Assessment of Spherical Reactors Comparing with Other Plants

Kozo YAMAZAKI1, Tetsutarou OISHI1, Kanae BAN1, Takuya KONDOH1, Hideki ARIMOTO1, Yoshio NAGAYAMA2

1Nagoya Univ., Chikusa-ku, Nagoya 464-8603, Japan 2 National Institute for Fusion Science, Toki, Gifu 509-5292, Japan

1 OUTLINE

1. Introduction 2. Burning Simulation using “TOTAL” (Toroidal Transport Analysis Linkage) 3. Life-Cycle Assessment using “PEC” (Plasma, Engineering and Cost) Cost, CO2, Energy Payback etc. 4. Summary

2 1. Introduction ST-Relevant Research (1/3) TOKASTAR (miniature hybrid experiment) 29-3-2 (Oral) M. Hasegawa et al., “Low aspect ratio plasma in tokamak-helical hybrid device TOKASTAR-2”. 28-3P-5 H. Ozeki et al., “Confinement Analysis of Spherical Tokamak- Hybrid Configurations”.

TOKASTAR ST-Relevant Research (2/3)

TOTAL code (burning plasma simulation) 28-3P-18 D. Kurita et al., “Neoclassical Tearing Mode Analysis in Spherical Tokamak Burning Plasmas”. 28-3P-20 T. Oishi et al., “Integrated Analysis on the Current Profile and the Operational Scenario of D-3He Spherical Tokamak Reactors”. 35 12 3.5 n e 30 10 jz 3 25 Te(no NTM) 8 2.5

20 n e ) ( V q

6 m

e 2 Toroidal Transport Analysis Linkage - k 15 3 ) (

e q

T 4 10 1.5 Te(m/n=3/2) 5 2 jBS 1 Te(m/n=2/1) 0 0 0.5 0 0.2 0.4 0.6 0.8 1 0 0.2 0.4 0.6 0.8 1 ρ ρ ST-Relevant Research (3/3) PEC (reactor system analysis) 30-2-1 (Oral) K. Yamazaki et al., “Assessment of Spherical Tokamak Reactors Comparing with Other Fusion Power Plants”. 30-2-2 (Oral) K. Ban et al., “Life Cycle Assessment for Energy Payback of Spherical Tokamak Reactors”.

Reflector

SC or NC Coil SC Coil Plasma Plasma Shield Physics, Engineering, Cost Shield

Gap Gap Blanket Blanket First Wall First Wall Rm B B max R m max R0 B 0 B 0 R 0 2. Burning Plasma Simulation: Using “TOTAL” Code History Start (~1980) Tokamak (2nd stability) Vol.25 (1985) 1543. Helical Analysis Nuclear Fusion Vol.32 (1992) 633. Burning Simulation (Tokamak & Helical) Nuclear Fusion Vol.49 (2009) 055017.

Based on JT-60U ITB operation and

LHD e-ITB data. 6 Main Feature of “TOTAL” is to perform both Tokamak and Helical Analyses

Core Plasma Equilibrium Tokamak: :2D APOLLO Helical: 3D VMEC, DESCUR, NEWBOZ Transport Tokamak： TRANS、 GLF23、NCLASS Toroidal Transport Analysis Linkage Helical： HTRANS, GIOTA World Integrated Stability NTM, Sawtooth, Ballooning mode Modeling

Edge transport H-mode edge transport TOTAL-T,-H(J) Impurity IMPDYN (rate equation) Tungsten TOPICS(J) TASK(J) ADPAC （various cross-section） CRONOS(EU) Fueling NGS (neutral gas shielding) model TRANSP(US) mass relocation model ASTRA(RU) NBI HFREYA, FIFPC Puffing AURORA

Divertor density control, two-point divertor

7 Time evolution of reversed-shear-mode operation with continuous HFS pellet injection

1GWe Tokamak Reactor TR-1

8 3. System Code Assessment: using “PEC” Code History Tokamak Ignition Design (~1980) Helical Design Assessment IAEA-Montreal (1996) Helical & Tokamak Assessment IAEA-Lyon (2002)

Cost/CO2/EPR Analysis in MFE reactors IAEA-Geneva (2008)

Cost/CO2/EPR Analysis in MFE & IFE reactors

IAEA-Daejeon (2010) 9 Main Feature is Comparative Assessment of Various Reactors

MFE IFE Physics, Engineering, Cost Plasma beta value β Driver Energy (MJ) Ignition margin Driver Efficiency (%) Target Electric Output (MWe) Target Electric Output (MWe) input input Assumption of Fuel Mass (g) ①Thickness of Blanket (m) Fusion Energy (MJ) ②Plasma major radius Rp(m) Plasma Parameters Assumption of Repetition Ratio frep(Hz) Power Balance Power Balance Check

