ITC17 & ISHW2007
Oct.15-19, 2007 at Toki, Japan.

DesignDesign Integration
Integration toward
toward Optimization
Optimization ofofLHD-type Fusion
LHD-typeFusion Reactor
Reactor FFHR
FFHR
Akio Sagara, O. Mitarai1, S. Iamagawa, Y. Kozaki, T. Watanabe, T. Tanaka,
N. Yanagi, K. Takahata, H. Tamura, H. Chikaraishi, S. Yamada, T. Mito,A. Shishkin2, Y. Igitkhanov3, H. Igami, T. Morisaki, M. Kobayashi,
R. Sakamoto, H. Yamada, O. Motojima and FFHR design group
National Institute for Fusion Science, Japan
1Kyushyu Tokai University, Japan
2Kharkov Institute of Physics and Technology, Ukraine
3Max-Planck-Institut fur Plasmaphysik, Germany

FFHR
2 GWth
6 Tesla
25,000 ton
1 / 22

FFHR design collaborations
Ignition access& heat flux
Thermo-mechanical
FFHR design
O.Mitarai
analysis
Blanket system/ Sytem Integration(Kyusyu Tokai
H.Matsui (Tohoku
T.Terai (Univ. of Tokyo)/ Replacement
Helical core plasma
Univ.)
Univ.)
A.Sagara (NIFS)
K.Yamazaki(NIFS)
Neutronics
T.Tanaka, Sagara
Structural Materials
Magnetic structure(NIFS)
T.Morisaki, Yanagi
Blanket(NIFS)
Alternative blanketT.Muroga
SC magnet
14MeV neutron(NIFS)
SC magnet
Be Liquidshielding/ Gas& supprt
materials.
& support
Safety20°C
S.ImagawaCore surface heat flux
high temp.
SafetyT. UdaT.MitoT-disengager
Plasma&Cost(NIFS)
Takahata,
Protection wall
Ph =0.2 MW/m 2
Pn =1.5 MW/m 2
Nd =450 dpa/30y

Radiation shield
reduction> 5 orders
Thermal shieldSelf-cooled
T breeder
TBRlocal > 1.2

Thermo-mechanical analysis
Optimum first wall thickness
604020Vacuum vessel&T boundarySCmagnet

Yamada,Tamura,(NIFS)Virtual Reality tool
N.Mizuguchi(NIFS)
Power supply
H.Chikaraishi(NIFS)
External heating
O.Kaneko, Igami(NIFS)

ChemistryTritiumPump

HX

Heating Divertor
pumping

Fueling
Sakamoto(NIFS)
Divertor pumping
S.Masuzaki,
Kobayashi (NIFS)

Fueling

Pump

T storage

Purifier

Thermofluid

Thermofluid MHD
S.Satake(Tokyo Univ. Sci.)
Advanced first wall
T.Norimatsu(Osaka Univ.)

MHD on Flibe

Tank

TurbineThermofluid system
K.Yuki, H.Hashizume(Tohoku Univ.)

0010

Permeation window system-4

Tritium leak
ferrite-steel permeation window
3T=843K, QT=190g-T/day, WFlibe=2.3m /s

10

Cost evaluation
Y. Kozaki(NIFS)
T-disengager system
S.Fukada, M.Nishikawa(Kyusyu Univ.)
Heat exchanger &gas turbine system
A.Shimizu(Kyusyu Univ.)

610-5

10

Fluid film diffusion controlling
Fluid-film diffusion + permeation

5-6

10-7

10104

10

Permeation controlling
3Large apparatus
10-8

10

1

10

Advanced thermofluid
T.Kunugi(Kyoto Univ.)

