Multifarious Physics Analyses of the Core Plasma Properties in a Helical DEMO Reactor FFHR-d1

Size: px
Start display at page:

Download "Multifarious Physics Analyses of the Core Plasma Properties in a Helical DEMO Reactor FFHR-d1"

Transcription

1 1 FTP/P7-34 Multifarious Physics Analyses of the Core Plasma Properties in a Helical DEMO Reactor FFHR-d1 J. Miyazawa 1, M. Yokoyama 1, Y. Suzuki 1, S. Satake 1, R. Seki 1, Y. Masaoka 2, S. Murakami 2, Y. Narushima 1, M. Nunami 1, T. Goto 1, C. Suzuki 1, H. Funaba 1, I. Yamada 1, R. Sakamoto 1, G. Motojima 1, H. Yamada 1, A. Sagara 1, and the FFHR Design Group 1 National Institute for Fusion Science, Toki, Gifu , Japan 2 Department of Nuclear Engineering, Kyoto University, Kyoto , Japan contact of main author: miyazawa@lhd.nifs.ac.jp Abstract. Physics assessments on the MHD equilibrium and stability, the neoclassical transport, and the alpha particle transport, etc., are being carried out for a helical fusion DEMO reactor named FFHR-d1, using radial profiles extrapolated from LHD. A large Shafranov shift is foreseen in FFHR-d1 due to its high-beta property. This leads to deterioration in the neoclassical transport and alpha particle confinement. Plasma position control using the vertical magnetic field has been examined and shown to be effective for Shafranov shift mitigation. Especially in a high aspect ratio configuration, it is possible to keep the magnetic surfaces similar to those in vacuum with high central beta of 8.5 % by applying a proper vertical magnetic field. As long as the Shafranov shift is mitigated, the neoclassical thermal transport can be kept at a level compatible with the alpha heating power. The alpha particle loss can also be kept at a low level as long as the loss boundary of alpha particles is on the blanket surface and the plasma position control is properly applied. The lost positions of alpha particles are localized around the divertor region that is located behind the blanket in FFHR-d1. 1. Introduction FFHR-d1 is a heliotron type DEMO reactor of which the conceptual design activity has been started since 2010 [1,2]. It is possible to sustain the burning plasma without auxiliary heating (i.e., self-ignition) in FFHR-d1, since there is no need of plasma current drive in heliotron plasmas. The device size is 4 times enlarged from LHD [3], i.e., the major radius of the helical coil center is 15.6 m, the magnetic field strength at the helical coil center is 4.7 T, and the fusion output is ~3 GW [1,2]. One of the distinguished features of FFHR-d1 compared with the former FFHR design series is the robust similarity with LHD. The arrangement of superconducting magnet coils in FFHR-d1 is similar to that of LHD, except a pair of planar poloidal coils omitted to maximize the maintenance ports [4]. This makes it reasonable to assume a similar MHD equilibrium as observed in LHD for FFHR-d1, as long as the beta profiles in these two are similar. In FFHR-d1, radial profiles of density and temperature are determined by multiplying proper enhancement factors on those obtained in LHD, according to the DPE (Direct Profile Extrapolation) method [5]. The enhancement factors are calculated consistently with the gyro-bohm model. Therefore, the global confinement properties as expressed in ISS95 [6] or ISS04 [7] are kept in FFHR-d1. It should be noted that the temperature profile is not necessarily fixed in DPE and various kinds of temperature profiles observed in LHD can be extrapolated to FFHR-d1. This paper discusses the results of detailed physics analyses of the core plasma properties in FFHR-d1. The radial profiles employed are extrapolated from the data obtained in LHD as described in Section 2. Results of physics analyses on the MHD equilibrium and stability, the

2 2 FTP/P7-34 neoclassical transport, and the alpha particle transport are given in Section 3. Finally, these are summarized in Section Radial Profiles for Detailed Physics Analyses Two sets of radial profiles are used in the detailed physics analyses described in the next section. The first set is named case A, which is shown in FIG. 1. These profiles are extrapolated using the DPE method from the data obtained in the standard configuration of R ax vac = 3.60 m, B 0 = 2.75 T, and γ c = in LHD, where R ax vac is the major radius of the magnetic axis in vacuum, B 0 is the magnetic field strength at R ax vac, and γ c = (m a c ) / (l R c ) is the pitch of helical coils (m, l, a c, and R c are the toroidal mode number (= 10), the number of helical coils (= 2), the minor radius (~ m) and the major radius (= 3.9 m) of the helical coils, respectively). The other set is the case B shown in FIG. 2, of which the profiles are extrapolated from the data obtained in the high aspect ratio configuration of R ax vac = 3.60 m, B 0 = 1.50 T, and γ c = The averaged plasma aspect ratio, R/a, in vacuum FIG. 1. Radial profiles in the case A of (a) the electron density, (b) the electron temperature, and (c) the plasma beta, (d) the alpha birth profile, and (e) the integrated heating power in FFHR-d1 (closed circles) extrapolated from experimental results of LHD (#96164, t = s, R ax vac = 3.60 m, B 0 = 2.75 T, γ c = 1.254, open circles), where enhancement factors for the energy confinement, γ DPE, the density, f n, the temperature, f T, and the plasma beta, f β, are assumed to be 1.29, 2.65, 6.59, and 5.10, respectively. FIG. 2. Radial profiles in the case B of (a) the electron density, (b) the electron temperature, and (c) the plasma beta, (d) the alpha birth profile, and (e) the integrated heating power in FFHR-d1 (closed circles) extrapolated from experimental results of LHD (#109602, t = s, R ax vac = 3.60 m, B 0 = 1.50 T, γ c = 1.20, open circles), where enhancement factors for the energy confinement, γ DPE, the density, f n, the temperature, f T, and the plasma beta, f β, are assumed to be 1.14, 12.0, 5.66, and 5.89, respectively.

3 3 FTP/P7-34 increases from ~5.6 to ~6.4 as γ c is decreased from to The temperature and density profiles in LHD are measured by Thomson scattering [8]. The density signal of the Thomson scattering is calibrated by the line-density measured by the millimetre wave interferometer [9] that is measuring the same line of sight with the Thomson scattering. In both cases, no auxiliary heating is applied and the conduction power needed to sustain the plasma, P reactor, which is estimated by the gyro-bohm model, is given by the alpha heating power, P α, minus the Bremsstrahlung loss, P B. In other words, the self-ignition condition is satisfied in both cases. The beta profile in the case A is more peaked than that in the case B. As a result, the central beta, β 0, is ~9 % in the case A and ~8 % in the case B. Note that peaked beta profiles are not favourable from the point of view of Shafranov shift mitigation. Furthermore, the high aspect ratio configuration itself is effective for Shafranov shift mitigation [10]. The beta profiles in the standard and the high aspect ratio configurations are compared in FIG. 3. In both cases, the peak position of the beta profile, i.e., the magnetic axis, is moved from the initial position of R ax vac = 3.60 m, due to the Shafranov shift. As seen in the figure, a higher central beta than that in the standard configuration is achieved with a smaller Shafranov shift in the high aspect ratio configuration. 3. Detailed Physics Analyses Using radial profiles shown in FIGs. 1 and 2, detailed physics analyses have been carried out. At first, the MHD equilibrium for these profiles is calculated by HINT2 [11] and VMEC [12]. The MHD stability of these equilibriums is evaluated by TERPSICHORE [13]. Then, the neoclassical transport is calculated by GSRAKE [14] and FORTEC-3D [15]. The alpha particle confinement is estimated by GNET [16] and MORH [17] MHD Equilibrium FIG. 3. Comparison of the beta profiles in the standard configuration (circles) and the high aspect ratio configuration (plusses and crosses). One can use the MHD equilibrium that can fit the radial profiles observed in the experiment, as long as the beta enhancement factor, f β, is equal to one. In the cases shown in FIGs. 1 and 2, f β is 5 6. Therefore, it is necessary to calculate the MHD equilibrium for these cases. As a first step, we tried to reconstruct the MHD equilibrium for the case A by HINT2 [11]. However, it was difficult to obtain the MHD equilibrium with the beta profile shown in FIG. 1(c). As shown in FIG. 4(b), the Shafranov shift is already significant at β 0 ~ 8.5 % and the magnetic surfaces in the edge region become destructed. To resolve this, plasma position control using the vertical magnetic field, B v, is effective. In the LHD type heliotron devices, B v can alter the magnetic axis position in vacuum. In the case of FIG. 4(c), B v identical to that used to form an inward shifted configuration of R ax vac = 14.0 m, which corresponds to R ax vac = 3.50 m in LHD, is applied instead of B v for R ax vac = 14.4 m. Remember that FFHR-d1 is 4 times enlarged from LHD and R ax vac = 14.4 m in FFHR-d1 corresponds to R ax vac = 3.60 m in