Ptarget, tblanket, igm Check Ptarget

Final Radial Build Final Radial Build output output

Cost & CO2 emission amounts Cost & CO2 emission amounts 10 Reference Designs

11 ST(NC coil) TR(SC coil) 10 10 1.5 8 2.0 8 2.5 6 6 K=3.0 K=1.5 4 4 2.0 2.5

2 2

0 0 1 1.5 2 2.5 3 3.5 4 1 1.5 2 2.5 3 3.5 4 A 10 10 A

8 8

6 6 K=1.5 2.0 K=1.5 4 4 2.0 2.5 2.5 2 3.0 2

0 0 1 1.5 2 2.5 3 3.5 4 1 1.5 2 2.5 3 3.5 4 A 12 A ST(NC coil) TR(SC coil) 20 20 1.5 15 2.0 2.5 K=3.0 15

10 K=1.5 10 2.0 2.5 5 5

0 0 1 1.5 2 2.5 3 3.5 4 1 1.5 2 2.5 3 3.5 4 A A 20 20 1.5 2.0 2.5 15 15 K=3.0

10 10 K=1.5 2.0 2.5 5 5

0 0 1 1.5 2 2.5 3 3.5 4 1 1.5 2 2.5 3 3.5 4 A A 13 16 7

14 6 TR-1 12 HR-1 10 5 HR-1 8 TR-1 4 6 ST-1 3 ST-1 4

2 2 2 3 4 5 6 7 8 9 2 3 4 5 6 7 8 9  (TR,ST)  (TR,ST) N N <>(%) (HR) <>(%) (HR)

6 5

5 4 TR-1 ST-1 4 HR-1 3 3 ST-1 2 2 TR-1 HR-1 1 1

0 0 2 3 4 5 6 7 8 9 2 3 4 5 6 7 8 9  (TR,ST)  (TR,ST) N N <>(%) (HR) <>(%) (HR) Multiple Approaches Have Similar COE,CO2, EPR

TR ST HR IR Reflector Shield SC or SC Coil Plasma Plasma NC Coil Plasma Shield Blanket Shield Shield SC Coil Gap Gap Gap Blanket Blanket Laser Rm First Wall Blanket First Wall R First Wall R Bmax m B First Wall 0 B R Bm max 0 m B R0 B R0 ax 0 0 Tokamak Reactor Spherical Torus Helical Reactor Inertial Reactor 9.8 ¢/kWh 10.4 ¢/kWh 11.1 ¢/kWh 10.7 ¢/kWh

9.2 g-CO2/kWh 9.1 g-CO2/kWh 9.9 g-CO2/kWh 12.9 g-CO2/kWh EPR ~ 19.3 EPR ~ 18.1 EPR ~ 12.1 EPR ~ 26.6

1GWe Plant 1¢~1¥

15 Comparisons with Other Power Plants

wind 13.2 * 29.5 • solar * 62.3 Hondo et al., Socio-economic 53.4 •Research Center (2000) water 11.2 * 11.3 coal* 5.4 975 oil * 10.1 742 fission 5 * 21.6 IR 10.7 12.9 TR (D-3He) 9.4 12.4 TR(F-F) 8.7 10.4 HR 11.1 9.9 ￠/kWh ST 10.4 g-CO2/kWh 9 TR 9.8 1¢~1¥ 9.2 0 20 40 60 80 100 ￠/kWh , g-CO2/kWh 16 ST (with SC coil) is better than ST(NC coil) and TR (SC coil).

20

ST(NC1.5 coil) 2.0 2.5 COE 15 K=3.0 CO2 TR(SCK=1.5 coil) 10 1/EPR 2.0 2.5 5 ST(SC coil, no CS, no inboard blanket) 0 1 1.5 2 2.5 3 3.5 4 A ４. Summary

● Comparative system studies have been done for several magnetic fusion energy reactors (TR, ST, HR) and inertial fusion energy reactor (IR) . ● The advantages of TR in COE and of ST in lifetime CO2 emission reduction are clarified. ● The optimized ST system with SC coils and without inboard blanket is suggested. ● Comparing fusion reactors with other electric power generation systems, we confirmed that fusion reactor, especially ST, does not emits large CO2 amount. 18