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Present activities on LHD-type Present activities on LHD-type Reactor Designs FFHRReactor Designs FFHR

Akio SAGARA, Akio SAGARA,

National Institute for Fusion Science, JapanNational Institute for Fusion Science, JapanContributed withContributed with

S. Imagawa, O. MitaraiS. Imagawa, O. Mitarai**, T. Dolan, T. Tanaka, , T. Dolan, T. Tanaka, Y. Kubota, K. Yamazaki, K. Y. Watanabe, N. Mizuguchi, Y. Kubota, K. Yamazaki, K. Y. Watanabe, N. Mizuguchi, T. Muroga, N. Noda, O. Kaneko, H. Yamada, N. Ohyabu, T. Muroga, N. Noda, O. Kaneko, H. Yamada, N. Ohyabu,

T. Uda, A. Komori, S. Sudo, and O. MotojimaT. Uda, A. Komori, S. Sudo, and O. Motojima

Japan-US Workshop on Fusion Power Plants and Related Advanced Technologies

with participation of EUJanuary 11-13, 2005 at Tokyo, JAPAN

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Presentation OutlinePresentation Outline

0

5

10

15

0 5 10 15 20

Mag

net

ic f

ield

B

ax (

T)

Coil major radius R (m)

LHDAp ~ 6

FFHR2Ap ~ 8

FFHR2m1Ap ~ 8

FFHR2m2Ap ~ 6

Direction of Compactness1. Introduction of FFHR1. Introduction of FFHR

2. New design approach2. New design approach

This work

The size is increasedThe size is increased Why ?Why ?

3. Conclusions3. Conclusions

Presented in 20th IAEA Fusion Energy Conference

1 - 6 November 2004Vilamoura, Portugal

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Reactor design collaborations in NIFS

1993〜

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.)

FFHRFFHR

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Tritium recovery systems Kyushyu Univ.: S.Fukada

Small amount of Flibe or He gas flow in the double tube are good as permeation barrier to reduce < 10Ci/day.

The most serious problem is permeation leak of ~34 kCi/day through the heat exchanger to the He loop

Pump

Flibe blanket

Tritiumstorage

Flow rate2.3m3/s

Tritium : 190 g-T/day = 1.8 MCi/day

Permeationbarrier< 10Ci/day

Double tubePlasma

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.%) Tritiumrecovery

Heat exchanger

Permeationbarrier

FFHR tritium recovery system (1GWth)

Permeation leak through the recovery system is a crucial problem

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Energy conversion systemsfor Flibe in/out temperature of 450˚C and 550˚C

Kyushyu Univ.: A.Shimizu

max ~ 37% for compression ratio of 1.5,

However, max decreases rapidly with the increase of pressure drop.

Therefore the layout of energy conversion system is a key design issue.

H.

C.� C.� C.�

I.C.� I.C.�

～

RH.� RH.�

E.� E.� E.�

Pr.C .

Rec.

Three-stage compression-expansion He-GT system was newly proposed

0 0.1 0.2 0.3 0.4 0.50

0.1

0.2

0.3

0.4

o

o

P

P

th

Relative pressure drop

Th

erm

al e

ffic

ien

cy

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innovative free surface wall* design Kyoto Univ.: T.Kunugi

*KSF wall (Kunugi-Sagara type Free surface wall)

Micro grooves are made on the first wall to use capillary force to withstand the gravity force

Numerical simulation has explored the formation of a pair of symmetrical spiral flow,

which enhances heat transfer efficiency about one order. 3.01 3.02 3.03

0

1000

2000

3000

4000

5000

6000

500

520

540

560

Time [sec]

hloca

l [W

/m2 ・

K]

Steady State

Tem

pera

ture

[C

]

Tbulk

Twall

1 m/sec

g

t = 3.016

Liq.Temp.° C

0.1 MW/m2

1 m/sec

2. 4 mm(48cell)

g

Solution domain14 cell

20 mm(98cell)

0. 2

1. 2

1. 2

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Thermofluid R&D activitiesTohoku Univ.: H.Hashizume

for enhancing heat-transfer in such high P

randtl-number fluid as Flibe “TNT loop” (Tohoku-NIFS Thermoflui

d loop) has been operated using HTS (Heat Transfer Salt, Tm= 142°C)

Results are converted into Flibe case at the same Pr=28.5 (Tin=200oC for HTS and 536oC for Flibe)

Same performance as turbulent flow is obtained at one order lower flow rate.

