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

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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

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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

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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

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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

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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

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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)

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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

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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

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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.