present activities on lhd-type reactor designs...
TRANSCRIPT
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Present activities on LHDPresent activities on LHD--type type Reactor Designs FFHRReactor Designs FFHR
Akio SAGARA, Akio SAGARA,
National Institute for Fusion Science, JapanNational Institute for Fusion Science, JapanContributed withContributed with
S. S. ImagawaImagawa, O. , O. MitaraiMitarai**, T. Dolan, T. Tanaka, , T. Dolan, T. Tanaka, Y. Kubota, K. Yamazaki, K. Y. Watanabe, N. Y. Kubota, K. Yamazaki, K. Y. Watanabe, N. MizuguchiMizuguchi, , T. T. MurogaMuroga, N. Noda, O. Kaneko, H. Yamada, N. , N. Noda, O. Kaneko, H. Yamada, N. OhyabuOhyabu, ,
T. T. UdaUda, A. Komori, S. , A. Komori, S. SudoSudo, and O. , and O. MotojimaMotojima
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
Bax
(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 increasedWhy ?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 systemsKyushyu Univ.: S.Fukada
Permeation leak through the recovery system is a crucial problem
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)
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Energy conversion systemsfor Flibe in/out temperature of 450˚C and 550˚C
Kyushyu Univ.: A.Shimizu
ηmax ~ 37% for compression ratio of 1.5,
However, ηmaxdecreases 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
PP
∆∑
thη
Relative pressure drop
The
rmal
effi
cien
cy
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innovative free surface wall* designKyoto 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 Prandtl-number fluid as Flibe“TNT loop” (Tohoku-NIFSThermofluid loop) has been operated using HTS (Heat Transfer Salt, Tm= 142°C)
Results are converted into Flibecase 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-Ç�ÅB
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
σ insulator
σV
≈10−8 −10−9
10-25 µm
Channel wall (V)
5 mm10µm
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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Present R & D activities on Present R & D activities on Flibe Flibe blanket in Japanblanket in JapanPresented by A.Sagara(NIFS) , Feb.’04
FY1993 1997 2001 2004 2007 2015
Helical reactor FFHR design
with R&D
JUPITER-II
R&D/LHDTNT loopUltrahigh HT
ITER-TBMFrame?Resouce?
$
$
$
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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.ChemistryINEELJapan
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 :currentcurrent--less less Steady stateSteady stateno current drive powerno current drive powerIntrinsic Intrinsic divertordivertor
LHD operation: 1998 ~LHD operation: 1998 ~
LHDLHD--type Dtype D--TT Reactor FFHRReactor FFHRSelection of lowerSelection of lower
To reduce To reduce magmag. . foopfoop forceforceTo expand blanket spaceTo expand blanket space
1993 1993 FFHRFFHR--11 ((l=3, m=18 l=3, m=18 ) ) R=20, Bt=12T, R=20, Bt=12T, ββ=0.7%=0.7%
1995 1995 FFHRFFHR--22 ((l=2, m=10 l=2, m=10 ))R=10, Bt=10T, R=10, Bt=10T, ββ=1.8%=1.8%
γ=tanθ
-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 l
coil cross-section
NIFS-960812 / S.I.
l
Coil pitch parameter c =(m/ )( acγ /R) l
FFHR-2
ac/H W/H < fa > (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)
γ=1.18γ=1.33
LHD Rax=3.6m, Bq=100%, Φ = 0°
γ =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 Insufficient tritium breeding ratio (TBR) ratio (TBR)
Insufficient nuclear shielding Insufficient nuclear shielding for for superconductingsuperconducting (SC) (SC) magnets, magnets,
Replacement of blanket due to Replacement of blanket due to high neutron wall loading high neutron wall loading
Narrowed maintenance ports Narrowed maintenance ports due to the support structure for due to the support structure for high magnetic field
Blanket spaceBlanket spacelimitationlimitation
ReplacementReplacementdifficultydifficulty
high magnetic field
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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 longa long––life & life & thickerthicker breeder blanketbreeder blanket
Increasing the reactor sizeIncreasing the reactor sizewith decreasing the magnetic with decreasing the magnetic fieldfield
Improving Improving the coilsthe coils--support support structure
Blanket Blanket spacespace
ReplacementReplacementstructure
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Lifetime ofLifetime of FlibeFlibe/RAFS liquid/RAFS liquidblanket in FFHR blanket 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((SSpectralpectral--shifter and 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
0200
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
SC�magnet
First wall
550�C�coolant out
LCFS
Thermal shield 450�C20�C
53
T boundary&
Flibe�6Li�40%
217
50.3
&
JLF-1� + B4C�
(30 vol.%)
Be (60 vol.%)16
7.7
Be2C
C
STB
1,600�C
Tiles mechanical�joint
10
(Redox)
Thermal�insulation
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 Neu
tron
Flu
x (
x 10
18 /m
2 s) F
ast Neutron F
lux ( x 1013 /m
2s)
