 5 10 15 20 25

Mag LHD

Ap ~ 6

FFHR2m2Ap ~ 6γ =1.25

outer shift

inner shift?3. Blanket system integration3. Blanket system integration

4. Concluding remarks4. Concluding remarks

5 / 255 / 21

0 5 10 15 20 25Coil major radius R

c (m)

Issues on blanket and divertor spaceIssues on blanket and divertor spaceIssues on blanket and divertor spaceIssues on blanket and divertor space(a) (b)

HoweverHoweverFor For αα--heating efficiency over heating efficiency over 90%, the importance of the 90%, the importance of the ergodic layers has been found ergodic layers has been found by collisionless orbitsby collisionless orbits

For For αα--heating efficiency over heating efficiency over 90%, the importance of the 90%, the importance of the ergodic layers has been found ergodic layers has been found by collisionless orbitsby collisionless orbitsby collisionless orbits by collisionless orbits simulation of 3.52MeV alpha simulation of 3.52MeV alpha particles.particles.

by collisionless orbits by collisionless orbits simulation of 3.52MeV alpha simulation of 3.52MeV alpha particles.particles.

By T WatanabeTo remove the interference between the To remove the interference between the first walls and the ergodic layers first walls and the ergodic layers surrounding the last closed flux surrounding the last closed flux

To remove the interference between the To remove the interference between the first walls and the ergodic layers first walls and the ergodic layers surrounding the last closed flux surrounding the last closed flux

By T. Watanabe

su ou d g t e ast c osed usu ou d g t e ast c osed usurface, helical xsurface, helical x--point divertor (HXD) point divertor (HXD) has been proposed.has been proposed.

su ou d g t e ast c osed usu ou d g t e ast c osed usurface, helical xsurface, helical x--point divertor (HXD) point divertor (HXD) has been proposed.has been proposed.

Helical X point Divertor (HXD) T Morisaki et al

6 / 256 / 21

Helical X-point Divertor (HXD) T. Morisaki et al., FED 81 (2006) 2749.

By T. Watanabe

7 / 257 / 21

Three candidates are proposed to increase blanket space > 1.1 mThree candidates are proposed

to increase blanket space > 1.1 m

1. Reduction of the inboardthe inboardshielding thickness

i WCi WC

2. Improvement of the symmetrythe symmetryof magnetic surfaces by increasing the current densityusing WCusing WC

14

1015

1016 JLF-1 JLF-1 (70vol.%)

+B4C (30vol.%)

MeV

)

14

1015

1016 JLF-1 JLF-1 (70vol.%)

+B4C (30vol.%)

MeV

)

14

1015

1016 JLF-1 JLF-1 (70vol.%)

+B4C (30vol.%)

MeV

)

by increasing the current density at the inboard side of the helical coilsby splitting the helical coils.by splitting the helical coils.N Yanagi et al in ITC-18 conference

agne

t)

asm

a

1011

1012

1013

1014 B4C

WC

on flu

x （>0.1

M

/cm

2/s)

agne

t)

asm

a

1011

1012

1013

1014 B4C

WC

on flu

x （>0.1

M

/cm

2/s)

agne

t)

asm

a

1011

1012

1013

1014 B4C

WC

on flu

x （>0.1

M

/cm

2/s)

N. Yanagi et al., in ITC-18 conference.

Breedinglayer

[Flibe+Be](32cm)

Radiationshield

(60 cm)

(Ma

Pla

7

108

109

1010

Fas

t neutr

o (n/

Breedinglayer

[Flibe+Be](32cm)

Radiationshield

(60 cm)

(Ma

Pla

7

108

109

1010

Fas

t neutr

o (n/

Breedinglayer

[Flibe+Be](32cm)

Radiationshield

(60 cm)

(Ma

Pla

7

108

109

1010

Fas

t neutr

o (n/

( )

180 200 220 240 260 280 300107

Radial position（cm)

( )

180 200 220 240 260 280 300107


( )

180 200 220 240 260 280 300107


25cmthinner

Rc = 15.0 m, ac = 3.0 m, γ = 1.0Baxis = 6 T, ap = 1.5 m, W = 143.2 GJ

Smaller size and higher field with γ = 1

8 / 25

thinner FFHR-2S Type-Ig γ

(reduction of total mass)