4 FTP/P (a) (a) vacuum vacuum (b) (b) w/o Bv w/o Bv (c) (c) w/ Bv FIG. 4. Magnetic surfaces calculated by the HINT2 code for the case A with (a) β0 = 0 %, Raxvac = 14.4 m (vacuum), (b) β0 ~ 8.5 %, Raxvac = 14.4 m (w/o Bv control), and (c) β0 ~ 10 %, Raxvac = 14.0 m (w/ Bv control). The shape of the beta profile is identical to that shown in Fig. 1(c), while the amplitude is varied. w/ Bv FIG. 5. Magnetic surfaces calculated by the HINT2 code for the case B with (a) β0 = 0 %, Raxvac = 14.4 m (vacuum), (b) β0 ~ 7.5 %, Raxvac = 14.4 m (w/o Bv control), and (c) β0 ~ 8.5 %, Raxvac = 14.0 m (w/ Bv control). The shape of the beta profile is identical to that shown in Fig. 2(c), while the amplitude is varied. LHD. When the plasma position control by Bv is applied, as seen in FIG. 4(c), the destructed magnetic surfaces are reformed. The Shafranov shift is also mitigated by the Bv control, although the Shafranov shift of the magnetic axis is still large. A drastic effect of Bv control is obtained in the case B as shown in FIG. 5. Even in this case, the magnetic surfaces in the edge region become destructed at β0 ~ 7.5 % if Bv control is not applied (FIG. 5(b)). When the Bv control is applied (FIG. 5(c)), however, the magnetic surfaces similar to those in vacuum (FIG. 5(a)) are formed with a high central beta of β0 ~ 8.5 %. The magnetic surfaces in the edge region are reformed and the position of the magnetic axis is effectively pushed back to the initial position of Raxvac ~ 14.4 m MHD Stability In the LHD type heliotron, the inward-shifted configurations of, for example, Raxvac 3.60 m in LHD and Raxvac 14.4 m in FFHR-d1 are characterized by a better neoclassical transport and a worse MHD stability due to the magnetic hill property compared with those of outwardshifted configurations [3,10]. Since the equilibriums used in this study are the inward-shifted configurations, these are expected to be MHD unstable. Especially, the equilibriums shown in FIGs. 4(c) and 5(c) correspond to the strongly inward-shifted configuration of Raxvac = 3.50 m in LHD, where various kinds of MHD instability have been observed. Shown in FIG. 6 is the typical result of 3-D ideal linear MHD stability analysis using TERPSICHORE [13]. The

5 5 FTP/P7-34 equilibrium analyzed here is the case B with β 0 ~ 8.5 % and B v control (FIG. 5(c)). The positive Mercier index, D I, in the entire region means that the interchange modes can be unstable (FIG. 6(b)). The mode structure composed of m / n = (10 14) / 11 appears as the most unstable mode in the n = 1 mode family, where m and n are the poloidal and toroidal mode number, respectively. These modes are localized around ρ ~ 0.8, where the rational surface of ι / 2π = 1 exists. The consecutive poloidal mode numbers coupled by the toroidal effect show the ballooning-like structure. The low mode number instabilities, e.g., m / n = 1 / 1 or 2 / 1, do not appear as the most unstable mode. Therefore, the high mode number instabilities will become unstable in this case. In LHD, no serious confinement degradation due to the high mode number instabilities has been clearly observed yet. The impact of the high mode number instabilities as shown in FIG. 6 on the MHD stability and the energy confinement property of FFHR-d1 should be carefully considered in the future studies. The core resonant mode of m / n = 2 / 1 does not appear in spite of the large D I. If the rotational transform in the core region decreases below 0.5, then the m / n = 2 / 1 mode will become destabilized. It would be better to take into account the plasma current of, for example, the bootstrap current and/or the neutral-beam driven current, since it can affect the MHD stability by modifying the rotational transform profile. This is also remained for the future studies. In this study, we assume that MHD instability will not be problematic in FFHR-d1, based on the observation in LHD that the plasma can be generated and sustained even in the inward-shifted configurations of R ax vac 3.60 m, which are expected to be Mercier unstable [3,10] Neoclassical Transport FIG. 6. Radial profiles of (a) the rotational transform ι/2π, (b) Mercier index D I, (c) radial deviation ξ S and (d) the potential energy δw. Neoclassical thermal transport is calculated using radial profiles shown in FIGs. 1 and 2, and the MHD equilibriums shown in FIGs. 4 and 5. Radial profiles of integrated total neoclassical flux, Q tot neo = Q i neo + Q e neo, multiplied by the surface area, S, in various cases are shown in FIG. 7, where Q i neo and Q e neo are the neoclassical heat flux transported by ions and electrons, respectively. To calculate Q i neo in the cases with large Shafranov shift, the δf Monte Carlo simulation code FORTEC-3D [15] has been used to take into account the finite drift motion by following the exact guiding-center of ions, while Q i neo in vacuum and Q e neo in all cases are calculated by GSRAKE [14]. In FORTEC-3D, the deuteron plasma is assumed and consideration of deuteron-triton plasma is left for future studies (see Ref. 18 for more details). In the case A without B v control, where the Shafranov shift is large (FIG. 4(b)), the neoclassical heat flow is as large as ~3.5 GW at ρ ~ 0.6. It is reduced to ~2 GW when the B v control is applied. In the case A, however, the neoclassical heat flow cannot be reduced to less than ~1 GW, which is calculated using the vacuum magnetic configuration. This is twice

6 6 FTP/P7-34 larger than the expected P reactor (= P α P B ) in the case A of ~0.5 GW (see FIG. 1(e)). In the case B with B v control, on the other hand, the neoclassical transport is ~0.4 GW, which is similar to P reactor in the case B (see FIG. 2(e)). Therefore, in terms of energy balance, the case B can be a sustainable option, only if the anomalous transport and the direct loss of alpha particles are negligibly small. Estimation of the anomalous transport is left for the future study. The alpha particle loss is discussed in the next subsection Alpha Particle Confinement The birth and heat deposition profiles of alpha particles calculated by GNET [16] for the case A without B v control are shown in FIG. 8. The energy loss is ~40 % and the particle loss is ~60 % in this case. These results are not affected by the B v control. One of the reasons of this large energy loss can be attributed to the definition of the alpha FIG. 7. Radial profiles of the integrated total neoclassical heat flux. Closed circles, open circles, closed rhombuses, and open squares denote the case A without B v control, the case A with B v control, vacuum (R ax vac = 14.4 m), and the case B with B v control, respectively. particle loss boundary. In GNET, the alpha loss boundary is set at the last-closed-flux-surface (LCFS). However, alpha particles are not practically lost at LCFS. The alpha particles deviated from LCFS can reenter the confinement region. To examine the impact of the definition of loss boundary position, alpha orbit tracing by MORH [17] has been carried out, where the alpha particles loss boundary can be arbitrarily defined by using the magnetic data in the real space. The ratio of confined alpha particles started from various positions with FIG. 8. The birth (broken curve) and heat deposition (thick curve) profiles of alpha particles in FFHR-d1 calculated by GNET, where the MHD equilibrium shown in Fig. 4(b) is used. FIG. 9. The Ratio of confined alpha particles in the case A. Open and closed symbols denote the loss boundary set at LCFS and blanket, while squares and circles denote w/o and w/ B v control, respectively.