This is a big advantage for MHD effects and the pumping power.

QuickTimeý DzTIFF (LZW) êLí£ÉvÉçÉOÉâÉÄ

ǙDZÇÃÉsÉNÉ`ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅB

TNT loop ~ 0.1m3, < 600°C

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Bird’s eye view of TNT loop

Upper Tank

Main PumpDump Tank

Air Cooler

Test Section

Control Room

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Self-cooled Be-free Li/V blanketNIFS : T.Tanaka, T.Murogabased on R&D progress on in-situ MHD coatings and high purity V fabrication

Simple models are evaluated as alternatives for FFHR2 blanket..

Balance of TBR and the shielding performance is examined, because shielding is poor w/o Be.

TBR of Li/V is higher than 1.3 at about 50 cm with an acceptable shielding efficiency for super-conducting magnets.

0 120 cm

Super-conducting

magnetPlasma

x cm (120 - x) cm

JLF-1 (70 vol. %)+ B4C (30 vol. %)

W armor(5mm)

First wall(5mm)

JLF-1(5cm)

Radiation shieldSelf-cooled

tritium breederchannels

Structural material (~17 vol. %):V-4Cr-4Ti or JLF-1

Breeding coolant (~83 vol. %):Liquid lithium or Flibe

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Modeling to Evaluate MHD pressure drop is established for self-cooled lithium blanket.Tohoku Univ. : H.Hashizume

Three-layered wall is proposed, where the inner thin metal layer protects permeation of lithium into the crack of coated layer

Extremely good agreement between FEM and theory has been obtained

The performance required to the insulator is evaluated to be

insulator

V

10 8 10 9

10-25 m

Channel wall (V)

5 mm10m

Insulator100 mm

Inner layer(V)

B field

1.00E+03

1.00E+04

1.00E+05

1.00E+06

1.00E+07

1.00E-12 1.00E-10 1.00E-08 1.00E-06 1.00E-04 1.00E-02 1.00E+00

σ(insulator)/σ(V)

dp/d

z(Pa

/m)

10 μm (numerical)

15μm(numerical)

20μm numerical)（

25μm(numerical)

10μm(analytical)

25μm(analytical)

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Helical reactor FFHR design with R&D

JUPITER-II

R&D/LHDTNT loopUltrahigh HT

FY1993 1997 2001 2004 2007

Present R & D activities on Flibe blanket in JapanPresent R & D activities on Flibe blanket in Japan

2015

ITER-TBM Frame?Resouce?

$

$

$

Presented by A.Sagara(NIFS) , Feb.’04

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ORNLANL(2001)

1-1-B: FLiBe Thermofluid Flow Simulation2-2 : SiC System Thermomechanics

1-2-A: Coatings for MHD Reduction

1-2-B: V Alloy Capsule Irradiation2-1 : SiC Fundamental Issues, Fabrication, and Materials Supply2-3 : SiC Capsule Irradiation

1-1-A: FLiBe Handling/Tritium.ChemistryINEEL

Japan

3-1: Design-based Integration Modeling3-2: Materials Systems Modeling

UCLA

Flibe/RAF blanket, Li/V blanket, SB-He/SiC Blanket R&D in Japan-US joint project JUPITER-II(FY’01~’06)

http://jupiter2.iae.kyoto-u.ac.jp/index-j.html

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Many advantages :Many advantages : current-less current-less Steady stateSteady state no current drive powerno current drive power Intrinsic divertorIntrinsic divertor

LHD operation: 1998 ~LHD operation: 1998 ~

LHD-type D-TLHD-type D-T Reactor FFHR Reactor FFHR

Selection of lowerSelection of lower

To reduce mag. foop forceTo reduce mag. foop forceTo expand blanket spaceTo expand blanket space

1993 1993 FFHR-1FFHR-1 ( (l=3, m=18 l=3, m=18 ) ) R=20, Bt=12T, R=20, Bt=12T, =0.=0.7%7%

1995 1995 FFHR-2FFHR-2 ( (l=2, m=10 l=2, m=10 )) R=10, Bt=10T, R=10, Bt=10T, =1.=1.8%8%