K. Saito et al., in ITC-18 conference.ICRF

antenna

Three candidates are proposed to increase blanket space > 1.1 mThree candidates are proposed

to increase blanket space > 1.1 m

3. Enlargement of reactor size3. Enlargement of reactor size3. Enlargement of reactor size3. Enlargement of reactor sizeNeutron wall loading should be kept at < 2 MW/mNeutron wall loading should be kept at < 2 MW/m22

15 200

T)

W

J = 25 ~ 33 A/mm 2

Γn = 1.3 ~ 1.5 MW/m 2

blanket space > 1 m Rp=16m (Rc=17.3m) is selected.(simple expansion of LHD gives too large R )

Neutron wall loading should be kept at 2 MW/mNeutron wall loading should be kept at 2 MW/m

10

100

150

C (G

$),

Bo (

T

Wg (G

J), CO

Wg (magnetic energy)

COEB0

Thermal insulation for SC magnet

gives too large Rc.)

5

50

100

Pf (

GW

), T

CC

OE

(mil/kW

h)TCC (total capital cost)

0 05 10 15 20

P

C il j di R ( )

Pf (fusion output)

FFHR2FFHR2m1

FFHR2m2

(I l d

9 / 25

Coil major radius R c (m) (Includes SOL and V/V)

A. Sagara et al., in ISFNT-8, FED 83 (2008) 1690.


1. Blanket and divertor space

Design windows and costDesign windows and costgg

Nuclear shield on SC coilsNuclear shield on SC coils

Reactor size optimizationReactor size optimizationReactor size optimizationReactor size optimization

New ignition regime

Design parameters

2. SC magnet and supports

3. Blanket system integration

4. Concluding remarks

10 / 25

Design windows and costDesign windows and costDesign windows and costDesign windows and cost

β=5%, γ = 1.20, j=26A/mm2 are selected

Y. Kozaki et al., in IAEA-2008 conference.

W GJ (β 5%, Pf 4GW)W GJ (β 4%, Pf 3GW)

COE3GW

90

100

Pf3GW

160

180

160

180COE (mill/kWh)

4GW

80

90

4GW

4.75GW

140

160

140

160

60

70

14.0 14.5 15.0 15.5 16.0 16.5100

120

14 15 16 17

HF<=1.16

100

120

14 15 16 17

The design windows limited with ∆d≥1.1m,Hf≤1.16, W<160GJ, depending on γ and β.

Rp mRp m Rp m

The COEs of helical reactors, which depend on Rp, γ and β, show the bottom as the result

11 / 25

f p g γ βHf=1.16 means the 1.2 times value achieved inLHD experiment. j=26A/mm2 is premised.

show the bottom as the result of the trade-off between the Cmag and Cbs, i.e., B0 versus plasma volume.

Issues on nuclear shield on SC coilsIssues on nuclear shield on SC coilsIssues on nuclear shield on SC coilsIssues on nuclear shield on SC coils

Acceptable level achievedAcceptable level achieved

Discrete Pumping with Semi-closed Shield (DPSS) has been proposedA. Sagara et al., in ISFNT-8, FED 83 (2008) 1690.

Acceptable level achievedAcceptable level achievedCover rate > 90%Cover rate > 90%Fast neutron < 1E22 n/mFast neutron < 1E22 n/m22 in 30 yearsin 30 yearsMax. nuclear heating < 0.2 mW/cmMax. nuclear heating < 0.2 mW/cm33

Total nuclear heating ~ 40 kWTotal nuclear heating ~ 40 kWCryogenics power ~ 12 MW (1% of PCryogenics power ~ 12 MW (1% of Pff))

3D design with 3 gDPSS

12 / 25

Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2inner shifted (equivalent to Rax=3.6m in LHD),

Vacuum Magnetic γ = 1.20, α = +0.1Rp=16m, Rc=17.3m, B0=4.84T, j=26A/mm2 , WWmagmag=167.7GJ=167.7GJ

