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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)..
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