Design concept of near term DEMO reactor with high temperature blanket

Transcription

1 Design concept of near term DEMO reactor with high temperature blanket Japan-US Workshop on Fusion Power Plants and Related Advanced Technologies March 16-18, 2009 Tokyo Univ. Mai Ichinose, Yasushi Yamamoto and Satoshi Konishi Inst. 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

2 Introduction Near term DEMO reactor with high temperature blanket Characteristics 1. Low initial cost 2. Near future technology 3. Non-nuclear hybrid with biomass Small Reactor R p <5.5 m Relaxed plasma requirements β N < 3.5 Q ~5 Modest wall loading P n < 2 MW/m 2 Fuel production LiPb blanket with SiC cooling panel T LiPb,out = 900 o C P fus < 1GW Net plant power output > 0

3 Non-Nuclear Hybrid Reactor Konishi s presentation nt ITER biofuel Q=5, Break-even η f =2.7 Q=1,η= T(kev) Electricity generation Q=20, η e =0.33 DEMO -0.6 Negative power High Plasma Q is not required Small major radius Relaxed plasma requirements Modest wall loading High Temperature extraction is required High temperature blanket should be developed.

4 Objectives 1.Investigation of possible design windows Small major radius, Q and fusion power Relaxed plasma requirements Modest wall loading 2. Design a realistic high temperature blanket Shielding TBR Thermal / hydraulic design The Purpose of this study is to examine the technical Feasibility of DEMO reactor based on this biomass-fusion Hybrid concept

5 Analysis flow Plasma analysis Plasma parameter Radial Build Q~5 P fus <1GW Neutronic analysis Neutron Shielding TBR Nuclear Heating Thermal hydraulic analysis Protection of RAFS vessel LiPb mass flow MHD pressure loss Temperature profile Realistic flow velocity

6 5.5 Plasma analysis (1/2) Δtf = 1.4 [m] Pfus [MW] Pn fus [MW/m2] 2 by TOPPER code Δ BKT n β N < 3.5 P n < 2 MW/m 2 P fus < 1GW Rp [m] βn 1.2 CS coil : Nb 3 Sn TF coil : Nb 3 Al Δ BKT R p 1.6 m a

7 5.5 Plasma analysis (2/2) Δ BKT Δtf = 1.4 [m] Pfus fus [MW] Pn [MW/m2] 2 n by TOPPER code Rp [m] βn Δ BKT R p, min 3.2< β N < < P n < 2 MW/m m 4.8 m

8 Main Parameter of the reactor Small size like ITER Positive output Driven Major radius, R p (m) Normalized beta, β N Neutron wall load, P n [MW/m 2 ] Energy multiplication factor, Q Safety factor, q ψ Bootstrap current fraction, f BS (%) Current Drive power, P CD (MW) Fusion power, P fus (MW) 763 Fuel energy ~2200MW Minor radius, a (m) Aspect ratio, A Elongation, κ 95 Maximum field, B max (T) Toroidal field, B T [T] Confinement enhancement, HH y2 1.1 Normalized Density, <n e >/n GW 0.47 Temperature, T e [kev] 17.0 Heat flux to divertor, P div [MW] 271 Plasma current, I p (MA) 13.4

9 Neutronic analysis Condition ~0.6 =1.4m Plasma analysis ~0.6 ΔBKT P n =1.7MW/m 2 Vacuum Vessel Low Temp. Shield High Temp. Shield Back Plates 0.2 Gap Breeder Zone Easy to remove the heat including radiation First Wall W armor Δ BKT =1.4m Plasma P n 1m 1m Blanket module

10 Blanket Structure Vessel material: F82H He 550 Breeder: Li 17 Pb 83 W armor ( 6 Li =90% enrichment ) Vacuum Vessel Low Temp. Shield High Temp. Shield Back Plates Breeder Zone ~900 First Wall W armor Cooling panel: SiC f /SiC He Scale model of SiC f /SiC Cooling panel

11 Shielding Performance Neutron flux [n/s s -1 cm2] -2 ] Breeder Shield F.W. + Zone Requirement [n/s s -1 cm2] -2 ] Gamma radiation dose? En > 0.1MeV V.V Blanket Thickness [cm] by ANISN code

12 TBR LocalTBR = NetTBR ( 1.05) 1.40 Coverage (= 0.75 ) SiC Panel Inboard TBR inboard = 1.35 LiPb SiC LiPb [cm] Outboard 40 TBR outboard 2 30 = 1.43 Breeder Zone LiPb SiC LiPb LocalTBR= 1.41 [cm]

13 Protection of RAFS vessel (1/2) Nuclear Heating [MW/m 3] RAFS temperature Heat (Plasma) MW/m2 LiPb 900 o C ~50MW/m3 Heat (W armor) MW/m 2 50 ~30MW/m MPa He First Wall Thickness [cm] W armor F82H SiC

14 Heat flux Temperature [ ] Protection of RAFS vessel (2/2) m/s 60 m/s 80 m/s F82H He F82H SiC He SiC Tmax 550 ANSYS Ver Wall thickness [cm] LiPb 900

15 LiPb mass flow Mass flow (LiPb) [kg/s] Nuclear heating (LiPb) = 4.18 MW / module (Area:1m 2 -Thickness0.5m) T out,lipb =900 o C 200 Temperature difference is required T in, LiPb [ ]

16 MHD Pressure Loss (LiPb) C: Wall conductance ratio Ha: Hartmann number σ: Electrical conductivity P loss= {C/(1+C)+1/Ha}σ(V B) B V: Velocity B: Magnetic flux density Maximum at the inboard blanket duct MHD Ploss [Pa/m] C=0.01 C=0 B=10 [T] (metal wall) (insulated wall) Velocity(LiPb) [m/s] 2 V LiPb [m/s] 1.5 φ=0.10m φ φ=0.20m Velocity(LiPb) [m/s] T [ ] in, LiPb

17 Summary of the design 1. Design windows was obtained Small major radius, Q and fusion power Relaxed plasma requirements Modest wall loading 2. High temperature blanket was designed Shielding TBR Sufficient for power demonstration Thermal/hydraulic design Satisfy the request Near future technology High Temp. extraction with SiC cooling panel Reasonable MHD pressure loss

18 Conclusion Univ. 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 Characteristics 1. Low initial cost 2. Near future technology 3. Non-nuclear hybrid with biomass

19 The following is support document