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Naitou H.,Yamaguchi University | Yamada Y.,Yamaguchi University | Kajiwara K.,Yamaguchi University | Lee W.-L.,Princeton Plasma Physics Laboratory | And 3 more authors.
Plasma Science and Technology | Year: 2011

In order to implement large-scale and high-beta tokamak simulation, a new algorithm of the electromagnetic gyrokinetic PIC (particle-in-cell) code was proposed and installed on the Gpic-MHD code [Gyrokinetic PIC code for magnetohydrodynamic (MHD) simulation]. In the new algorithm, the vorticity equation and the generalized Ohm's law along the magnetic field are derived from the basic equations of the gyrokinetic Vlasov, Poisson, and Ampere system and are used to describe the spatio-temporal evolution of the field quantities of the electrostatic potential and the longitudinal component of the vector potential Az. The basic algorithm is equivalent to solving the reduced-MHD-type equations with kinetic corrections, in which MHD physics related to Alfven modes are well described. The estimation of perturbed electron pressure from particle dynamics is dominant, while the effects of other moments are negligible. Another advantage of the algorithm is that the longitudinal induced electric field, ETz = -∂Az/∂t, is explicitly estimated by the generalized Ohm's law and used in the equations of motion. Furthermore, the particle velocities along the magnetic field are used (vz-formulation) instead of generalized momentums (p z-formulation), hence there is no problem of 'cancellation', which would otherwise appear when Az is estimated from the Ampere's law in the pz-formulation. The successful simulation of the collisionless internal kink mode by the new Gpic-MHD with realistic values of the large-scale and high-beta tokamaks revealed the usefulness of the new algorithm.


Tobita K.,Japan Atomic Energy Agency | Federici G.,Fusion for Energy | Okano K.,International Fusion Energy Research Center
Fusion Engineering and Design | Year: 2014

The main objective of DEMO design activity under the Broader Approach is to develop pre-conceptual design of DEMO options by addressing key design issues on physics, technology and system engineering. This paper describes the latest results of the design activity, including DEMO parameter study, divertor and remote maintenance. DEMO parameter study has recently started with "pulsed" DEMO having a major radius (Rp) of 9 m, and "steady state" DEMO of Rp = 8.2 m or more. Divertor design study has focused on a computer simulation of fully detached plasma under DEMO divertor conditions and the assessment of advanced divertor configuration such as super-X. Comparative study of various maintenance schemes for DEMO and narrowing down the schemes is in progress. © 2014 Elsevier B.V.


Utoh H.,Japan Atomic Energy Agency | Utoh H.,International Fusion Energy Research Center | Takase H.,Japan Atomic Energy Agency | Takase H.,International Fusion Energy Research Center | And 11 more authors.
Fusion Engineering and Design | Year: 2016

In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell. © 2015 Elsevier B.V. All rights reserved.


Klepper C.C.,Oak Ridge National Laboratory | Biewer T.M.,Oak Ridge National Laboratory | Graves V.B.,Oak Ridge National Laboratory | Andrew P.,ITER Organization | And 11 more authors.
Fusion Engineering and Design | Year: 2015

One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (mainly in the form of diatomic molecules of H, D, and T). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N2), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (~8m long, ~110mm diameter) sampling pipe originating from a pressure reducing orifice, confirm that the desired response time (~1s for He or D2) is achieved with the present design. © 2015.


Rivas J.C.,International Fusion Energy Research Center | Rivas J.C.,Polytechnic University of Catalonia | Nakamura M.,Japan Atomic Energy Agency | Someya Y.,Japan Atomic Energy Agency | And 10 more authors.
Fusion Engineering and Design | Year: 2015

In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times.The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case the transient time until melting is nearly two orders of magnitude higher. The results point out that the recovery time for plasma control system should be at least one order of magnitude lower than confinement time to avoid melting of divertor targets. © 2015 Elsevier B.V.


Takase H.,International Fusion Energy Research Center | Takase H.,Japan Atomic Energy Agency | Utoh H.,International Fusion Energy Research Center | Utoh H.,Japan Atomic Energy Agency | And 6 more authors.
Fusion Engineering and Design | Year: 2015

Plasma position control for DEMO reactor has been investigated using numerical simulation, which consists of plasma equilibrium, eddy current and active feedback control analyses. The stabilization effect of in-vessel components, the influence on the magnetic detector and the power of active feedback control coils are evaluated. Especially, the influence of breeding blanket modules on plasma position control is shown in this paper. © 2015 Elsevier B.V.

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