Observations of fast ion loss to the plasma facing wall during quiescent H-modes on DIII-D

The Quiescent H-mode exhibits H-mode levels of confinement and edge pedestal pressures, but does not exhibit ELMs. To date this mode has only been observed in tokamaks during beam heating with some or all of the beams injected counter to the direction of plasma current. During QH-mode, fast ion loss to the low field side plasma facing surfaces has been observed. Some of the fast ion loss is calculated to be the result of outwardly directed banana orbits of the energetic beam ions created in the edge region. Other fast ion loss has been observed to be associated with bursts or oscillations in broadband, high-frequency, magnetic fluctuations. The relationship of the fast ion loss to the ELM stabilization or edge particle transport during QH-mode is not yet understood. © 2004 Elsevier B.V. All rights reserved

Non-linear MHD modelling of ELM triggering by pellet injection in DIII-D and implications for ITER

Edge localized mode (ELM) triggering by pellet injection in the DIII-D tokamak has been simulated with the non-linear MHD code JOREK with a view to validating its physics models. JOREK has been subsequently applied to evaluate the requirements for ELM control by pellet injection in ITER. JOREK modelling results for DIII-D show that the key parameter for the triggering of ELMs by pellets is the value of the localized pressure perturbation caused by pellet injection which leads to a threshold minimum pellet size for a given injection velocity, injection geometry and H-mode plasma characteristics. The minimum pellet size for ELM triggering is found to depend on injection geometry with the largest value being required for injection at the outer midplane, intermediate for injection near the X-point and the smallest one for injection at the high-field side. The first results of studies for ELM triggering by pellet injection in ITER 15 MA Q = 10 plasmas with the foreseen injection geometry in ITER are presented

Initial development of the DIII-D snowflake divertor control

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII-D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII-D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad-Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII-D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5× reduction in the peak heat flux for many energy confinement times (2-3 s) without any adverse effects on core plasma performance

Advanced tokamak physics in DIII-D

Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-β advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D

Long pulse high performance discharges in the DIII-D tokamak

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5> with q > 1.5. (The normalized performance is measured by the product β H , indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes as previously observed. Confinement remains good despite q > 1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ≈75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with β H ≈ 7 for up to 6.3 s or ≈34τ . These discharges appear to have stationary current profiles with q ≈ 1.05 in agreement with the current profile relaxation time ≈1.8 s. E min N 89 min N 89 E mi

Advanced tokamak physics in DIII-D

Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-β advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper