1. Japan-US Workshop on Fusion Power Plants Related Advanced Technologies with participants from China and Korea （ Kyoto University, Uji, Japan, 26-28 Feb. 2013 ）

2. J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies （ Kyoto Univ., 26-28 Feb. 2013 ） 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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

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) 123007. 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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

13. Case A Rax = 3.60 m  c = 1.254 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., 26-28 Feb. 2013 ）

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. 2013 ） 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 ) 0.6 - The heating profile peaking factor with the central auxiliary heating can be larger than 0.65 - 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 = 7.0 18/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies （ Kyoto Univ., 26-28 Feb. 2013 ） P  – P B = (1 / C aux ) P reactor P aux = (1 – 1 / C aux ) P reactor

19. FFHR-c1.0, C aux = 1.8 19/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies （ Kyoto Univ., 26-28 Feb. 2013 ）

20. FFHR-c1.1, C aux = 1.0 20/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies （ Kyoto Univ., 26-28 Feb. 2013 ）

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., 26-28 Feb. 2013 ）

23. FFHR-d1, C aux = 1.0 23/22 J. Miyazawa, “Core Plasma Design for FFHR-d1 and c1”, Japan-US WS on Fusion Power Plants Related Advanced Technologies （ Kyoto Univ., 26-28 Feb. 2013 ）

24. The central pressures similar to those in the  c = 1.254 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