20Thickness, mmTritium molar fraction in Flibe [-]In-vessel
Components

Stress by internalpressure, 0.5MPa
Stress by internalpressure, 1MPa
Stress by internalpressure, 1.5MPa
Thermal stress

Tritium partial pressure [Pa]Fusion Engineering

Confinement scaling
H.Yamada,Miyazawa (NIFS)

Stress / MPa

LHD2310104105102 ]2( α =1, t=1mm)
Permeation window
area [m area [mPermeation]

Closed He gas turbine
cycle with three-stagecompression/expansion
March 10. 2007, A.Sagara
T or hTn

550°C

Tl

450°C

TNT Loop : Max.20L/min @ 600°C

T en

30°C

2 / 22

Presentation outline
toward design optimization

1. Reactor size and cost
2. Minimization of heating power
3. High density ignition and control
4. Large SC magnet system
5. Access to Demo
6. Concluding remarks
2233 // 22

Two main features in FFHR
(1) Quasi-force free γ optimizationon continuous helical windingJ x Bsurf = 0, when ϑ = 45 deg.

¾

to reduce the magnetic hoop force(Force Free Helical Reactor: FFHR)¾ to expand the blanket spaceγ =⎛ m ac ⎞⎝ l R⎠

2Rax =3.6m, B q=100%, Φ = 0°LHD

1Z (m)

γ=tanθ

γ=1.33γ=1.180(2) Self-cooled liquid Flibe(BeF2-LiF) blanket-1-2234R (m)56¾¾¾¾

low MHD pressure loss
low reactivity with air
low pressure operation
low tritium solubility

444 / 2222

LongLong life
life blanket
blanketandand wide
wide maintenancemaintenance ports
ports
1. Simplified support structurewith wide maintenance ports
due to FF concept.
2. Neutron wall loading ~ 1.5 MW/m2 is a
key for long life or replacement freeblanket.

STB (Spectral-shifter and Tritium
Neutron damage (dpa/30years)

breeder Blanket)
A.Sagara, Nuclear Fusion 45 (2005) 258250FFHR2m1

STB in FFHR2m1: after 30 years at 1.5 MW/m2200

MaintenancePort
Nuclear Shield

150

VacuumVessel

BreederBlanket

Poloidal
Coil

Flibe40%+Be60%

100
C50

Be2C

JLF-1

JLF-1

CFlibe00100
200
Position (mm)

300

400

Coil Support

Screw Coaster
Helical Coil

5 / 220Magnetic field B (T)151998FFHR2Ap ~ 8γ =1.15102004
Ap~4.5
FFHR2m1
ARIES-CS

ITER

5Ap~3.1

LHD
Ap ~ 600

Ap ~ 8γ =1.15
outer shift

1995
FFHR1(l=3)
Ap ~ 10
γ =1

optimum ?

Reactor
Reactor size
sizeoptimization
w/o
optimization w/oincreasing
increasing neutron
neutronwallwall loading
loading inin FFHR
FFHR

FFHR2m2
Ap ~ 6γ =1.25
inner shift

510152025
Coil major radius Rc (m)Ergodic
Ergodiclayer
layerisisimportant
importantfor
foralpha-heatingalpha-heatingefficiency
efficiency>>90
90%,%,

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

by T. Watanabe

6 / 22

ThreeThree candidates
candidates areare proposed
proposed
toto increase
increase blanket
blanket space
space >> 1.1
1.1 mm(1)Reduction of the inboard
shielding thickness
using WC
1015
1014

1011
1010

(Magnet)

1013
1012

9

Breedinglayer

108[Flibe+Be]

10107

by increasing the current density
at the inboard side of the helical coils

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

PlasmaFast neutron flux （>0.1MeV)(n/cm2/s)

1016

(2)
Improvement of the symmetryof magnetic surfaces(32cm)

Radiation
shield(60 cm)

180 200 220 240 260 280 300
Radial position（cm)

LHD

FFHR-2Sby splitting the helical coils.