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
flux
(n/m
2s/
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
SC�magnet
First wall
550�C�coolant out
LCFS
Thermal shield 450�C20�C
53
T boundary&
Flibe�6Li�
40%
217
50.3
&
JLF-1� + B4C�
(30 vol.%)
Be (60 vol.%)16
7.7
Be2C
C
STB
1,600�C
Tiles mechanical�joint
10
(Redox)
Thermal�insulation
10
JLF-1JLF-1 JLF-1
C
14
Be
1
Results :Results :Fast neutron flux at the first Fast neutron flux at the first wall is factor 3 reduced.wall is factor 3 reduced.Local TBR > 1.2 is possible.Local TBR > 1.2 is possible.Fast neutron Fast neutron fluencefluence to SC is to SC is reduced to 5x10reduced to 5x102222n/mn/m2 2
((TcTc//TcoTco>90% in 30y)>90% in 30y)Surface temperature < Surface temperature < 20002000°°C(C(~mPa~mPa of C) is possible 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
λλ for Cfor C--BeBe--C tilesC tiles> 100W/> 100W/mKmK
SuperSuper--G sheet for tiles joint G sheet for tiles joint > 6kW/m> 6kW/m22KK
Heat removal by Heat removal by Flibe Flibe > 1MW/m> 1MW/m22
Key R&D issues :Key R&D issues :(1)(1) Impurity shielding in edge Impurity shielding in edge
plasmaplasma(2)(2) Neutron irradiation effects Neutron irradiation effects
on tiles at high temperatureon tiles at high temperature(3)(3) Heat transfer enhancement Heat transfer enhancement
in in Flibe Flibe flowflow
Design Design parameterparameter
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Improved design parametersImproved design parametersDesign parameters LHD FFHR2 FFHR2m1 FFHR2m2
Polarity l 2 2 2 2Field periods m 10 10 10 10Coil pitch parameter γ 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 100α heating efficiency ηα 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 <β>=2*n*T/(B^2/ µ) (parabolic distribution) 1.8 3.0 4.1COE Yen/kWh 21.00 14.00 9.00
ImprovedImprovedinputinput
τ ISS95 = 0.26P−0.59n e0.51B0.83R0.65a2.21ι 2/ 3
0.4
RequiredRequiredBlanket Blanket spacespace
RequiredRequiredWall loadingWall loading
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SelfSelf--ignition access in ignition access in FFHR2m1FFHR2m1
ZeroZero--dimensional analysisdimensional analysisH H ISS95ISS95 = 1.2 x 1.6= 1.2 x 1.6ττ**
αα / / ττE E = 3 ( < 7)= 3 ( < 7)parabolic profilesparabolic profilesnnee< 1.5 x < 1.5 x SudoSudo limitlimitαα 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
Pex
t ,
S
DT
ne , Te , PF , β
Time (s)
PEXT
(10MW)
SDT
(1019 /m3s)
ne (1020 /m3)
Te (10keV)
PF (GW)
FFHR2m1
β (%)
0
1
2
3
4
0 5 10 15 20 25 30
Den
sity
(x
1020
m- 3
)
Temperature (keV)
Pext
=
10MW
20MW
30MW
40MW
50MW
ne limit (P
ext=0)
β=5 %
4 %
3 %
2 %P
f=0.5GW
1GW
2GW
3GW
Operation path
FFHR2m1
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SelfSelf--ignition access in ignition access in FFHR2m2FFHR2m2
ZeroZero--dimensional analysisdimensional analysisH H ISS95ISS95 = 1.1 x 1.6= 1.1 x 1.6ττ**
αα / / ττE E = 3 ( < 7 )= 3 ( < 7 )parabolic profilesparabolic profilesnnee< 1.5 x < 1.5 x SudoSudo limitlimitαα 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
Pex
t ,
S
DT
ne , Te , PF , β
Time (s)
PEXT
(10MW)
SDT
(1019 /m3s)
ne (1020 /m3)
Te (10keV)
PF (GW)
FFHR2m2
β (%)
0
1
2
3
4
0 5 10 15 20 25 30
Den
sity
(x
1020
m- 3
)
Temperature (keV)
Pext
=
10MW
20MW30MW
40MW50MW
ne limit (P
ext=0)
β=5 %4 %
Pf =0.5GW
3GW
Operation path
FFHR2m2
3 %
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Improved Design of Improved Design of CoilCoil--supporting Structure (1/2)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 bydetermined byBmaxBmax ~ 13 T for such as Nb~ 13 T for such as Nb33Sn Sn or Nbor Nb33Al.Al.Then J=25~35A/mmThen J=25~35A/mm22
Poloidal Poloidal coils layout as for coils layout as for stored magnetic energy and stored magnetic 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
stray field
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Improved Design of Improved Design of CoilCoil--supporting Structure (2/2)supporting Structure (2/2)
Coil current (MA)HC 43.257OV -21.725IV -22.100
The maximum stress can be The maximum stress can be reduced less than 1000 reduced less than 1000 MPaMPa(<1.5Sm)(<1.5Sm)
This value is allowable for This value is allowable for strengthened stainless steel.strengthened 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)
onon poloidalpoloidal coilscoils
FFHR2m1
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Replacement of InReplacement of In--Vessel Vessel Components (1/2)Components (1/2)
Large size maintenance portsLarge size maintenance ports at at top, bottom, outer and inner sides.top, bottom, outer and inner sides.The vacuum boundary locatedThe vacuum boundary locatedjust inside of the helical coils and just inside of the helical coils and supporting structure. supporting structure. Blanket units supported on the Blanket units supported on the permanent shielding structurespermanent shielding structures, , which are mainly supported at which are mainly supported at their helical bottom position. their helical bottom position.