7 7 FTP/P7-34 (a) (b) FIG. 10. Distributions of hitting positions of lost alpha particles on the blanket for the case A, at different toroidal angles of (a) φ = 0 and (b) φ = 27. Four circles in each figure denote the divertor region. Crosses and plusses correspond to without and with B v control, respectively. various pitch angles is summarized in FIG. 9. These are calculated for the case A with and without B v control. It should be noted that the slowing down process is not taken into account in MORH, at this moment. Therefore, the ratio of confined alpha particles shown in FIG. 9 is not necessarily comparable with the deposition to the birth ratio as seen in FIG. 8. When the alpha loss boundary is set at LCFS, the ratio of confined alpha particles is larger than 60 % at ρ < 0.7, in both cases of w/ and w/o B v control. The B v control affects the ratio of confined alpha at ρ > 0.7. However, the number of alpha particles generated in this region is small and this presumably is the reason why no clear difference between w/ and w/o B v control is observed in GNET. When the alpha particles loss boundary is set at the blanket tentatively designed for FFHR-d1 (will be shown in FIG. 10), the ratio of confined alpha particles is improved to > 80 % at ρ < 0.7, even for the case without B v control. Further improvement is obtained when the B v control is applied. In this case, the ratio of confined alpha particles increases to larger than ~85 % at ρ < Hitting Positions of Alpha Particles To check the hitting positions of alpha particles on in-vessel components, the lost positions of alpha particles calculated by MORH are plotted on the slices of FFHR-d1 at different toroidal angles in FIG. 10. It should be noted that the divertors are placed behind the blanket in FFHRd1, and therefore the divertors are protected from the direct neutron irradiation. This is one of the strong merits of FFHR-d1. According to the MORH results, the majority of lost alpha particles reach the divertor region. A small portion of lost alpha particles started from the edge region of ρ ~ 0.95 will hit the sidewall of the blankets. However, the number of alpha particles generated at ρ ~ 0.95 is small since the temperature in this region is low (see FIGs. 1 and 2). Therefore, we expect that the damage of blanket sidewall will be small. It is necessary to properly design the divertor plates to receive the heat load of lost alpha particles, based on the results obtained here. Note that recalculation of alpha particle orbit is necessary if baffle plates are adopted to increase the neutral pressure in the divertor region.

8 8 FTP/P Summary The plasma performance in the helical DEMO reactor FFHR-d1 has been analyzed using the radial profiles extrapolated from LHD. Although the large Shafranov shift and resultant destruction of peripheral magnetic surfaces are expected in the high-beta reactor core plasma, it is possible to mitigate the Shafranov shift and restore the magnetic surfaces by selecting a high aspect ratio configuration and applying a proper vertical magnetic field. As long as the Shafranov shift is mitigated, the neoclassical thermal transport can be reduced to less than ~400 MW, which is comparable to the alpha heating power of ~500 MW in FFHR-d1. The energy loss of alpha particles is expected to be less than 20 %, as long as the loss boundary is set on the blanket surface and the proper plasma position control is applied. The lost alpha particles go to the divertor region, which are located behind the blanket in FFHR-d1. These results form the basis for the successive study of the reactor core plasma in FFHR-d1. It is necessary to carry out similar analyses iteratively in close link with engineering design. The MHD equilibrium, the neoclassical transport, and the alpha particle loss obtained in this study can be used as the input parameters for the next step analyses. The MHD stability analysis including the effect of bootstrap current on the MHD equilibrium and evaluation of the anomalous transport will be important as the next step analyses to assure the selfconsistency of the radial profiles in FFHR-d1. References [1] Sagara, A., et al., Fusion Eng. Des. 87 (2012) 594. [2] Goto, T., et al., Plasma Fusion Res. 7 (2012) [3] Komori, A., et al., Fusion Sci. Tech. 58 (2010) 1. [4] Miyazawa, J., et al., Plasma Fusion Res. 7 (2012) [5] Miyazawa, J., et al., Fusion Sci. Tech. 58 (2010) 29. [6] Stroth, U, et al, Nucl. Fusion 36 (1996) [7] Yamada, H., et al., Nucl. Fusion 45 (2005) [8] Yamada, I., et al., Fusion Sci. Tech. 58 (2010) 345. [9] Akiyama, T., et al., Fusion Sci. Tech. 58 (2010) 352. [10] Sakakibara, S., et al., Fusion Sci. Tech. 58 (2010) 176. [11] Suzuki, Y., et al., Nucl. Fusion 46 (2006) L19. [12] Hirshman, S.P., and Whitson, J.C., Phys. Fluids 26 (1983) [13] Cooper, W.A, Plasma Phys. Control. Fusion 34 (1992) [14] Beidler, C.D., et al., Plasma Phys. Control. Fusion 37 (1995) 463. [15] Satake, S., et al., Plasma Fusion Res. 3 (2008) S1062. [16] Murakami, S., et al. Nucl. Fusion 46 (2006) S425. [17] Seki, R., Plasma Fusion Res. 5 (2010) 027. [18] Satake, S., et al., 22 nd International Toki Conference (2012).

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

Japan-US Workshop on Fusion Power Plants Related Advanced Technologies with participants from China and Korea (Kyoto University, Uji, Japan, 26-28 Japan-US Workshop on Fusion Power Plants Related Advanced Technologies with participants from China and Korea (Kyoto University, Uji, Japan, 26-28 Feb. 2013) 2/22 FFHR-d1 R c = 15.6 m B c = 4.7 T P fusion

More information

Design window analysis of LHD-type Heliotron DEMO reactors

Design window analysis of LHD-type Heliotron DEMO reactors Design window analysis of LHD-type Heliotron DEMO reactors Fusion System Research Division, Department of Helical Plasma Research, National Institute for Fusion Science Takuya GOTO, Junichi MIYAZAWA, Teruya

More information

Configuration Optimization of a Planar-Axis Stellarator with a Reduced Shafranov Shift )

Configuration Optimization of a Planar-Axis Stellarator with a Reduced Shafranov Shift ) Configuration Optimization of a Planar-Axis Stellarator with a Reduced Shafranov Shift ) Shoichi OKAMURA 1,2) 1) National Institute for Fusion Science, Toki 509-5292, Japan 2) Department of Fusion Science,

More information

Estimations of Beam-Beam Fusion Reaction Rates in the Deuterium Plasma Experiment on LHD )

Estimations of Beam-Beam Fusion Reaction Rates in the Deuterium Plasma Experiment on LHD ) Estimations of Beam-Beam Fusion Reaction Rates in the Deuterium Plasma Experiment on LHD ) Masayuki HOMMA, Sadayoshi MURAKAMI, Hideo NUGA and Hiroyuki YAMAGUCHI Department of Nuclear Engineering, Kyoto

More information

DT Fusion Ignition of LHD-Type Helical Reactor by Joule Heating Associated with Magnetic Axis Shift )

DT Fusion Ignition of LHD-Type Helical Reactor by Joule Heating Associated with Magnetic Axis Shift ) DT Fusion Ignition of LHD-Type Helical Reactor by Joule Heating Associated with Magnetic Axis Shift ) Tsuguhiro WATANABE National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan (Received

More information

Ion Heating Experiments Using Perpendicular Neutral Beam Injection in the Large Helical Device

Ion Heating Experiments Using Perpendicular Neutral Beam Injection in the Large Helical Device Ion Heating Experiments Using Perpendicular Neutral Beam Injection in the Large Helical Device Kenichi NAGAOKA, Masayuki YOKOYAMA, Yasuhiko TAKEIRI, Katsumi IDA, Mikiro YOSHINUMA, Seikichi MATSUOKA 1),