N. Yanagi et al., in this conference.
7 / 22Three
Three candidates
candidates areare proposed
proposed
toto increase
increase blanket
blanket space
space >> 1.1
1.1 m
m
(3)
(3)Optimization
Optimizationof
ofreactor
reactorsize
sizearoundaroundRc
Rcof
of1616mm

FFHR2 FFHR2m1 FFHR2m2

4.6
155

2
10
1.15
14.0
3.22
14.0
1.73
827
1.1
6.18
13.3
26.6
133
1.9
1.5
80
0.9
1.5
1.92
1.34
26.7
15.8
3.0
290
1.6
5.6
106

2
10
1.25
17.3
4.33
16.0
2.80
2471
1.15
4.43
13.0
32.8
118
3
1.3
100
0.9
1.5
1.76
1.3519.0
16.1
4.1463
2.3
6.987(Cost evaluation based on the ITER (2002) report)

15200
J = 25 ~ 33 A/mm2
Γn = 1.3 ~ 1.5 MW/m2

Pf (GW), TCC (G$), Bo (T)

210
1.15
10
2.3
10
1.24
303
0.7
10
14.8
25
147
1
1.5
70
0.7
1
2.40
1.40
27.4
21
1.6

blanket space > 1 m

Wg (magnetic energy)

150

10COE

B0100

550

TCC (total capital cost)
FFHR2m2

Pf (fusion output)
0FFHR2510Wg (GJ), COE (mil/kWh)Design parameters
LHDPolarityl2
Field periods
m10
Coil pitch parameterγ1.25Rc
m 3.9
Coil major Radius
acm 0.98
Coil minor radius
Rpm 3.75
Plasma major radius
ap
Plasma radius
m 0.6130
Plasma volume
Vp
m3Blanket space
Δm 0.12B0T
4Magnetic field
Bmax
T 9.2
Max. field on coils
53
Coil current density
j MA/m2
Magnetic energy
GJ 1.64
PFGW
Fusion power
Γn MW/m2
Neutron wall load
MW
External heating power Pextηαα heating efficiency
Density lim.improvement
H factor of ISS95
ZeffEffective ion charge-3nElectron densitye(0) 10^19 mTi(0)TemperaturekeV
Plasma beta
<β>%PMW
Plasma conduction loss L
Γdiv MW/m2
Diverter heat load
Total capital cost
G$(2003)COE
mill/kWhFFHR2m1

15020

Coil major radius Rc (m)8 / 22ITER
ITERbasedbasedcostcostanalyses
analysesshow
show

Similar
Similarweights
weightsand
andcosts
costsbetween
betweenHelical
HelicalandandTokamak
Tokamakreactor
reactormagnet
magnetsystems
systems(FFHR2m1)(FFHR2m1)
Strands
Coil Case

Weight ton
05000

10000

01500016,000 ton

Poloidal CoilSupps. & Others

ITER TF

ITER TF

ITER CSITER CS
10,200 ton

ITER PFSupps. & Others

Supps. & Others

FDR TF

FDR TF

FDR CS

FDR CS

FDR PF
Supps. & Others

50100150200

Helical CoilSupps. & Others

ITER PF

Radial Plate

Cost B Yen

Helical Coil
Poloidal CoilConduit
Winding

21,900 ton

FDR PF
Supps. & Others

210 BYen

110 BYen

275 BYen

9 / 22

Presentation outline
toward design optimization

1. Reactor size and cost
2. Minimization of heating power
3. High density ignition and control
4. Large SC magnet system
5. Access to Demo
6. Concluding remarks
1010/ /2222

2 1020NE0NE0LMT40(a)30201 1020

T

FALPHA

1nlimit(0)/n(0)04DLM

WPV(c)

2400

10TAUE

3E

Reduction ofthermal stress

0800

04τ (s)

10(b)

0.02

2

rise-up time 2 h

10150
PEXT(MW)+ (PL + PB + PS )02PFPF03dt

Wp(MJ)

falpha00.04

PEX T + Pα =Ti(0) (keV)

3 10

20O. Mitarai al. Nucl. Fusion 47 (2007) 1411.

