Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 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 Slide 13 Case A Rax = 3.60 m c = 1.254 n eo /n ebar ~ 1.5 MHD equilibrium Shafranov shift is too large Neoclassical 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 confinement heating profile (hollow) Case C Rax = 3.55 m c = 1.20 n eo /n ebar ~ 1.5 MHD equilibrium ? Neoclassical ? confinement ? heating profile ? Both NC and confinement are bad in the Case A, due to the large Shafranov shift. Both NC and confinement are good in the Case B where Shafranov shift is mitigated. However, 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 ? confinement ? heating profile ? heating efficiency ? In the Case C, 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 Slide 14 14/22 Slide 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 Slide 16 J. Miyazawa, Core Plasma Design for FFHR-d1 and c1, FFHR-d1 1 Feb. 2013 16/22 Slide 17 17/22 Slide 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 Slide 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 Slide 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 Slide 21 21/22 Slide 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 Slide 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 Slide 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