More information

1 EX/P5-9 International Stellarator/Heliotron Database Activities on High-Beta Confinement and Operational Boundaries

1 EX/P5-9 International Stellarator/Heliotron Database Activities on High-Beta Confinement and Operational Boundaries 1 International Stellarator/Heliotron Database Activities on High-Beta Confinement and Operational Boundaries A. Weller 1), K.Y. Watanabe 2), S. Sakakibara 2), A. Dinklage 1), H. Funaba 2), J. Geiger 1),

More information

Analyses of Visible Images of the Plasma Periphery Observed with Tangentially Viewing CCD Cameras in the Large Helical Device

Analyses of Visible Images of the Plasma Periphery Observed with Tangentially Viewing CCD Cameras in the Large Helical Device Analyses of Visible Images of the Plasma Periphery Observed with Tangentially Viewing CCD Cameras in the Large Helical Device M. SHOJI, T. WATANABE, S. MASUZAKI, H. YAMADA, A. KOMORI and LHD Experimental

More information

Extension of High-Beta Plasma Operation to Low Collisional Regime

Extension of High-Beta Plasma Operation to Low Collisional Regime EX/4-4 Extension of High-Beta Plasma Operation to Low Collisional Regime Satoru Sakakibara On behalf of LHD Experiment Group National Institute for Fusion Science SOKENDAI (The Graduate University for

More information

Recent Development of LHD Experiment. O.Motojima for the LHD team National Institute for Fusion Science

Recent Development of LHD Experiment. O.Motojima for the LHD team National Institute for Fusion Science Recent Development of LHD Experiment O.Motojima for the LHD team National Institute for Fusion Science 4521 1 Primary goal of LHD project 1. Transport studies in sufficiently high n E T regime relevant

More information

- Effect of Stochastic Field and Resonant Magnetic Perturbation on Global MHD Fluctuation -

- Effect of Stochastic Field and Resonant Magnetic Perturbation on Global MHD Fluctuation - 15TH WORKSHOP ON MHD STABILITY CONTROL: "US-Japan Workshop on 3D Magnetic Field Effects in MHD Control" U. Wisconsin, Madison, Nov 15-17, 17, 2010 LHD experiments relevant to Tokamak MHD control - Effect

More information

Energetic Ion Confinement and Lost Ion Distribution in Heliotrons

Energetic Ion Confinement and Lost Ion Distribution in Heliotrons Energetic Ion Confinement and Lost Ion Distribution in Heliotrons S. Murakami, Y. Masaoka, T. Yamamot, A. Fukuyama, M. Osakabe1), K. Ida1), M. Yoshinuma1) T. Ozaki1), T. Tokuzawa1), M. Isobe1), M. Nishiura1),

More information

Design of structural components and radial-build for FFHR-d1

Design of structural components and radial-build for FFHR-d1 Japan-US Workshop on Fusion Power Plants and Related Advanced Technologies with participations from China and Korea February 26-28, 2013 at Kyoto University in Uji, JAPAN 1 Design of structural components

More information

Reduced-Size LHD-Type Fusion Reactor with D-Shaped Magnetic Surface )

Reduced-Size LHD-Type Fusion Reactor with D-Shaped Magnetic Surface ) Reduced-Size LHD-Type Fusion Reactor with D-Shaped Magnetic Surface ) Tsuguhiro WATANABE National Institute for Fusion Science, Toki 509-59, Japan (Received 6 December 011 / Accepted 1 June 01) A new winding

More information

Study of High-energy Ion Tail Formation with Second Harmonic ICRF Heating and NBI in LHD

Study of High-energy Ion Tail Formation with Second Harmonic ICRF Heating and NBI in LHD 21st IAEA Fusion Energy Conference Chengdu, China, 16-21 October, 2006 IAEA-CN-149/ Study of High-energy Ion Tail Formation with Second Harmonic ICRF Heating and NBI in LHD K. Saito et al. NIFS-851 Oct.

More information

Simulation of alpha particle current drive and heating in spherical tokamaks

Simulation of alpha particle current drive and heating in spherical tokamaks Simulation of alpha particle current drive and heating in spherical tokamaks R. Farengo 1, M. Zarco 1, H. E. Ferrari 1, 1 Centro Atómico Bariloche and Instituto Balseiro, Argentina. Consejo Nacional de

More information

Significance of MHD Effects in Stellarator Confinement

Significance of MHD Effects in Stellarator Confinement Significance of MHD Effects in Stellarator Confinement A. Weller 1, S. Sakakibara 2, K.Y. Watanabe 2, K. Toi 2, J. Geiger 1, M.C. Zarnstorff 3, S.R. Hudson 3, A. Reiman 3, A. Werner 1, C. Nührenberg 1,

More information

Finite-Orbit-Width Effect and the Radial Electric Field in Neoclassical Transport Phenomena

Finite-Orbit-Width Effect and the Radial Electric Field in Neoclassical Transport Phenomena 1 TH/P2-18 Finite-Orbit-Width Effect and the Radial Electric Field in Neoclassical Transport Phenomena S. Satake 1), M. Okamoto 1), N. Nakajima 1), H. Sugama 1), M. Yokoyama 1), and C. D. Beidler 2) 1)

More information

Comparison of Ion Internal Transport Barrier Formation between Hydrogen and Helium Dominated Plasmas )

Comparison of Ion Internal Transport Barrier Formation between Hydrogen and Helium Dominated Plasmas ) Comparison of Ion Internal Transport Barrier Formation between Hydrogen and Helium Dominated Plasmas ) Kenichi NAGAOKA 1,2), Hiromi TAKAHASHI 1,2), Kenji TANAKA 1), Masaki OSAKABE 1,2), Sadayoshi MURAKAMI

More information

Role of Low-Order Rational Surfaces in Transport Barrier Formation on the Large Helical Device

Role of Low-Order Rational Surfaces in Transport Barrier Formation on the Large Helical Device Role of Low-Order Rational Surfaces in Transport Barrier Formation on the Large Helical Device K. Toi, F. Watanabe a, K. Tanaka, T. Tokuzawa, K. Ogawa b, M. Isobe, M. Osakabe, Y. Suzuki, T. Akiyama, K.

More information

Integrated Particle Transport Simulation of NBI Plasmas in LHD )

Integrated Particle Transport Simulation of NBI Plasmas in LHD ) Integrated Particle Transport Simulation of NBI Plasmas in LHD Akira SAKAI, Sadayoshi MURAKAMI, Hiroyuki YAMAGUCHI, Arimitsu WAKASA, Atsushi FUKUYAMA, Kenichi NAGAOKA 1, Hiroyuki TAKAHASHI 1, Hirohisa

More information

Bunno, M.; Nakamura, Y.; Suzuki, Y. Matsunaga, G.; Tani, K. Citation Plasma Science and Technology (

Bunno, M.; Nakamura, Y.; Suzuki, Y. Matsunaga, G.; Tani, K. Citation Plasma Science and Technology ( Title The Finite Beta Effects on the Toro Tokamak Plasma Author(s) Bunno, M.; Nakamura, Y.; Suzuki, Y. Matsunaga, G.; Tani, K. Citation Plasma Science and Technology (2013 Issue Date 2013-02 URL http://hdl.handle.net/2433/173038

More information

Study on supporting structures of magnets and blankets for a heliotron-type type fusion reactors

Study on supporting structures of magnets and blankets for a heliotron-type type fusion reactors JA-US Workshop on Fusion Power Plants and Related Advanced Technologies with participation of EU, Jan. 11-13, 2005, Tokyo, Japan. Study on supporting structures of magnets and blankets for a heliotron-type

More information

Tokamak/Helical Configurations Related to LHD and CHS-qa

Tokamak/Helical Configurations Related to LHD and CHS-qa 9TH WORKSHOP ON MHD STABILITY CONTROL: "CONTROL OF MHD STABILITY: BACK TO THE BASICS" NOVEMBER 21-23, 2004, PRINCETON PLASMA PHYSICS LABORATORY Tokamak/Helical Configurations Related to LHD and CHS-qa

