Upload
dinhnguyet
View
214
Download
0
Embed Size (px)
Citation preview
Design concept of near term DEMO
reactor with high temperature
blanket
○○○○Mai Ichinose, Yasushi Yamamoto and
Satoshi Konishi
Japan-US Workshop on
Fusion Power Plants and Related Advanced Technologies
March 16-18, 2009
Tokyo Univ.
Inst. of Advanced Energy, Kyoto Univ.
Institute of Advanced Energy, Kyoto Univ.
Acknowledgement : The authors would like to express their gratitude to Dr.
K.Tobita and S.Nishio for their support and permission for TOPPER code
Introduction
1. Low initial cost
2. Near future technology
3. Non-nuclear hybrid
with biomass
Near term DEMO reactor with high temperature blanket
Characteristics
Fuel production
Small Reactor
Institute of Advanced Energy, Kyoto Univ.
Relaxed plasma requirements
Rp<5.5 m
ββββN < 3.5
Pn < 2 MW/m2
Modest wall loading
Q ~5
LiPb blanket with SiC cooling panel
TLiPb,out = 900oC
Pfus < 1GW Net plant power output > 0
Konishi’s presentation
1810
1022
1021
1020
1019
1 10 100
nnnnt
T(kev)T(kev)T(kev)T(kev)
Break-even
Q=Q=Q=Q=1111,,,,ηηηη====1111
Electricity
generation
Q=Q=Q=Q=20,,,,ηηηηe====0.33
biofuel
Q=Q=Q=Q=5555,,,,ηηηηffff====2222....7777
-0.6
Negative power
ITER
DEMO
Non-Nuclear Hybrid ReactorInstitute of Advanced Energy, Kyoto Univ.
High Plasma Q is not required
High Temperature extraction
is required
Small major radius
Relaxed plasma requirements
Modest wall loading
High temperature blanket
should be developed.
Objectives
Shielding
TBR
Thermal / hydraulic design
Small major radius, Q and fusion power
Relaxed plasma requirements
Modest wall loading
2. Design a realistic high temperature blanket
The Purpose of this study is to examine the
technical Feasibility of DEMO reactor based on
this biomass-fusion Hybrid concept
1.Investigation of possible design windows
Institute of Advanced Energy, Kyoto Univ.
Analysis flow
Neutron Shielding
TBR
Nuclear Heating
Protection of RAFS vessel
Plasma analysis
Radial Build
Neutronic analysis
Thermal hydraulic
analysis
Plasma parameter
LiPb mass flow
MHD pressure loss
Institute of Advanced Energy, Kyoto Univ.
Q~5
Pfus<1GW
Temperature profile
Realistic flow velocity
Plasma analysis (1/2)
3 3.5 4 4.54
4.5
5
5.5
8001000
1400
1200
ΔΔΔΔtftftftf = 1.4 [m]
Rp [m]
ββββNNNN
Pfus [MW]Pn [MW/m2]
1
2
1.5
3 3.5 4 4.54
4.5
5
5.5
8001000
1400
1200
ΔΔΔΔtftftftf = 1.4 [m]
Rp [m]
ββββNNNN
Pfus [MW]Pn [MW/m2]
1
2
1.5
by TOPPER code
CS coil : Nb3Sn
TF coil : Nb3Al
Rp a
1.6 m1.2
ΔΔΔΔBKT
Institute of Advanced Energy, Kyoto Univ.
ΔΔΔΔBKT
Pn < 2 MW/m2
Pfus < 1GW
2n
fusββββN < 3.5
3 3.5 4 4.54
4.5
5
5.5
8001000
1400
1200
ΔΔΔΔtftftftf = 1.4 [m]
Rp [m]
ββββNNNN
Pfus [MW]Pn [MW/m2]
1
2
1.5
3 3.5 4 4.54
4.5
5
5.5
8001000
1400
1200
ΔΔΔΔtftftftf = 1.4 [m]
Rp [m]
ββββNNNN
Pfus [MW]Pn [MW/m2]
1
2
1.5
ΔΔΔΔBKT
by TOPPER code
Institute of Advanced Energy, Kyoto Univ.
Plasma analysis (2/2)
ΔΔΔΔBKT 1.6 m1.2
Rp, min 4.8 m4.2
3.2< ββββN < 3.5
1.7< Pn < 2 MW/m2
2n
fus
Main Parameter of the reactor
Institute of Advanced Energy, Kyoto Univ.
Minor radius, a (m)
Aspect ratio, A
Elongation, κκκκ95
Maximum field, Bmax (T)
Toroidal field, BT [T]
1.4
3.2
1.8
15.7
5.9
Safety factor, qψψψψBootstrap current fraction, fBS (%))))Current Drive power, PCD (MW)
3.4
1.7
4.5
Normalized beta, ββββN
Neutron wall load, Pn [MW/m2]
Energy multiplication factor, Q
2.9
45.9
168
Confinement enhancement, HHy2
Normalized Density, <ne>/nGW
Temperature, Te [keV]
Heat flux to divertor, Pdiv [MW]
Plasma current, Ip (MA)
1.1
0.47
17.0
271
13.4
like ITER
Driven
Fusion power, Pfus(MW) 763
4.5
Major radius, Rp (m) Small size
Positive
output Fuel energy ~2200MW
Neutronic analysis Condition
Plasma analysis
Pn =1.7MW/m2
=1.4mΔΔΔΔBKT
Back Plates
High Temp. Shield
Breeder Zone
First Wall
Low Temp. Shield
Vacuum Vessel
Pn
Plasma
~0.6
~0.6
0.2
=1.4m
ΔΔΔΔ BKT
Gap
W armor
Institute of Advanced Energy, Kyoto Univ.
