A new technique and results are presented for the estimation of the open loop frequency response of the plasma on TCV. Voltages were applied to poloidal field coils and the resulting plasma current, position and shape related parameters were measured. The results are compared with the CREATE-L model, and good agreement is confirmed. The results are a significant advance on previous comparisons with closed loop data, which were limited by the role of feedback in the system. A simpler circuit equation model has also been developed in order to understand the reasons for the good agreement and identify which plasma properties are important in determining the response. The reasons for the good agreement with this model are discussed. An alternative modelling method has been developed, combining features of both the theoretical and experimental techniques. Its advantage is that it incorporates well defined knowledge of the electromagnetic properties of the tokamak with experimental data to derive plasma related parameters. This new model provides further insight into the plasma behaviour.
Results of 2D disruption modelling for validation of benchmark ITER scenarios using two established codes—DINA and TSC, are compared. Although the simulation models employed in those two codes ought to be equivalent in the resistive time scale, quite different defining equations and formulations are adopted in their approaches. Moreover there are considerable differences in the implemented model of solid conducting structures placed on the periphery of the plasma such as the vacuum vessel and blanket modules. Thus it has long been unanswered whether the one of the two codes is really able to reproduce the other's results correctly, since a large number of code-wise differences render the comparison task exceedingly complicated. In this paper, it is demonstrated that after the simulations are set up accounting for the model differences, a reasonably good agreement is generally obtained, corroborating the correctness of the code results. When the halo current generation and its poloidal path in the first wall are included, however, the situation is more complicated. Because of the surface averaged treatment of the magnetic field (current density) diffusion equation, DINA can only approximately handle the poloidal electric currents in the first wall that cross the field lines. Validation is carried out for DINA simulations of the halo current generation by comparing with TSC simulations, where the treatment of halo current dynamics is more justifiable. The specific details of each code, affecting the consequence in ITER disruption prediction, are highlighted and discussed.
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.
DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.
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