More information

Plasmoid Motion in Helical Plasmas

Plasmoid Motion in Helical Plasmas Plasmoid Motion in Helical Plasmas Ryuichi ISHIZAKI and Noriyoshi NAKAJIMA National Institute for Fusion Science, Toki 509-5292, Japan (Received 12 December 2009 / Accepted 18 May 2010) In order to explain

More information

19 th IAEA- Fusion Energy Conference FT/1-6

19 th IAEA- Fusion Energy Conference FT/1-6 1 19 th IAEA- Fusion Energy Conference FT/1-6 RECENT DEVELOPMENTS IN HELIAS REACTOR STUDIES H. Wobig 1), T. Andreeva 1), C.D. Beidler 1), E. Harmeyer 1), F. Herrnegger 1), Y. Igitkhanov 1), J. Kisslinger

More information

Der Stellarator Ein alternatives Einschlusskonzept für ein Fusionskraftwerk

Der Stellarator Ein alternatives Einschlusskonzept für ein Fusionskraftwerk Max-Planck-Institut für Plasmaphysik Der Stellarator Ein alternatives Einschlusskonzept für ein Fusionskraftwerk Robert Wolf robert.wolf@ipp.mpg.de www.ipp.mpg.de Contents Magnetic confinement The stellarator

More information

Integrated Heat Transport Simulation of High Ion Temperature Plasma of LHD

Integrated Heat Transport Simulation of High Ion Temperature Plasma of LHD 1 TH/P6-38 Integrated Heat Transport Simulation of High Ion Temperature Plasma of LHD S. Murakami 1, H. Yamaguchi 1, A. Sakai 1, K. Nagaoka 2, H. Takahashi 2, H. Nakano 2, M. Osakabe 2, K. Ida 2, M. Yoshinuma

More information

Effect of rotational transform and magnetic shear on confinement of stellarators

Effect of rotational transform and magnetic shear on confinement of stellarators Effect of rotational transform and magnetic shear on confinement of stellarators E. Ascasíbar(1), D. López-Bruna(1), F. Castejón(1), V.I. Vargas(1), V. Tribaldos(1), H. Maassberg(2), C.D. Beidler(2) R.

More information

MHD Simulation of High Wavenumber Ballooning-like Modes in LHD

MHD Simulation of High Wavenumber Ballooning-like Modes in LHD 1 TH/P9-16 MHD Simulation of High Wavenumber Ballooning-like Modes in LHD H. Miura and N. Nakajima National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, JAPAN e-mail contact of main

More information

Innovative fabrication method of superconducting magnets using high T c superconductors with joints

Innovative fabrication method of superconducting magnets using high T c superconductors with joints Innovative fabrication method of superconducting magnets using high T c superconductors with joints (for huge and/or complicated coils) Nagato YANAGI LHD & FFHR Group National Institute for Fusion Science,

More information

Dynamics of ion internal transport barrier in LHD heliotron and JT-60U tokamak plasmas

Dynamics of ion internal transport barrier in LHD heliotron and JT-60U tokamak plasmas Dynamics of ion internal transport barrier in LHD heliotron and JT-60U tokamak plasmas K. Ida 1), Y. Sakamoto 2), M. Yoshinuma 1), H. Takenaga 2), K. Nagaoka 1), N. Oyama 2), M. Osakabe 1), M. Yokoyama

More information

Plasma and Fusion Research: Regular Articles Volume 10, (2015)

Plasma and Fusion Research: Regular Articles Volume 10, (2015) Possibility of Quasi-Steady-State Operation of Low-Temperature LHD-Type Deuterium-Deuterium (DD) Reactor Using Impurity Hole Phenomena DD Reactor Controlled by Solid Boron Pellets ) Tsuguhiro WATANABE

More information

Extension of Wavelength Range in Absolute Intensity Calibration of Space-Resolved EUV Spectrometer for LHD Diagnostics )

Extension of Wavelength Range in Absolute Intensity Calibration of Space-Resolved EUV Spectrometer for LHD Diagnostics ) Extension of Wavelength Range in Absolute Intensity Calibration of Space-Resolved EUV Spectrometer for LHD Diagnostics ) Chunfeng DONG 1), Shigeru MORITA 1,2), Motoshi GOTO 1,2) and Erhui WANG 2) 1) National

More information

Physics of fusion power. Lecture 14: Anomalous transport / ITER

Physics of fusion power. Lecture 14: Anomalous transport / ITER Physics of fusion power Lecture 14: Anomalous transport / ITER Thursday.. Guest lecturer and international celebrity Dr. D. Gericke will give an overview of inertial confinement fusion.. Instabilities

More information

Design Windows and Economic Aspect of Helical Reactor

Design Windows and Economic Aspect of Helical Reactor Design Windows and Economic Aspect of Helical Reactor Y. Kozaki, S. Imagawa, A. Sagara National Institute for Fusion Science, Toki, Japan Japan-US Workshop, Kashiwa, March 16-18,2009 - Background and Objectives

More information

ISOTOPE EFFECTS ON CONFINEMENT AND TURBULENCE IN ECRH PLASMA OF LHD

ISOTOPE EFFECTS ON CONFINEMENT AND TURBULENCE IN ECRH PLASMA OF LHD K. TANAKA et al, ISOTOPE EFFECTS ON CONFINEMENT AND TURBULENCE IN ECRH PLASMA OF LHD K. TANAKA National Institute for Fusion Science, National Institutes on Natural Sciences Toki, Japan Email: ktanaka@nifs.ac.jp

More information

US-Japan workshop on Fusion Power Reactor Design and Related Advanced Technologies, March at UCSD.

US-Japan workshop on Fusion Power Reactor Design and Related Advanced Technologies, March at UCSD. US-Japan workshop on Fusion Power Reactor Design and Related Advanced Technologies, March 5-7 28 at UCSD. Overview Overview of of Design Design Integration Integration toward toward Optimization -type

More information

MHD. Jeff Freidberg MIT

MHD. Jeff Freidberg MIT MHD Jeff Freidberg MIT 1 What is MHD MHD stands for magnetohydrodynamics MHD is a simple, self-consistent fluid description of a fusion plasma Its main application involves the macroscopic equilibrium

More information

Simulation Study of Interaction between Energetic Ions and Alfvén Eigenmodes in LHD

Simulation Study of Interaction between Energetic Ions and Alfvén Eigenmodes in LHD 1 Simulation Study of Interaction between Energetic Ions and Alfvén Eigenmodes in LHD Y. Todo 1), N. Nakajima 1), M. Osakabe 1), S. Yamamoto 2), D. A. Spong 3) 1) National Institute for Fusion Science,

More information

High Beta Discharges with Hydrogen Storage Electrode Biasing in the Tohoku University Heliac

High Beta Discharges with Hydrogen Storage Electrode Biasing in the Tohoku University Heliac J. Plasma Fusion Res. SERIES, Vol. 8 (2009) High Beta Discharges with Hydrogen Storage Electrode Biasing in the Tohoku University Heliac Hiroyasu UTOH, Kiyohiko NISHIMURA 1), Hajime UMETSU, Keiichi ISHII,

More information

(a) (b) (c) (d) (e) (f) r (minor radius) time. time. Soft X-ray. T_e contours (ECE) r (minor radius) time time

(a) (b) (c) (d) (e) (f) r (minor radius) time. time. Soft X-ray. T_e contours (ECE) r (minor radius) time time Studies of Spherical Tori, Stellarators and Anisotropic Pressure with M3D 1 L.E. Sugiyama 1), W. Park 2), H.R. Strauss 3), S.R. Hudson 2), D. Stutman 4), X-Z. Tang 2) 1) Massachusetts Institute of Technology,