Pf (GW)

4 10

2015PEXT

100(d)

DWDTV

550
010dW/dt(MW)n(0) (m-3)

MinimizedMW
Minimized heatingheating power
power ~~ 3030MWdW001000

2000

3000

4000

5000

Time (s)

6000

7000

8000

11 / 22IDB
IDB Ignition
Ignition Scenario
Scenario withwith SDC
SDC
Conventional design

IDB scenarion, T parabola profile: (1-x2)

Design parameters

FFHR2m1

Fusion powerPF
Density lim.improvement
H factor of ISS95

SDC

1.9

GW
1.5

7.5

IDB profile

1.92
-3

2.4
15.8
1.48
3.0
1.9
57
282

9.8
6.27
1.52
2.5
4.7
248
96

Electron density
ne (0) 10^20 mkeV
TemperatureTi(0)Effective ion charge
Zeff<β>Plasma beta%τΕEnergy confinem. timesMW
Bremsstrahlung loss
PBMW
Plasma conduction loss P L
Heat load to first wall

Γ div

MW/m

20.06

0.25

Heat load to divertor

Γ div

MW/m

2

1.6

0.54

Divertor heat load is
drastically reduced.12 / 22

NewNew control
control method
methodinin aa thermallythermally unstable
unstable regime
regimeProportional-Integration-Derivative (PID) control
The error of the fusion power with an opposite
sign of e(Pf)= - (Pfo-Pf) can stabilize the thermal
instability through fueling.

SDT(t)=0if SDT(t)<0O. Mitarai al. Plasma and Fusion Research,Rapid Communications, 2, 021 (2007) .
τα*/τE=3 ~ 5
τp*/τE=2 ~ 8
But,effectively
reduced dueto burning
13 / 22

22
Neutronwall
loading
isave.
1.5
MW/m
Neutron wall loading is ave. 1.5 MW/mandand peaking
peaking factor
factor << 1.3
1.3

Cell 2

Quick feedback
Cell 1
between 3D CAD
and 3D Monte-Carlocode MCNP5,
using numerical
Q uadrangu lar m esh
Quadrangularpreparedin Fig. 2mesh(a) helical equations.

Two cases of neutron sources
0.14

Source distribution (a.u.)

Evaluated using the recently developed 3D
neutronics calculation system for nonaxisymmetric helical systems
T. Tanaka et al. 21th IAEA (2006)

Helical source
Torus uniform source

0.120.10

Ion density: n∝(1-r2)
Temperature: T∝(1-r2)
Max. Temp.: 16 keV

0.08
0.06

p（ｒ） ∝ n2・<σv>(T)

0.04
0.02
0.00

0

prepared in Fig. (a).

Total TBR
> 1.1

50

100

150

200

250

Position (cm)First wall of blanket
on helical coilMax. 1.8
for helicalsource
Max. 2.0
for uniformsource
Average 1.5First wall of blanket
at divertor area0(unit : MW/m2)

14 / 22

Presentation outline
toward design optimization

1. Reactor size and cost
2. Minimization of heating power
3. High density ignition and control
4. Large SC magnet system
5. Access to Demo
6. Concluding remarks
1515/ /2222

Base
Base design
design ofofCICC
CICC magnet
magnet systemsystem
S. Imagawa et al., in this conference.
101

3

Nuclear heating (mW/cm )Nuclear heating in FFHR2m1

10100-1

Neutron+GammaGamma

5 times

10-2
10-3

10-4-20

0 20 40 60 80 100
Radial position (cm)

9 Max. cooling path is 500 m forthe nuclear heat of 1 mW/cm3.
9 This value is 5 times larger on
the FFHR magnets.
9 Gamma-ray heating is dominant.
16 / 22

Nuclear
Nuclear shielding
shielding ofof SC
SC magnets
magnets
bybyDiscrete
DiscretePumping
Pumpingwith
withSemi-closed
Semi-closedShield
Shield (DPSS)(DPSS)

cover
cover rate
rate >> 90%
90%
Acceptable level achieved
9 Fast neutron < 1E22 n/m2 in 30 years
9 Max. nuclear heating < 0.2 mW/cm3
9 Total nuclear heating ~ 40 kW
Cryogenics power ~ 12 MW (1% of Pf)
3D design withDPSS

17 / 22

Indirect
Indirect cooling
cooling systemsystemasasan
analternative
alternativewith
withquench
quenchprotection
protectioncandidates
candidates