=tan

-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

> /

(B0

I 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

-2

-1

0

1

2

2 3 4 5 6

Z (

m)

R (m)

LHD Rax=3.6m, Bq=100%,

m

l

ac

R

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However, direction of compact However, direction of compact design has engineering issuesdesign has engineering issues

Insufficient tritium breeding ratio Insufficient tritium breeding ratio (TBR) (TBR)

Insufficient nuclear shielding for sInsufficient nuclear shielding for superconducting (SC) magnets, uperconducting (SC) magnets,

Replacement of blanket due to higReplacement of blanket due to high neutron wall loading h neutron wall loading

Narrowed maintenance ports due Narrowed maintenance ports due to the support structure for high mto the support structure for high magnetic field agnetic field

Blanket spaceBlanket spacelimitationlimitation

ReplacementReplacementdifficultydifficulty

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New design approach is proposed New design approach is proposed to overcome all these issuesto overcome all these issues

Introducing Introducing a long–life & a long–life & thickerthicker breeder blanket breeder blanket

Increasing the reactor sizeIncreasing the reactor size with decreasing the magnetic with decreasing the magnetic fieldfield

Improving Improving the coils-support the coils-support structurestructure

Blanket Blanket spacespace

ReplacementReplacement

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Lifetime of Lifetime of Flibe/RAFS liquidFlibe/RAFS liquid b blanket in FFHR lanket in FFHR ~ 15MWa/m~ 15MWa/m22

Neutron wall loading Neutron wall loading in FFHR2 in 30 yearsin FFHR2 in 30 years1.5MW/m1.5MW/m22 x 30y x 30y = 45MWa/m= 45MWa/m22

factor factor 33

Proposal and Optimization of Proposal and Optimization of STBSTB((SSpectral-shifter and pectral-shifter and TTritium breeder ritium breeder BBlanket)lanket)

CarbonCarbon First wallFirst wallBreederBreeder

FastFastneutronneutron

ISSEC : by Kulchinski, ‘75

CarbonCarbon First wallFirst wallBreederBreeder

FastFastneutronneutron

STB & optimization

BeBe

TBRthis workthis work

Neutron wall loading MWa/m20 2 4 6 8 10 12 14 16 18 20

0

200

400

600

800

1000

1200

1400

Tem

pera

ture

°C

DBTT increase by irradiation

He effect (?)

Void swelling (>1%)

Thermal creep (1% creep strainunder applied stress sy/3 )

He effect (?)

Reduced Activation Ferritic Steel

operation

Plasma

107 cm0

Self-cooled T breeder

Radiation shield Vacuum vessel

SCmagnet

First wall

550Ccoolant out

LCFS

Thermal shield 450C

20C

53

T boundary&

Flibe6Li

40%

217

50.3

&

JLF-1 + B4C

(30 vol.%)

Be (60 vol.%)

167.7

Be2C

C

STB

1,600C

Tiles mechanicaljoint

10

(Redox)

Thermalinsulation

10

JLF-1JLF-1 JLF-1

C

14

Be

1

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Shielding efficiencyShielding efficiencyTritium breedingTritium breeding

1.0

1.1

1.2

0 20 40 60 80 100 120

Loca

l TB

R

Thickness of Be2C (mm)

2nd C =80 mm

120 mm

100 mm

STB

design point

C10

CBe2C breeder

& shieldLi-6 : 40%

First wall10 mm

0.9

1

1.1

1.2

1.3

1.4

1.5

1

2

3

4

5

6

7

70 80 90 100 110 120 130Fas

t N

eutr

on F

lux

( x

101

8 /m

2s)

Fast N

eutron Flux ( x 10

13

/m2s)

Thickness of 2nd carbon (mm)

at the SC magnet

at the first wall (4.2 w/o STB)

Be2C=70mm

design point

1016

1017

1018

1019

1020

10-710-610-510-410-310-210-1 100 101 102

Original FFHR2 designFFHR2 with STB

Neu

tron

flu

x (n

/m2s/

leth

argy

)

Neutron energy (MeV)

0.8

1

1.2

1.4

1.6

1.8

0 5 10 15 20

Loca

l TB

R

Thickness of the first wall JLF-1 (mm)

STB

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Results and Key R&D IssuesResults and Key R&D Issues