PC positionPC position

Vacuum Magnetic Surface

expandedRpRb

Rc Type AType A

13 / 25aPC = 8.2 m, HC : 38.72 MA, OV : -23.47 MA, IV : -17.04 MA

inner shifted (equivalent to Rax=3.6m in LHD),Vacuum Magnetic

Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2

R

γ = 1.20, α = +0.1Rp=16m, Rc=17.3m, B0=4.84T, j=26A/mm2 , WWmagmag=149.1GJ=149.1GJ



movedRp

RbRc Type BType B

14 / 25HC : 38.72 MA, OV : -18.39 MA, IV : -13.94 MA

inner shifted (equivalent to Rax=3.6m in LHD),Vacuum Magnetic

Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2Reactor size optimization of FFHR2m2

γ = 1.20, α = +0.1Rp=15.5m, Rc=16.7m, B0=4.90T, j=25A/mm2 , WWmagmag=135.6GJ=135.6GJ



Type B’Type B’WC shield at inboard

15 / 25HC : 37.87 MA, OV : -18.01 MA, IV : -14.79 MA



Design windows and costg

Nuclear shield on SC coils


New ignition regime New ignition regime

Design parametersDesign parameters

2. SC magnet and supports

3. Blanket system integration


16 / 25

New control method in a thermally unstable New control method in a thermally unstable ignition regimeignition regimeNew control method in a thermally unstable New control method in a thermally unstable ignition regimeignition regime O. Mitarai et al., in IAEA-2008 conference.

ProportionalProportional--IntegrationIntegration--Derivative (PID) controlDerivative (PID) controlThe error of the fusion power The error of the fusion power with an opposite signwith an opposite sign of of

(P(P )) (P(P PP )) t bili th th l i t bilitt bili th th l i t bilite(Pe(Pff)= )= -- (P(Pfofo--PPff)) can stabilize the thermal instability can stabilize the thermal instability through fuelingthrough fueling. .

O Mitarai al Plasma and Fusion Research Rapid

SSDTDT(t)=0 (t)=0 if Sif SDTDT(t)<0(t)<0

O. Mitarai al. Plasma and Fusion Research, Rapid Communications, 2, 021 (2007) .

τταα*/*/ττEE=3 ~ 5=3 ~ 5ττ */*/ττ =2 ~ 8=2 ~ 8ττpp //ττEE=2 ~ 8=2 ~ 8

But, But, effectively effectively

17 / 25

reduced due reduced due to burningto burning

Design parametersDesign parametersDesign parametersDesign parametersDesign parameters LHD FFHR2 FFHR2m1FFHR2m2 SDC 1 1Design parameters S C

Polarity l 2 2 2Field periods m 10 10 10Coil pitch parameter γ 1.25 1.15 1.15Coil major Radius Rc m 3.9 10 14.0Coil minor radius a m 0 98 2 3 3 22

210

1.2017.34 16

1

1.1

ZOV=ZIVZOV/Rc=3.8/14Z /R =4 5/14on

ly

LHD

aPC to 0.95Rc

Coil minor radius ac m 0.98 2.3 3.22Plasma major radius Rp m 3.75 10 14.0Plasma radius <ap> m 0.61 1.24 1.73Plasma volume Vp m3 30 303 827Blanket space ∆ m 0.12 0.7 1.1M i fi ld B T 4 10 6 18

4.1616.02.3517441.054 84

0.9

ZOV/Rc=4.5/14ZOV/Rc=5/14

rgy,

W/W

HC

o

aPC to Rc

Type A

Magnetic field B0 T 4 10 6.18

Max. field on coils Bmax T 9.2 14.8 13.3Coil current density j MA/m2 53 25 26.6Magnetic energy GJ 1.64 147 133Fusion power P GW 1 1 9

4.84

11.926

3 0 7

0.8

Sto

red

Ene

ZIV (RIV) increase.Type B

Fusion power PF GW 1 1.9Neutron wall load Γn MW/m2 1.5 1.5External heating powePext MW 70 80 43 100α heating efficiency ηα 0.7 0.9 0.9 0.9Density lim.improvement 1 1.5 1.5 7.5