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Proposal of the Proposal of the ““Screw coasterScrew coaster”” concept concept to replace STB armor tiles
Replacement of InReplacement of In--Vessel Vessel Components (2/2)Components (2/2)
to replace STB armor tiles
Replacement of bolted tilesReplacement of bolted tilesduring the planned inspection during the planned inspection period.period.Using the merit of helical Using the merit of helical structure,structure, where the normal where the normal cross section of blanket is cross section of blanket is constant.constant.ToroidalToroidal effects can be effects can be adjustedadjusted with flexible with flexible actuators. actuators.
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Cost EstimationCost EstimationUsing PEC code developed in NIFS.Using PEC code developed in NIFS.
Calibrated with ARIESCalibrated with ARIES--AT, ARIESAT, ARIES--SPPS, resulting in good agreement within 5 %. SPPS, resulting in good agreement within 5 %. (T.J.Dolan,K.Yamazaki, A.(T.J.Dolan,K.Yamazaki, A.SagaraSagara, in press in , 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
beta=1%
2%
4%
6%
γ=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 24Major radius Rp, m
CO
E, Y
en/k
Wh
beta=1%
beta=2%
beta=4%
beta=7%
HISS=2 HISS=1.5γ=1.25∆=1.1 mRp/<a>=5.71
FFHR2m2
Pf=1 GW
2 GW
4 GW7 GW
6
7
8
9
10
11
0 5 10 15 20 25 30 35 40
Blanket Lifetime, years
favail = 0.85 - favail *1.4 �w *tm / Wtb
COECOE’’ss for FFHR2, FFHR2m1 and for FFHR2, FFHR2m1 and FFHR2m2 decreases with increasing FFHR2m2 decreases with increasing the reactor size,the reactor size, because the fusion output because the fusion output increases in ~ Rincreases in ~ R22, while the weight of coil , while the weight of coil supporting structure increases in~ R supporting structure increases in~ R 0.40.4 ( not ~R( not ~R33). ).
When the blanket lifetime ~ 30y, the When the blanket lifetime ~ 30y, the COE decreases ~20%COE decreases ~20% due to higher due to higher availability and lower cost for replacement.availability and lower cost for replacement.
0
5
10
15
20
25
30
35
8 10 12 14 16 18Rp, m
beta=1%
2%
3%
4%
Pf=1 GW
2 GW
4 GW
7 GW
Hiss=2
Hiss=1.5
1
γ = 1.15Rp/<a>=8.33∆ = 0.7 m
FFHR2
Sagara 30
ConclusionsConclusions
Design studies on FFHR have focused on new design Design studies on FFHR have focused on new design approaches to solve the key engineering issues of approaches to solve the key engineering issues of blanket space limitation and replacement difficulty. blanket space limitation and replacement difficulty.
(1)(1) The combination of improved support structure The combination of improved support structure and longand long––life breeder blanket STB is quite life breeder blanket STB is quite successful. successful.
(2)(2) The The ““screw coasterscrew coaster”” concept is advantageous inconcept is advantageous inheliotronheliotron reactors to replace inreactors to replace in--vessel vessel components. components.
(3)(3) The COE can be largely reduced by those The COE can be largely reduced by those improved designs. improved designs.
(4)(4) The key R&D issues to develop the STB concept The key R&D issues to develop the STB concept are elucidated.are elucidated.