Plasma

107 cm0

Self-cooled T breeder

Radiation shield Vacuum vessel

SCmagnet

First wall

550Ccoolant out

LCFS

Thermal shield 450C

20C

53

T boundary&

Flibe6Li

40%

217

50.3

&

JLF-1 + B4C

(30 vol.%)

Be (60 vol.%)

167.7

Be2C

C

STB

1,600C

Tiles mechanicaljoint

10

(Redox)

Thermalinsulation

10

JLF-1JLF-1 JLF-1

C

14

Be

1

Results :Results : Fast neutron flux at the first waFast neutron flux at the first wa

ll is factor 3 reduced.ll is factor 3 reduced. Local TBR > 1.2 is possible.Local TBR > 1.2 is possible. Fast neutron fluence to SC is rFast neutron fluence to SC is r

educed to 5x10educed to 5x102222n/mn/m2 2 (Tc/Tco(Tc/Tco>90% in 30y)>90% in 30y)

Surface temperature < 2000°CSurface temperature < 2000°C(~mPa of C) is possible (~mPa of C) is possible

Required conditions :Required conditions : Neutron wall loading Neutron wall loading < 1.5MW/m< 1.5MW/m22

Blanket thicknessBlanket thickness > 1100 mm> 1100 mm for C-Be-C tilesfor C-Be-C tiles > 100W/mK> 100W/mK Super-G sheet for tiles joint Super-G sheet for tiles joint > 6kW/m> 6kW/m22KK Heat removal by Flibe Heat removal by Flibe > 1MW/m> 1MW/m22

Key R&D issues :Key R&D issues :(1)(1) Impurity shielding in edge plImpurity shielding in edge pl

asmaasma(2)(2) Neutron irradiation effects oNeutron irradiation effects o

n tiles at high temperaturen tiles at high temperature(3)(3) Heat transfer enhancement Heat transfer enhancement

in Flibe flowin Flibe flow

Design Design parameterparameter

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Design parameters LHD FFHR2 FFHR2m1 FFHR2m2

Polarity l 2 2 2 2Field periods m 10 10 10 10Coil pitch parameter 1.25 1.15 1.15 1.25Coil major Radius Rc m 3.9 10 14.0 17.3Coil minor radius ac m 0.98 2.3 3.22 4.33Plasma major radius Rp m 3.75 10 14.0 16.0Plasma radius ap m 0.61 1.2 1.73 2.80Blanket space m 0.12 0.7 1.2 1.1Magnetic field B0 T 4 10 6.18 4.43Max. field on coils Bmax T 9.2 13 13.3 13Coil current density j MA/m2 53 25 26.6 32.8Weight of support ton 400 2880 3020 3210Magnetic energy GJ 1.64 147 120 142Fusion power PF GW 1.77 1.9 3.0Neutron wall load MW/m2 2.8 1.5 1.3External heating power Pext MW 100 80 100 heating efficiency 0.7 0.9 0.9Density lim.improvement 1 1.5 1.5H factor of ISS95 2.53 1.92 1.68Effective ion charge Zeff 1.32 1.34 1.35Electron density ne(0) 10^19 m-3 28.0 26.7 19.0Temperature Ti(0) keV 27 15.8 16.1 <>=2*n*T/(B^2/) (parabolic distribution) 1.8 3.0 4.1

COE Yen/kWh 21.00 14.00 9.00

ISS95 0.26P 0.59n e0.51B0.83R0.65a2.212/ 3

0.4

ImprovedImprovedinputinput

RequiredRequiredBlanket Blanket spacespace

RequiredRequiredWall loadingWall loading

Improved design parametersImproved design parameters

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Self-ignition access in Self-ignition access in FFHR2m1FFHR2m1

Zero-dimensional analysisZero-dimensional analysis H H ISS95ISS95 = 1.2 x 1.6 = 1.2 x 1.6 **

/ / E E = 3 ( < 7)= 3 ( < 7) parabolic profilesparabolic profiles nnee< 1.5 x Sudo limit< 1.5 x Sudo limit heating efficiency = 0.9heating efficiency = 0.9

O.Mitarai et al., Fusion Eng. Design 70 (‘04) 247.

