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Oikawa, Toshihiro; Tani, Keiji*; Azumi, Masafumi*
no journal, ,
An asymmetry particle trapping caused by the finite banana width of alpha particles and a current produced by the resulting imbalance between transit particles with positive and negative velocities parallel to the magnetic field line (simply referred to as transit current) are numerically demonstrated by using an orbit-following Monte-Carlo code. The transit current is considerably enhanced by a collaborative effect of the finite-banana-width and a non-uniformity of the alpha particle source. Banana-trapped particles drive a substantially negative current in the central region. Consequently, the positive transit current is degraded by the negative banana current and the net alpha current is almost flat near the plasma centre. The net alpha-driven total currents in an ITER-like normal aspect-ratio system (aspect ratio = 3.17) and in a low aspect-ratio system VECTOR (aspect ratio = 1.98) estimated with a conventional electron screening effect are about 0.4 MA and 1 MA, respectively.
Oikawa, Toshihiro
no journal, ,
The author has been leading a neutral beam current drive code benchmark in the ITPA IOS topical group frame. Presently, OFMC, NUBEAM, SPOT, ACCOME and ASTRA have been compared. The NB fast ion source profile agrees well among the codes. The heating profile generally agrees, but there is visible discrepancy. This is considerted to result from finite orbit width effects. Large differenence is observed in the NBCD profile. OFMC and ACCOME calculates the initial pitch angle of a fast ion against the toroidal magnetic field, which should correctly be against the equilibrium field. By correcting this the difference is expected to be resolved to some degree. ASTRA's deviation from other codes results from its too simplified NBI geometry and a Fokker-Planck equation derived for a cylinder plasma being used, which does not accordingly include toroidal effects. Other codes such as CQL3D, ASCOT and TRANSP are going to submit results soon. The author will discuss on how to proceed the benchmark.
Honda, Mitsuru; Hayashi, Nobuhiko
no journal, ,
We will report the status of the integrated modeling activities in JAEA and latest simulation results in ITER by using an integrated code TOPICS-IB. For integrated modelings, we have integrated a neutral transport code with TOPICS-IB so as to study the energy loss and the cycle of ELMs with density dynamics. By using TASK/TX with OFMC and EPOC, we have reproduced the toroidal rotation due to the radial current induced by the charge separation in a plasma with NBI and clarified its physical mechanism. We briefly explain the present status of the integrated divertor code SONIC. For steady-state simulations in ITER, we confirmed the validity of the integration of TOPICS-IB with an LH code, comparing the simulation result with that by ACCOME. Furthermore, we compared the neoclassical heat diffusivity by the Matrix Inversion Code with that by NCLASS because a performance of an ITER plasma in a steady state is found to depend on it.
Miyamoto, Seiji; Nakamura, Yukiharu*; Fujieda, Hirobumi; Hamamatsu, Kiyotaka; Oikawa, Toshihiro; Sugie, Tatsuo; Kusama, Yoshinori; Yoshino, Ryuji
no journal, ,
Recently, we developed a simulation model in which an ECRF ray tracing and current drive calculations are combined with TSC. This model is applied to the evaluation of magnetic flux consumption in the ITER current ramp-up scenario. In this model, real geometry of PF/CS coils and EC launcher is taken into account, and EC deposition/current drive profile are calculated in self-consistent with the plasma profile evolution. Central current drive (present ITER design) and off-axis current drive (test case) is compared. Resistive flux is lowered in both cases. Internal flux is also reduced by the off-axis EC due to reduction of internal inductance. It is although shown that, even in the case of central EC, comparable reduction of internal flux is expected due to the skin effect of inductive current.
Nakamura, Yukiharu*; Miyamoto, Seiji; Fujieda, Hirobumi; Oikawa, Toshihiro; Sugie, Tatsuo; Kusama, Yoshinori; Yoshino, Ryuji
no journal, ,
In tokamak fusion reactors with superconducting PF coil system, ramping-up of the plasma current for startup of discharges is essentially restrained at a rate much slower than the current tokamak with normal PF conductors. Therefore, the induced plasma current can penetrate deeply into the core region with higher electron temperature, i.e. low Ohmic resistivity, leading to a centrally peaked current profile. Consequently, such the high li, highly elongated configuration imperative to divertor formation has the dangerous property of causing vertical instability. Additional heating at early phase of the ramp-up has been addressed as one of startup procedures that retard the penetration of plasma current. A potential drawback, however, is the substantial increase of direct heat load to limiter. Hence, it follows that discharge optimization, e.g. timing of the heating, has been critical to plasma formation during the startup phase. The feasible ramp-up scenario to obtain a stable current profile while increasing bootstrap current fraction is discussed, as well as the interests: how quickly the plasma current needs to be ramped up and how slowly it can be.
Ide, Shunsuke; JT-60 Team
no journal, ,
Results of international joint experiments on JT-60U that have been proposed by the Steady state operation ITPA topical group are reported. The results are mainly on plasma development experiments towards establishment of operation scenarios on ITER. Research structure after closure of the JT-60U experiments is also reported.