More information

STELLARATOR REACTOR OPTIMIZATION AND ASSESSMENT

STELLARATOR REACTOR OPTIMIZATION AND ASSESSMENT STELLARATOR REACTOR OPTIMIZATION AND ASSESSMENT J. F. Lyon, ORNL ARIES Meeting October 2-4, 2002 TOPICS Stellarator Reactor Optimization 0-D Spreadsheet Examples 1-D POPCON Examples 1-D Systems Optimization

More information

Introduction to Nuclear Fusion. Prof. Dr. Yong-Su Na

Introduction to Nuclear Fusion. Prof. Dr. Yong-Su Na Introduction to Nuclear Fusion Prof. Dr. Yong-Su Na What is a stellarator? M. Otthe, Stellarator: Experiments, IPP Summer School (2008) 2 Closed Magnetic System ion +++ ExB drift Electric field, E - -

More information

Effect of neoclassical poloidal viscosity and resonant magnetic perturbation on the response of the m/n=1/1 magnetic island in LHD

Effect of neoclassical poloidal viscosity and resonant magnetic perturbation on the response of the m/n=1/1 magnetic island in LHD Effect of neoclassical poloidal viscosity and resonant magnetic perturbation on the response of the m/n=1/1 magnetic island in LHD B. Huang 1, S. Satake 2, R. Kanno 2, Y. Narushima 2, S. Sakakibara 2,

More information

Plasma Physics Performance. Rebecca Cottrill Vincent Paglioni

Plasma Physics Performance. Rebecca Cottrill Vincent Paglioni Plasma Physics Performance Rebecca Cottrill Vincent Paglioni Objectives Ensure adequate plasma power and H-mode operation with a reasonable confinement time. Maintain plasma stability against various magnetohydrodynamic

More information

Is the Troyon limit a beta limit?

Is the Troyon limit a beta limit? Is the Troyon limit a beta limit? Pierre-Alexandre Gourdain 1 1 Extreme State Physics Laboratory, Department of Physics and Astronomy, University of Rochester, Rochester, NY 14627, USA The plasma beta,

More information

0 Magnetically Confined Plasma

0 Magnetically Confined Plasma 0 Magnetically Confined Plasma 0.1 Particle Motion in Prescribed Fields The equation of motion for species s (= e, i) is written as d v ( s m s dt = q s E + vs B). The motion in a constant magnetic field

More information

Integrated Simulation of ELM Energy Loss Determined by Pedestal MHD and SOL Transport

Integrated Simulation of ELM Energy Loss Determined by Pedestal MHD and SOL Transport 1 Integrated Simulation of ELM Energy Loss Determined by Pedestal MHD and SOL Transport N. Hayashi, T. Takizuka, T. Ozeki, N. Aiba, N. Oyama Japan Atomic Energy Agency, Naka, Ibaraki-ken, 311-0193 Japan

More information

High-Density, Low Temperature Ignited Operations in FFHR

High-Density, Low Temperature Ignited Operations in FFHR High-Density, Low Temperature Ignited Operations in FFHR Osamu MITARAI, Akio SAGARA 1), Ryuichi SAKAMOTO 1), Nobuyoshi OHYABU 1), Akio KOMORI 1) and Osamu MOTOJIMA 1) Liberal Arts Education Center, Kumamoto

More information

Active MHD Control Needs in Helical Configurations

Active MHD Control Needs in Helical Configurations Active MHD Control Needs in Helical Configurations M.C. Zarnstorff 1 Presented by E. Fredrickson 1 With thanks to A. Weller 2, J. Geiger 2, A. Reiman 1, and the W7-AS Team and NBI-Group. 1 Princeton Plasma

More information

1) H-mode in Helical Devices. 2) Construction status and scientific objectives of the Wendelstein 7-X stellarator

1) H-mode in Helical Devices. 2) Construction status and scientific objectives of the Wendelstein 7-X stellarator Max-Planck-Institut für Plasmaphysik 1) H-mode in Helical Devices M. Hirsch 1, T. Akiyama 2, T.Estrada 3, T. Mizuuchi 4, K. Toi 2, C. Hidalgo 3 1 Max-Planck-Institut für Plasmaphysik, EURATOM-Ass., D-17489

More information

Physics Considerations in the Design of NCSX *

Physics Considerations in the Design of NCSX * 1 IAEA-CN-94/IC-1 Physics Considerations in the Design of NCSX * G. H. Neilson 1, M. C. Zarnstorff 1, L. P. Ku 1, E. A. Lazarus 2, P. K. Mioduszewski 2, W. A. Cooper 3, M. Fenstermacher 4, E. Fredrickson

More information

Measurements of rotational transform due to noninductive toroidal current using motional Stark effect spectroscopy in the Large Helical Device

Measurements of rotational transform due to noninductive toroidal current using motional Stark effect spectroscopy in the Large Helical Device REVIEW OF SCIENTIFIC INSTRUMENTS 76, 053505 2005 Measurements of rotational transform due to noninductive toroidal current using motional Stark effect spectroscopy in the Large Helical Device K. Ida, a

More information

Design concept of near term DEMO reactor with high temperature blanket

Design concept of near term DEMO reactor with high temperature blanket 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

More information

Toroidal confinement devices

Toroidal confinement devices Toroidal confinement devices Dr Ben Dudson Department of Physics, University of York, Heslington, York YO10 5DD, UK 24 th January 2014 Dr Ben Dudson Magnetic Confinement Fusion (1 of 20) Last time... Power

More information

Initial Experimental Program Plan for HSX

Initial Experimental Program Plan for HSX Initial Experimental Program Plan for HSX D.T. Anderson, A F. Almagri, F.S.B. Anderson, J. Chen, S. Gerhardt, V. Sakaguchi, J. Shafii and J.N. Talmadge, UW-Madison HSX Plasma Laboratory Team The Helically

More information

Modelling of JT-60U Detached Divertor Plasma using SONIC code

Modelling of JT-60U Detached Divertor Plasma using SONIC code J. Plasma Fusion Res. SERIES, Vol. 9 (2010) Modelling of JT-60U Detached Divertor Plasma using SONIC code Kazuo HOSHINO, Katsuhiro SHIMIZU, Tomonori TAKIZUKA, Nobuyuki ASAKURA and Tomohide NAKANO Japan

More information

LHD-type heliotron reactor design

LHD-type heliotron reactor design Published by Fusion Energy Division, Oak Ridge National Laboratory Building 5700 P.O. Box 2008 Oak Ridge, TN 37831-6169, USA Editor: James A. Rome Issue 126 June 2010 E-Mail: jar@ornl.gov Phone (865) 482-5643

More information

Electrode Biasing Experiment in the Large Helical Device

Electrode Biasing Experiment in the Large Helical Device 1 EXC/P8-07 Electrode Biasing Experiment in the Large Helical Device S. Kitajima 1), H. Takahashi 2), K. Ishii 1), J. Sato 1), T. Ambo 1), M. Kanno 1), A. Okamoto 1), M. Sasao 1), S. Inagaki 3), M. Takayama

More information

ITER operation. Ben Dudson. 14 th March Department of Physics, University of York, Heslington, York YO10 5DD, UK

ITER operation. Ben Dudson. 14 th March Department of Physics, University of York, Heslington, York YO10 5DD, UK ITER operation Ben Dudson Department of Physics, University of York, Heslington, York YO10 5DD, UK 14 th March 2014 Ben Dudson Magnetic Confinement Fusion (1 of 18) ITER Some key statistics for ITER are:

More information

Observation of Neo-Classical Ion Pinch in the Electric Tokamak*

Observation of Neo-Classical Ion Pinch in the Electric Tokamak* 1 EX/P6-29 Observation of Neo-Classical Ion Pinch in the Electric Tokamak* R. J. Taylor, T. A. Carter, J.-L. Gauvreau, P.-A. Gourdain, A. Grossman, D. J. LaFonteese, D. C. Pace, L. W. Schmitz, A. E. White,

More information

Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor

Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor 1 FIP/3-4Ra Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor N. Asakura 1, K. Hoshino 1, H. Utoh 1, K. Shinya 2, K. Shimizu 3, S. Tokunaga 1, Y.Someya 1, K. Tobita 1, N. Ohno