Already achieved in LHD

0

1

2

3

4

5

6

7

8

9

10

0

1

2

3

4

5

0 50 100 150 200

Pe

xt ,

SD

T

ne , Te , PF ,

Time (s)

PEXT

(10MW)

SDT

(1019 /m3s)

ne (1020 /m3)

Te (10keV)

PF (GW)

FFHR2m1

(%)

0

1

2

3

4

0 5 10 15 20 25 30

Den

sity

(x

102

0 m- 3

)

Temperature (keV)

Pext

=

10MW

20MW

30MW

40MW

50MW

ne limit (P

ext=0)

=5 %

4 %

3 %

2 %P

f=0.5GW

1GW

2GW

3GW

Operation path

FFHR2m1

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Self-ignition access in Self-ignition access in FFHR2m2FFHR2m2

Zero-dimensional analysisZero-dimensional analysis H H ISS95ISS95 = 1.1 x 1.6 = 1.1 x 1.6 **

/ / E E = 3 ( < 7 )= 3 ( < 7 ) parabolic profilesparabolic profiles nnee< 1.5 x Sudo limit< 1.5 x Sudo limit heating efficiency = 0.9heating efficiency = 0.9

O.Mitarai et al., Fusion Eng. Design 70 (‘04) 247.

Already achieved in LHD

0

1

2

3

4

5

6

7

8

9

10

11

0

1

2

3

4

5

0 50 100 150 200

Pe

xt ,

SD

T

ne , Te , PF ,

Time (s)

PEXT

(10MW)

SDT

(1019 /m3s)

ne (1020 /m3)

Te (10keV)

PF (GW)

FFHR2m2

(%)

0

1

2

3

4

0 5 10 15 20 25 30

Den

sity

(x

102

0 m- 3

)

Temperature (keV)

Pext

=

10MW

20MW

30MW

40MW50MW

ne limit (P

ext=0)

=5 %4 %

Pf=0.5GW

3GW

Operation path

FFHR2m2

3 %

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Improved Design of Improved Design of Coil-supporting Structure (1/2)Coil-supporting Structure (1/2)

Cylindrical supporting structure Cylindrical supporting structure under reduced magnetic force, under reduced magnetic force, which facilitates expansion of the which facilitates expansion of the maintenance ports.maintenance ports.

Helical coils supported at inner, Helical coils supported at inner, outer and bottom only.outer and bottom only.

W/H = 2 & H/aW/H = 2 & H/acc determined by determined by Bmax ~ 13 T for such as NbBmax ~ 13 T for such as Nb33Sn oSn o

r Nbr Nb33Al.Al. Then J=25~35A/mmThen J=25~35A/mm22

Poloidal coils layout as for stored Poloidal coils layout as for stored magnetic energy and stray fieldmagnetic energy and stray field

design point

118

119

120

121

17.1

17.2

17.3

17.4

3.2 3.3 3.4 3.5 3.6 3.7 3.8

Sto

red

Ene

rgy

(GJ)

Radius of O

V C

oil (m)

Z of IV Coil (m)

RIV=9.5m, ZOV=3.6 m

Sagara 26

Improved Design of Improved Design of Coil-supporting Structure (2/2)Coil-supporting Structure (2/2)

Coil current (MA)HC 43.257OV -21.725IV -22.100

The maximum stress can be reThe maximum stress can be reduced less than 1000 MPa (<duced less than 1000 MPa (<1.5Sm)1.5Sm)

This value is allowable for streThis value is allowable for strengthened stainless steel.ngthened stainless steel.

Electromagnetic forces on Electromagnetic forces on a helical coil of FFHR2m1.a helical coil of FFHR2m1.

-50

0

50

100

0 12 24 36 48 60 72

Fa (minor-radius)Fb (overturning)

Ele

ctro

mag

netic

For

ce (

MN

/m)

Toroidal Angle, Phi (deg)

outer top inner bottom outer

-40

-20

0

20

40

60

80

0 12 24 36 48 60 72

FZ of OVFR of OV

FZ of IVFR of IV

Ele

ctro

mag

netic

For

ce (

MN

/m)

Toroidal Angle, Phi (deg)

on poloidal coilson poloidal coils

FFHR2m1

Sagara 27

Replacement of In-Vessel Replacement of In-Vessel Components (1/2)Components (1/2)

Large size maintenance portsLarge size maintenance ports at t at top, bottom, outer and inner sides.op, bottom, outer and inner sides.