31.5

0.6

0.7

0.1 0.15 0.2 0.25 0.3 0.35

γ=1.20, Rp/Rc=3.6/3.9

Bleak(2.5 Rc) < 0.5 mTaIV=aOV, RIV<Rc-ac

H factor of ISS95 2.40 1.92 1.92 1.60Effective ion charge Zeff 1.40 1.34 1.48 1.55Electron density ne(0) 10^19 m-3 27.4 26.7 17.9 83.0Temperature Ti(0) keV 21 15.8 18 6.33Plasma beta <β> % 1.6 3.0 4.40 3.35

(aPC-ac)/Rc

In SDC, divertor heat load is

18 / 25

Plasma beta <β> % 1.6 3.0 4.40 3.35Plasma conduction lo PL MW 290 453 115Diverter heat load ΓdivMW/m2 1.6 2.3 0.6Total capital cost G$(2003) 4.6 5.6COE mill/kWh 155 106 93

7.0

drastically reduced (~1/4).



Design windows and costg

Nuclear shield on SC coils


New ignition regime

Design parameters

2. SC magnet and supports 2. SC magnet and supports

3. Blanket system integration3. Blanket system integration


19 / 2519 / 21

Base design of Base design of CICC magnet systemCICC magnet system

Base design of Base design of CICC magnet systemCICC magnet system

101

CICC magnet systemCICC magnet systemCICC magnet systemCICC magnet systemNuclear heating in FFHR2m1S. Imagawa et al., in IAEA-2008 conference

10-1

100

101

ng (

mW

/cm3

)

Neutron+Gamma Gamma

5 times

10-3

10-2

Nucle

ar h

eat

in

-20 0 20 40 60 80 10010-4N

Radial position (cm)

By T. Tanaka

Max. cooling path is 500 m for the Max. cooling path is 500 m for the nuclear heat of 1 mW/cmnuclear heat of 1 mW/cm33. . This value is 5 times larger on theThis value is 5 times larger on the

By T. Tanaka

20 / 25

This value is 5 times larger on the This value is 5 times larger on the FFHR magnets. FFHR magnets. GammaGamma--ray heating is dominant.ray heating is dominant.

IndirectIndirect--cooled helical coil systemcooled helical coil system(alternative for CICC)

with quench protection by internal dumping

IndirectIndirect--cooled helical coil systemcooled helical coil system(alternative for CICC)

with quench protection by internal dumpingwith quench protection by internal dumpingwith quench protection by internal dumping

Stress analysis inside of the coilStress analysis inside of the coilH Tamura et al in ITC 18 conference

SS316SS316 H. Tamura et al., in ITC-18 conference

Indirect Cooling

Circumferential strain distribution

Cross-sectional structure of the helical coil

Quench protection by internal dumpingQuench protection by internal dumping Hoop stress

Quench back with a secondary circuitto increase a decay time constant > t=20 s

K. Takahata et al., ITC-17.Quench protection by internal dumpingQuench protection by internal dumping Hoop stress

distribution

21 / 25

to reduce a transient voltage Vmax=10 kVto avoid a serious hot spot < 150 K Confirmed to be within the permissible

range.

LHDLHD--type support post for FFHR type support post for FFHR LHDLHD--type support post for FFHR type support post for FFHR

Gravity per support Gravity per support 16 000 t / 30 l 530 t16 000 t / 30 l 530 t

By H. Tamura

= 16,000 ton / 30 legs ~ 530 ton.= 16,000 ton / 30 legs ~ 530 ton.Thermal contraction < max. 55 mmThermal contraction < max. 55 mmTotal heat load to 4K ~ 0.34 kWTotal heat load to 4K ~ 0.34 kW( 1/20 of stainless steel post)( 1/20 of stainless steel post)( p )( p )

(Carbon Fiber Reinforcement Plastic)

6.18

5.66

3

22 / 25

Wide maintenance ports Wide maintenance ports Wide maintenance ports Wide maintenance ports

23 / 25

Fusion Research Network

Blanket system integration Blanket system integration by broad R&D collaboration activitiesby broad R&D collaboration activities

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

K Y ki(NIFS)

FFHR design / Sytem Integration

/ Replacement A Sagara (NIFS)

Neutronics T.Tanaka, A.Sagara (NIFS)

Confinement scaling H.Yamada, Miyazawa (NIFS)

LHDBlanket material system T.Muroga, T.Nagasaka, M.Kondo (NIFS)

Fusion Eng.R.C.Fusion Research Network

S C m agn et

S elf - cool ed T b ree der TB R lo cal > 1. 2

R adi ati on s h ield r educ tio n > 5 or der s T h erm al s hie l d

2 0 °C

Vacuu m ves s e l

T b oun dar y&

P r otect io n wa ll Ph = 0 .2 M W / m Pn = 1 .5 M W / m N d = 4 50 dp a/3 0y

22

Be14MeV neutron Liquid / G as

Structural MaterialsBlanket

shie ld ing mater ia ls.