More information

Optimization of Compact Stellarator Configuration as Fusion Devices Report on ARIES Research

Optimization of Compact Stellarator Configuration as Fusion Devices Report on ARIES Research Optimization of Compact Stellarator Configuration as Fusion Devices Report on ARIES Research Farrokh Najmabadi and the ARIES Team UC San Diego OFES Briefing November 16, 2005 Germantown Electronic copy:

More information

Shear Flow Generation in Stellarators - Configurational Variations

Shear Flow Generation in Stellarators - Configurational Variations Shear Flow Generation in Stellarators - Configurational Variations D. A. Spong 1), A. S. Ware 2), S. P. Hirshman 1), J. H. Harris 1), L. A. Berry 1) 1) Oak Ridge National Laboratory, Oak Ridge, Tennessee

More information

Energetic-Ion Driven Alfvén Eigenmodes in Large Helical Device Plasmas with Three-Dimensional Structure and Their Impact on Energetic Ion Transport

Energetic-Ion Driven Alfvén Eigenmodes in Large Helical Device Plasmas with Three-Dimensional Structure and Their Impact on Energetic Ion Transport Energetic-Ion Driven Alfvén Eigenmodes in Large Helical Device Plasmas with Three-Dimensional Structure and Their Impact on Energetic Ion Transport K. Toi, S. Yamamoto 1), N. Nakajima, S. Ohdachi, S. Sakakibara,

More information

Direct drive by cyclotron heating can explain spontaneous rotation in tokamaks

Direct drive by cyclotron heating can explain spontaneous rotation in tokamaks Direct drive by cyclotron heating can explain spontaneous rotation in tokamaks J. W. Van Dam and L.-J. Zheng Institute for Fusion Studies University of Texas at Austin 12th US-EU Transport Task Force Annual

More information

Impact of Energetic-Ion-Driven Global Modes on Toroidal Plasma Confinements

Impact of Energetic-Ion-Driven Global Modes on Toroidal Plasma Confinements Impact of Energetic-Ion-Driven Global Modes on Toroidal Plasma Confinements Kazuo TOI CHS & LHD Experimental Group National Institute for Fusion Science Toki 59-5292, Japan Special contributions from:

More information

Confinement Study of Net-Current Free Toroidal Plasmas Based on Extended International Stellarator Database

Confinement Study of Net-Current Free Toroidal Plasmas Based on Extended International Stellarator Database 1 X/1-5 Confinement Study of Net-Current Free Toroidal Plasmas Based on xtended International Stellarator Database H.Yamada 1), J.H.Harris 2), A.Dinklage 3),.Ascasibar 4), F.Sano 5), S.Okamura 1), J.Talmadge

More information

Stellarators. Dr Ben Dudson. 6 th February Department of Physics, University of York Heslington, York YO10 5DD, UK

Stellarators. Dr Ben Dudson. 6 th February Department of Physics, University of York Heslington, York YO10 5DD, UK Stellarators Dr Ben Dudson Department of Physics, University of York Heslington, York YO10 5DD, UK 6 th February 2014 Dr Ben Dudson Magnetic Confinement Fusion (1 of 23) Previously... Toroidal devices

More information

Characteristics of Internal Transport Barrier in JT-60U Reversed Shear Plasmas

Characteristics of Internal Transport Barrier in JT-60U Reversed Shear Plasmas Characteristics of Internal Transport Barrier in JT-6U Reversed Shear Plasmas Y. Sakamoto, Y. Kamada, S. Ide, T. Fujita, H. Shirai, T. Takizuka, Y. Koide, T. Fukuda, T. Oikawa, T. Suzuki, K. Shinohara,

More information

Reduction of Neoclassical Transport and Observation of a Fast Electron Driven Instability with Quasisymmetry in HSX

Reduction of Neoclassical Transport and Observation of a Fast Electron Driven Instability with Quasisymmetry in HSX 1 Reduction of Neoclassical Transport and Observation of a Fast Electron Driven Instability with Quasisymmetry in HSX J.M. Canik 1), D.L. Brower 2), C. Deng 2), D.T. Anderson 1), F.S.B. Anderson 1), A.F.

More information

OPTIMIZATION OF STELLARATOR REACTOR PARAMETERS

OPTIMIZATION OF STELLARATOR REACTOR PARAMETERS OPTIMIZATION OF STELLARATOR REACTOR PARAMETERS J. F. Lyon, L.P. Ku 2, P. Garabedian, L. El-Guebaly 4, L. Bromberg 5, and the ARIES Team Oak Ridge National Laboratory, Oak Ridge, TN, lyonjf@ornl.gov 2 Princeton

More information

arxiv: v1 [physics.plasm-ph] 24 Nov 2017

arxiv: v1 [physics.plasm-ph] 24 Nov 2017 arxiv:1711.09043v1 [physics.plasm-ph] 24 Nov 2017 Evaluation of ideal MHD mode stability of CFETR baseline scenario Debabrata Banerjee CAS Key Laboratory of Geospace Environment and Department of Modern

More information

Research of Basic Plasma Physics Toward Nuclear Fusion in LHD

Research of Basic Plasma Physics Toward Nuclear Fusion in LHD Research of Basic Plasma Physics Toward Nuclear Fusion in LHD Akio KOMORI and LHD experiment group National Institute for Fusion Science, Toki, Gifu 509-5292, Japan (Received 4 January 2010 / Accepted

More information

Integrated Modelling and Simulation of Toroidal Plasmas

Integrated Modelling and Simulation of Toroidal Plasmas 7th ITER International School on High performance computing in fusion science Aix-Marseille University, Aix-en-Provence, France 2014-08-28 Integrated Modelling and Simulation of Toroidal Plasmas Atsushi

More information

Low Beta MHD Equilibrium Including a Static Magnetic Island for Reduced MHD Equations in a Straight Heliotron Configuration

Low Beta MHD Equilibrium Including a Static Magnetic Island for Reduced MHD Equations in a Straight Heliotron Configuration Low Beta MHD Equilibrium Including a Static Magnetic Island for Reduced MHD Equations in a Straight Heliotron Configuration Kinya SAITO 1,a), Katsuji ICHIGUCHI 1,2) and Ryuichi ISHIZAKI 1,2) 1) The Graduate

More information

Characterization of neo-classical tearing modes in high-performance I- mode plasmas with ICRF mode conversion flow drive on Alcator C-Mod

Characterization of neo-classical tearing modes in high-performance I- mode plasmas with ICRF mode conversion flow drive on Alcator C-Mod 1 EX/P4-22 Characterization of neo-classical tearing modes in high-performance I- mode plasmas with ICRF mode conversion flow drive on Alcator C-Mod Y. Lin, R.S. Granetz, A.E. Hubbard, M.L. Reinke, J.E.