The vacuum boundary locatedThe vacuum boundary located just just inside of the helical coils and supinside of the helical coils and supporting structure. porting structure.

Blanket units supported on the peBlanket units supported on the permanent shielding structuresrmanent shielding structures, whic, which are mainly supported at their helh are mainly supported at their helical bottom position. ical bottom position.

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Proposal of the “Screw coaster” concept Proposal of the “Screw coaster” concept to replace STB armor tilesto replace STB armor tiles

Replacement of In-Vessel Replacement of In-Vessel Components (2/2)Components (2/2)

Replacement of bolted tilesReplacement of bolted tiles du during the planned inspection perring the planned inspection period.iod.

Using the merit of helical strucUsing the merit of helical structure,ture, where the normal cross s where the normal cross section of blanket is constant.ection of blanket is constant.

Toroidal effects can be adjusteToroidal effects can be adjustedd with flexible actuators. with flexible actuators.

Sagara 29

Cost EstimationCost EstimationUsing PEC code developed in NIFS.Using PEC code developed in NIFS.

Calibrated with ARIES-AT, ARIES-SPPS, resulting in good agreement within 5 %. Calibrated with ARIES-AT, ARIES-SPPS, resulting in good agreement within 5 %. (T.J.D(T.J.Dolan,K.Yamazaki, A.Sagara, in press in olan,K.Yamazaki, A.Sagara, in press in Fusion Science & Tech.)Fusion Science & Tech.)

0

5

10

15

20

25

30

35

10 12 14 16 18 20 22 24

Major radius Rp, m

CO

E, Y

en/k

Wh

beta=1%

2%

4%

6%

=1.15Rp/<a>=8.09=1.2 m

Pf = 1 GW

2 GW

4 GW

7 GWFFHR2m1

HISS=2

HISS=1.5

0

5

10

15

20

25

30

35

40

14 16 18 20 22 24

Major radius Rp, mC

OE

, Yen

/kW

h

beta=1%

beta=2%

beta=4%

beta=7%

HISS=2 HISS=1.5=1.25=1.1 mRp/<a>=5.71

FFHR2m2

Pf=1 GW

2 GW

4 GW

7 GW

6

7

8

9

10

11

0 5 10 15 20 25 30 35 40

Blanket Lifetime, years

CO

E, Y

en/k

Wh

favail = 0.85 - favail *1.4 w *tm / Wtb

COE’s for FFHR2, FFHR2m1 and FFHR2m2 COE’s for FFHR2, FFHR2m1 and FFHR2m2 decreases with increasing the reactor sizdecreases with increasing the reactor size,e, because the fusion output increases in ~ Rbecause the fusion output increases in ~ R22, wh, while the weight of coil supporting structure increases ile the weight of coil supporting structure increases in~ R in~ R 0.40.4 ( not ~R ( not ~R33). ).

When the blanket lifetime ~ 30y, the COE When the blanket lifetime ~ 30y, the COE decreases ~20%decreases ~20% due to higher availability and ldue to higher availability and lower cost for replacement.ower cost for replacement.

0

5

10

15

20

25

30

35

8 10 12 14 16 18

Rp, m

CO

E, Y

en

/kW

h

beta=1%

2%

3%

4%

Pf=1 GW

2 GW

4 GW

7 GW

Hiss=2

Hiss=1.5

1

= 1.15Rp/<a>=8.33 = 0.7 m

FFHR2

Sagara 30

Design studies on FFHR have focused on new design apDesign studies on FFHR have focused on new design approaches to solve the key engineering issues of blanket proaches to solve the key engineering issues of blanket space limitation and replacement difficulty. space limitation and replacement difficulty.

(1)(1) The combination of improved support structure and The combination of improved support structure and long–life breeder blanket STB is quite successful. long–life breeder blanket STB is quite successful.

(2)(2) The “screw coaster” concept is advantageous in heThe “screw coaster” concept is advantageous in heliotron reactors to replace in-vessel components. liotron reactors to replace in-vessel components.

(3)(3) The COE can be largely reduced by those improved The COE can be largely reduced by those improved designs. designs.

(4)(4) The key R&D issues to develop the STB concept arThe key R&D issues to develop the STB concept are elucidated.e elucidated.

ConclusionsConclusions