C

)K.Yamazaki(NIFS)

Blanket system

A.Sagara (NIFS)

SC magnet & support

( )

Magnetic structure T.Morisaki, Yanagi (NIFS)

SC magnet & supprt S Imagawa

Mag. material system A.Nishimura, Y.Hishinuma (NIFS)

2 0 °C

T s to rag e

Pu m p

Pu mp

H XPu rif ie r

T-d ise n ga ge r

T bi

In-vessel Components

Th

Tritium

Safety & Cost

Chemistry

high te mp.surface heat fluxCorePlasma

Blanket systemT.Terai (Univ. of Tokyo)

Virtual Reality tool

pp

Heating FuelingDivertor

S.ImagawaT.Mito Takahata, Yamada, Tamura, (NIFS)

Cost evaluation

Safety T.Uda(NIFS)

Tan k TurbineThermo-fluid

Thermofluid system K.Yuki, H.Hashizume

(Tohoku Univ.) Heat exchanger & t bi t

T-disengager system S.Fukada, M.Nishikawa

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

Virtual Reality tool N.Mizuguchi (NIFS)

Divertor pumping

Fueling Sakamoto (NIFS)

External heating

pumping

Power supply H.Chikaraishi (NIFS)

Cost evaluationY. Kozaki(NIFS)

Permeation window system10-4

ferrite-steel permeation windowTritium leak

April 10, 2008, A.Sagara

Advanced thermofluid T.Kunugi (Kyoto Univ.)

gas turbine systemA.Shimizu (Kyusyu Univ.)

Advanced first wall T.Norimatsu (Osaka Univ.)

Divertor pumpingS.Masuzaki, Kobayashi (NIFS)

O.Kaneko, Igami(NIFS)

Closed He gas turbinecycle with three-stagecompression/expansion

-7

10-6

10-5

ritiu

m m

olar

frac

tion

in F

libe

[-]

104

105

106

Triti

um p

artia

l pre

ssur

e [P

a]

Fluid-film diffusion + permeation

T=843K, QT=190g-T/day, WFlibe=2.3m3/s

Fluid film diffusion controlling

MHD on Flibe

24 / 25Ten

T or h

Tl

Tn 550°C

450°C

30°C

10-8

10 7

Tr

101 102 103 104 105

Permeation area [m2]

103

10

Permeation controlling

Permeation window area [m 2] ( α=1, t=1mm)

Large apparatus

TNT Loop : Max.20L/min @ 600°C

Concluding remarksConcluding remarksConcluding remarksConcluding remarks1. Helical reactor is superior in steady state operation and

Reduced neutron wall loading < 2MW/m2 with long-life blanket concept, High density operation with reduced heat load on divertorHigh density operation with reduced heat load on divertor.

2. On the 3D design, a simply defined blanket shape can be compatible with the ergodized magnetic layer which is essential on both of divertor and α-heating.

3 This simple methodology on LHD type reactors can be used for design optimization3. This simple methodology on LHD-type reactors can be used for design optimization towards DEMO, which is economically comparable to Tokamak.

4. The design parameters of FFHR2m2 have been totally improved, where a new ignition regime at SDC is an innovative alternativeignition regime at SDC is an innovative alternative.

5. The large scale SC magnet system and their support posts are conceptually feasible.

6 Large R&D progress has been made on blanket and magnet materials6. Large R&D progress has been made on blanket and magnet materials. 7. The next key work is to realize the DPSS divertor concept compatible with

replacement scenario and neutronics issues on TBR and nuclear shielding for SC magnets

25 / 25

magnets.

SupplementsSupplements

26 / 25

japan-us workshop on march 16-18, 2009 at the university...

Documents