Easy to remove the
heat including
radiation
1m
1m
Blanket module
Blanket StructureBack Plates
High Temp. Shield
Breeder Zone
First Wall
Low Temp. Shield
Vacuum Vessel
W armor
Breeder: Li17Pb83
(6Li =90% enrichment )
Vessel material: F82H – He
W armor
~900℃℃℃℃
≦≦≦≦550℃℃℃℃
Cooling panel: SiCf/SiC– He
Scale model of
SiCf/SiC Cooling
panel
Institute of Advanced Energy, Kyoto Univ.
V.V
Shielding PerformanceInstitute of Advanced Energy, Kyoto Univ.
1.3××××109Requirement
F.W. +ShieldBreeder
Zone
Gamma radiation dose ?
0 50 100108
109
1010
1011
1012
1013
1014
1015
1016
Blanket Thickness [cm]
Neutron flux [n/s・・・・cm2]
En > 0.1MeV
1.6××××1010 [n/s・・・・cm2]
by ANISN code[n・・・・s-1・・・・cm-2]
[n・・・・s-1・・・・cm-2]
TBRInstitute of Advanced Energy, Kyoto Univ.
20
LiPb LiPb
302
SiC
LocalTBR= 1.41
TBRoutboard
TBRinboard = 1.35
= 1.43
[cm]
Inboard
Outboard
LocalTBR = NetTBR (≧≧≧≧1.05) ≧≧≧≧1.40Coverage (= 0.75 )
40
LiPb LiPb
302
SiC
[cm]
Breeder Zone
SiC Panel
0 1 2 30
50
100
150
First Wall Thickness [cm]
Nuclear Heating [MW/m ]3
Protection of RAFS vessel (1/2)
Institute of Advanced Energy, Kyoto Univ.
LiPb
900oCHeat (W armor)
0.0347 MW/m2
Heat (Plasma)
0.45 MW/m2
~50MW/m3
~30MW/m3
SiC
He
RAFS temperature
≦≦≦≦550℃℃℃℃
350℃℃℃℃8MPa
F82H
W armor
Protection of RAFS vessel (2/2)Institute of Advanced Energy, Kyoto Univ.
≦≦≦≦550℃℃℃℃Tmax
HeF82H SiC SiCHeF82H
LiPb900℃℃℃℃
ANSYS Ver.10.0Heat
flux
0 1 2 3300
400
500
600
700
800
900
Wall thickness [cm]
Temperature [℃℃ ℃℃
]40 m/s60 m/s80 m/s
300 400 500 600 700 8000
100
200
300
400
500
Mass flow [kg/s]
T [℃℃℃℃]in, LiPb
(LiPb)
LiPb mass flow
Nuclear heating (LiPb)
= 4.18 MW / module
(Area:1m2-Thickness0.5m)
Temperature difference is required
Institute of Advanced Energy, Kyoto Univ.
out,LiPb=900 oCT
0 0.5 1 1.5 2100
101
102
103
104
105
Velocity(LiPb) [m/s]
MHD Ploss [Pa/m]
C=0
C=0.01
MHD Pressure Loss (LiPb)P loss= {C/(1+C)+1/Ha}σσσσ(V××××B)××××B
B=10 [T]
(insulated wall)
(metal wall) φφφφ
Maximum at the inboard blanket duct
300 400 500 600 700 8000
0.5
1
1.5
2
V [m/s]
T [℃℃℃℃]
φφφφ=0.10m
φφφφ=0.20m
in, LiPb
LiPb
σσσσ: Electrical conductivity
V: Velocity
B: Magnetic flux densityHa: Hartmann number
C: Wall conductance ratio
Institute of Advanced Energy, Kyoto Univ.
Velocity(LiPb) [m/s]
Summary of the design
Modest wall loading
2. High temperature blanket was designed
1. Design windows was obtained
Small major radius, Q and fusion power
Relaxed plasma requirements Near future
technology
Sufficient for power demonstration
Shielding
TBR
Thermal/hydraulic design
Satisfy the request
High Temp. extraction with SiC cooling panel
Reasonable MHD pressure loss
Institute of Advanced Energy, Kyoto Univ.
Institute of Advanced Energy, Kyoto
Univ.
Institute of Advanced Energy, Kyoto Univ.
Conclusion
1. Low initial cost
2. Near future technology
Characteristics
When we will take advantage of the hybrid of
biomass and fusion, it will be feasible to develop a
small DEMO reactor that has the features of
3. Non-nuclear hybrid with biomass
The following is support documentThe following is support document
Institute of Advanced Energy, Kyoto Univ.Institute of Advanced Energy, Kyoto Univ.