K. Takahata et al., 24th SOFT, 2006.
SS316
SS316

Nb3Sn+Cu
Indirect Cooling
Ceramic

100 kA Superconductor¾ High effective thermal conductivity¾ High mechanical rigidity and strength

Quench protection
by external dumping¾ Conventional protection circuit
using an external resistor τ=20 s¾ Six subdivisions Vmax=10 kV¾ Hot spot temperature < 150 K

Cross-sectional structure of the helical coil

Quench protection by internal dumping

K. Takahata et al., in this conference.
¾ Quench back with a secondary circuit¾ to increase a decay time constant¾ to reduce a transient voltage¾ to avoid a serious hot spot

18 / 22

LHD-type support
LHD-type
support post
post forfor FFHR
FFHR
H. Tamura et al., in this conference.

(Carbon Fiber
Reinforcement Plastic)

5.663¾ Gravity per support= 16,000 ton / 30 legs ~ 530 ton.
¾ Thermal contraction < max. 55 mm¾ Total heat load to 4K ~ 0.34 kW( 1/20 of stainless steel post)

18
6.

19 / 22

Design parameters
Polarity
Field periods
Coil pitch parameter
Coil major Radius
Coil minor radius
Plasma major radius
Plasma radius
Plasma volumeBlanket space
Magnetic field
Max. field on coils
Coil current density
Magnetic energy
Fusion power
Neutron wall load
External heating power
α heating efficiency
Density lim.improvement
H factor of ISS95lm
γ
Rcac
Rp
apmmmm

Vp
Δmm
T
T

B0
B maxjPFΓnP ext
ηα32MA/m
GJGWLHDFFHR2m12
10
1.25
3.90.98
3.75
0.61

210
1.15
14.0
3.22
14.0
1.73

30
0.12
4
9.2

827
1.1
6.18
13.3

External heating

53
1.64

26.6
133
1.9

High energy particle

2MW/m
MWSDC

LHD experiments

1.580
0.9
1.5

New density limit
7.5

1.92-3

Electron density
Temperature
Effective ion charge
Plasma beta
Energy confinem. time
α ash confimentne (0)
Ti(0)
Zeff
<β>
τΕτ α ∗ /τ Ε

10^20 mkeV

Bremsstrahlung loss
Plasma conduction loss

PBPLMW
MW

Heat load to first wall

Γ div

Heat load to divertorDiverter heat load
Total capital cost
COE%
s

2.4
15.8
1.48
3.0
1.9

9.8
6.27
1.52
2.5
4.7

Enhancement of τE
Impurity shielding

3<757282248
96

MW/m

20.06

0.25

Γ div

2MW/m

1.6

0.54

Γ div

2

1.6
5.6
106

0.5

MW/m
G$(2003)mill/kWhHe exhaust

20 // 222220

Role-Reactor
RoleofofDesign
DesignStudy
StudytotoHelical
HelicalDemo
Demo-Reactor
based
basedononLHD
LHDProject
Project
Tokamak Experimental Rector
ITER

LHD-type Helical
Reactor FFHR
Electric Power 1GW
Weight25,000ton
Magnetic Field6T

Helical Demo Rector(29 years to go)

Physics ofburning plasmas

LHD

Demonstration ofsteady-state, highdensity, high beta by
net-current free plasma

Multi-layer models covering
physics and engineering
LHD-NT
LHD Numerical
Test Reactor

Basic Science
21 / 22

Concluding
Concluding remarks
remarks
1. Helical reactor is superior in steady state operation and¾ Reduced neutron wall loading by optimization oflarge reactor size,¾ Minimized heating power by long access time toignition,¾ High density operation with reduced heat load ondivertor.
2. Large SC magnet system is conceptually feasible.
3. Helical reactor is economically comparable to Tokamak.
4. Numerical Test Reactor is planned to Helical Demo.
5. Large sized construction is important R&D issue.
6. LHD experiments can open new reactor regimes.
22 / 22