More information

- Plasma Control - Stellarator-Heliotron Control

- Plasma Control - Stellarator-Heliotron Control - Plasma Control - Stellarator-Heliotron Control H.Yamada National Institute for Fusion Science, NINS The University of Tokyo Acknowledgements to T.Akiyama, A.Dinklage, T.Goto, M.Koyabashi, J.Miyazawa,

More information

Additional Heating Experiments of FRC Plasma

Additional Heating Experiments of FRC Plasma Additional Heating Experiments of FRC Plasma S. Okada, T. Asai, F. Kodera, K. Kitano, T. Suzuki, K. Yamanaka, T. Kanki, M. Inomoto, S. Yoshimura, M. Okubo, S. Sugimoto, S. Ohi, S. Goto, Plasma Physics

More information

STABILIZATION OF m=2/n=1 TEARING MODES BY ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII D TOKAMAK

STABILIZATION OF m=2/n=1 TEARING MODES BY ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII D TOKAMAK GA A24738 STABILIZATION OF m=2/n=1 TEARING MODES BY ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII D TOKAMAK by T.C. LUCE, C.C. PETTY, D.A. HUMPHREYS, R.J. LA HAYE, and R. PRATER JULY 24 DISCLAIMER This

More information

Confinement Study of Net-Current Free Toroidal Plasmas Based on Extended International Stellarator Database

Confinement Study of Net-Current Free Toroidal Plasmas Based on Extended International Stellarator Database EX/1-5 IAEA FEC24, Vilamoura, Nov.2, 24 Confinement Study of Net-Current Free Toroidal Plasmas Based on Extended International Stellarator Database H.Yamada 1), J.H.Harris 2), A.Dinklage 3), E.Ascasibar

More information

Overview of Pilot Plant Studies

Overview of Pilot Plant Studies Overview of Pilot Plant Studies and contributions to FNST Jon Menard, Rich Hawryluk, Hutch Neilson, Stewart Prager, Mike Zarnstorff Princeton Plasma Physics Laboratory Fusion Nuclear Science and Technology

More information

Integrated Simulation of ELM Energy Loss and Cycle in Improved H-mode Plasmas

Integrated Simulation of ELM Energy Loss and Cycle in Improved H-mode Plasmas 1 Integrated Simulation of ELM Energy Loss and Cycle in Improved H-mode Plasmas N. Hayashi 1), T. Takizuka 1), N. Aiba 1), N. Oyama 1), T. Ozeki 1), S. Wiesen 2), V. Parail 3) 1) Japan Atomic Energy Agency,

More information

Introduction to Fusion Physics

Introduction to Fusion Physics Introduction to Fusion Physics Hartmut Zohm Max-Planck-Institut für Plasmaphysik 85748 Garching DPG Advanced Physics School The Physics of ITER Bad Honnef, 22.09.2014 Energy from nuclear fusion Reduction

More information

Advanced Tokamak Research in JT-60U and JT-60SA

Advanced Tokamak Research in JT-60U and JT-60SA I-07 Advanced Tokamak Research in and JT-60SA A. Isayama for the JT-60 team 18th International Toki Conference (ITC18) December 9-12, 2008 Ceratopia Toki, Toki Gifu JAPAN Contents Advanced tokamak development

More information

Nonlinear Simulation of Energetic Particle Modes in JT-60U

Nonlinear Simulation of Energetic Particle Modes in JT-60U TH/P6-7 Nonlinear Simulation of Energetic Particle Modes in JT-6U A.Bierwage,N.Aiba 2, K.Shinohara 2, Y.Todo 3,W.Deng 4,M.Ishikawa 2,G.Matsunaga 2 and M. Yagi Japan Atomic Energy Agency (JAEA), Rokkasho,

More information

Possibilities for Long Pulse Ignited Tokamak Experiments Using Resistive Magnets

Possibilities for Long Pulse Ignited Tokamak Experiments Using Resistive Magnets PFC/JA-91-5 Possibilities for Long Pulse Ignited Tokamak Experiments Using Resistive Magnets E. A. Chaniotakis L. Bromberg D. R. Cohn April 25, 1991 Plasma Fusion Center Massachusetts Institute of Technology

More information

GA A23168 TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO

GA A23168 TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO GA A23168 TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO by C.P.C. WONG and R.D. STAMBAUGH JULY 1999 DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United

More information

Recent Fusion Research in the National Institute for Fusion Science )

Recent Fusion Research in the National Institute for Fusion Science ) Recent Fusion Research in the National Institute for Fusion Science ) Akio KOMORI, Satoru SAKAKIBARA, Akio SAGARA, Ritoku HORIUCHI, Hiroshi YAMADA, Yasuhiko TAKEIRI and NIFS Team National Institute for

More information

Density Collapse in Improved Confinement Mode on Tohoku University Heliac

Density Collapse in Improved Confinement Mode on Tohoku University Heliac 1 EX/P5-12 Density Collapse in Improved Confinement Mode on Tohoku University Heliac S. Kitajima 1), Y. Tanaka 2), H. Utoh 1), H. Umetsu 1), J. Sato 1), K. Ishii 1), T. Kobuchi 1), A. Okamoto 1), M. Sasao

More information

Simulation Study of High-Frequency Magnetosonic Waves Excited by Energetic Ions in Association with Ion Cyclotron Emission )

Simulation Study of High-Frequency Magnetosonic Waves Excited by Energetic Ions in Association with Ion Cyclotron Emission ) Simulation Study of High-Frequency Magnetosonic Waves Excited by Energetic Ions in Association with Ion Cyclotron Emission ) Mieko TOIDA 1),KenjiSAITO 1), Hiroe IGAMI 1), Tsuyoshi AKIYAMA 1,2), Shuji KAMIO

More information

Issues of Perpendicular Conductivity and Electric Fields in Fusion Devices

Issues of Perpendicular Conductivity and Electric Fields in Fusion Devices Issues of Perpendicular Conductivity and Electric Fields in Fusion Devices Michael Tendler, Alfven Laboratory, Royal Institute of Technology, Stockholm, Sweden Plasma Turbulence Turbulence can be regarded

More information

The Path to Fusion Energy creating a star on earth. S. Prager Princeton Plasma Physics Laboratory

The Path to Fusion Energy creating a star on earth. S. Prager Princeton Plasma Physics Laboratory The Path to Fusion Energy creating a star on earth S. Prager Princeton Plasma Physics Laboratory The need for fusion energy is strong and enduring Carbon production (Gton) And the need is time urgent Goal

More information

Innovative Concepts Workshop Austin, Texas February 13-15, 2006

Innovative Concepts Workshop Austin, Texas February 13-15, 2006 Don Spong Oak Ridge National Laboratory Acknowledgements: Jeff Harris, Hideo Sugama, Shin Nishimura, Andrew Ware, Steve Hirshman, Wayne Houlberg, Jim Lyon Innovative Concepts Workshop Austin, Texas February

More information

Quasi-Symmetric Stellarators as a Strategic Element in the US Fusion Energy Research Plan

Quasi-Symmetric Stellarators as a Strategic Element in the US Fusion Energy Research Plan Quasi-Symmetric Stellarators as a Strategic Element in the US Fusion Energy Research Plan Quasi-Symmetric Stellarator Research The stellarator offers ready solutions to critical challenges for toroidal

More information

Diagnostics for Burning Plasma Physics Studies: A Status Report.

Diagnostics for Burning Plasma Physics Studies: A Status Report. Diagnostics for Burning Plasma Physics Studies: A Status Report. Kenneth M. Young Princeton Plasma Physics Laboratory UFA Workshop on Burning Plasma Science December 11-13 Austin, TX Aspects of Plasma

More information

ELM Suppression in DIII-D Hybrid Plasmas Using n=3 Resonant Magnetic Perturbations

ELM Suppression in DIII-D Hybrid Plasmas Using n=3 Resonant Magnetic Perturbations 1 EXC/P5-02 ELM Suppression in DIII-D Hybrid Plasmas Using n=3 Resonant Magnetic Perturbations B. Hudson 1, T.E. Evans 2, T.H. Osborne 2, C.C. Petty 2, and P.B. Snyder 2 1 Oak Ridge Institute for Science

More information

Progress of Confinement Physics Study in Compact Helical System

Progress of Confinement Physics Study in Compact Helical System 1st IAEA Fusion Energy Conference Chengdu, China, 16-1 October, 6 IAEA-CN-149/ EX/5-5Rb Progress of Confinement Physics Study in Compact Helical System S. Okamura et al. NIFS-839 Oct. 6 1 EX/5-5Rb Progress

More information

Development of a Systematic, Self-consistent Algorithm for the K-DEMO Steady-state Operation Scenario

Development of a Systematic, Self-consistent Algorithm for the K-DEMO Steady-state Operation Scenario Development of a Systematic, Self-consistent Algorithm for the K-DEMO Steady-state Operation Scenario J.S. Kang 1, J.M. Park 2, L. Jung 3, S.K. Kim 1, J. Wang 1, D. H. Na 1, C.-S. Byun 1, Y. S. Na 1, and

More information