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Japan-US Workshop on Fusion Power Plants Related Advanced Technologies with participants from China and Korea ( Kyoto University, Uji, Japan, 26-28 Feb.

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Presentation on theme: "Japan-US Workshop on Fusion Power Plants Related Advanced Technologies with participants from China and Korea ( Kyoto University, Uji, Japan, 26-28 Feb."— Presentation transcript:

1 Japan-US Workshop on Fusion Power Plants Related Advanced Technologies with participants from China and Korea ( Kyoto University, Uji, Japan, Feb )

2 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb ) 2/22

3 Schematic view of FFHR-d1 (by H. Tamura) Vertical slices of FFHR-d1 (by T. Goto) FFHR-d1 R c = 15.6 m B c = 4.7 T P fusion ~ 3 GW 3/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

4 Experiment Reactor J. Miyazawa et al., Fusion Eng. Des. 86, 2879 (2011) p GB-BFM (  ) =  0 n e (  ) 0.6 P 0.4 B 0.8 J 0 (2.4  /  1 ) 4/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

5  DPE ∝ ((P dep /P dep1 ) avg,reactor ) / (P dep /P dep1 ) avg,exp ) 0.6 = (0.65 / (P dep /P dep1 ) avg ) 0.6  …where 0.65 is the assumed peaking factor of the alpha heating in the reactor J. Miyazawa et al., Nucl. Fusion 52 (2012) FFHR-d1 5/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

6  c = 1.25 One from the standard configuration of R ax = 3.60 m,  c = 1.25  c = 1.20 Another from the high aspect ratio configuration of R ax = 3.60 m,  c = 1.20 The high-aspect ratio configuration is effective for Shafranov shift mitigation  c = (m a c ) / (l R c ) is the pitch of helical coils, where m = 10, l = 2, a c ~ 0.9 – 1.0 m, and R c = 3.9 m in LHD 6/22

7 ・ device sizeRc = 15.6 m ・ mag. fieldBc = 4.7 T ・ fusion output P DT ~ 3 GW ・ radial profiles given by DPE ・ HINT2 ・ VMEC ・ TERPSICHORE ・ GKV-X Detailed physics analyses on MHD, neoclassical transport, and alpha particle transport using profiles given by the DPE method have been started 7/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

8 By Y. Suzuki R ax = 14.4 m “Vacuum” R ax = 14.4 m “w/o B v ” R ax = 14.0 m “w/ B v ” R ax = 14.4 m “Vacuum” R ax = 14.4 m “w/o B v ” R ax = 14.0 m “w/ B v ” 8/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

9 By S. Satake Neoclassical heat flow in various cases Comparison with P  in Case B w/ B v 9/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

10 By Y. Masaoka and S. Murakami (Kyoto Univ.) 10/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

11 R ax = 14.4 m Vacuum R ax = 14.4 m  0 ~ 8.5 % R ax = 14.0 m  0 ~ 10 % By R. Seki 11/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

12 By R. Seki and T. Goto 12/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

13 Case A Rax = 3.60 m  c = n eo /n ebar ~ 1.5 MHD equilibrium Shafranov shift is too large Neoclassical → ✖ α confinement → ✖ Case B Rax = 3.60 m  c = 1.20 n eo /n ebar ~ 0.9 MHD equilibrium Shafranov shift is mitigated by applying Bv Neoclassical → α confinement → α heating profile → ✖ (hollow) Case C Rax = 3.55 m  c = 1.20 n eo /n ebar ~ 1.5 MHD equilibrium ? Neoclassical → ? α confinement → ? α heating profile → ? Both NC and α confinement are bad in the Case A, due to the large Shafranov shift. Both NC and α confinement are good in the Case B where Shafranov shift is mitigated. However, α heating profile is hollow (not consistent with the assumption for  DPE ). Case C’ Rax = 3.55 m  c = 1.20 n eo /n ebar ~ 1.5 MHD equilibrium ? Neoclassical → ? α confinement → ? α heating profile → ? α heating efficiency → ? In the Case C’, α heating efficiency,   = 0.85 is assumed (   = 1.0 in Cases A, B, and C). A new profile “Case C” has been chosen. Peaked density profile as the Case A. High-aspect ratio as the Case B but more inward-shifted. 13/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

14 14/22

15 15/22 R ax = 14.4 m,  0 ~ 7.5 %, “w/o B v ” R ax = 14.2 m,  0 ~ 7.6 %, “w/o B v ” R ax = 14.0 m,  0 ~ 7.6 %, “w/ B v ” R ax = 14.0 m,  0 ~ 9.1 %, “w/ B v ” R ax = 14.0 m,  0 ~ 8.2 %, “w/ B v ” MHD equilibrium is ready Bad? Good? ?? Bad Good

16 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, FFHR-d1 炉設計合同タスク会合( 1 Feb ) 16/22

17 17/22

18 - The confinement enhancement factor is proportional to the 0.6 power of the heating profile peaking factor:  DPE = ((P dep /P dep1 ) avg,reactor ) / (P dep /P dep1 ) avg,exp ) 0.6  = (0.65 / (P dep /P dep1 ) avg ) The heating profile peaking factor with the central auxiliary heating can be larger than Assume that the auxiliary heating is focused on the center and expressed by the delta function, i.e., (P dep /P dep1 ) avg,reactor = 0.65 (1/C aux ) +(1.0 – 1/C aux ) = 1.0 – 0.35/C aux - The confinement enhancement factor is given by  DPE* = ((1.0 – 0.35/C aux ) / (P dep /P dep1 ) avg ) 0.6 C aux = 1.0 C aux = 2.0 C aux = /22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb ) P  – P B = (1 / C aux ) P reactor P aux = (1 – 1 / C aux ) P reactor

19 FFHR-c1.0, C aux = /22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

20 FFHR-c1.1, C aux = /22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

21 21/22

22 22/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

23 FFHR-d1, C aux = /22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies ( Kyoto Univ., Feb )

24 The central pressures similar to those in the  c = configuration have also been achieved in the  c = 1.20 configuration, in spite of smaller plasma volume (i.e., better confinement in  c = 1.20) f  as low as ~3 has been achieved P reactor can be as low as ~200 MW Density peaking factor is high (density peaking was observed after NB#1 break down) 24/22


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