DIII-D Program Research Articles

Applications of ECH on the DIII-D tokamak and projections for future ECH upgrades

Electron Cyclotron Heating and Current Drive plays an important role in the DIII-D program. In high performance discharges EC power contributes greatly to MHD stability, and this is particularly important for discharges with low rotational torque applied, as will be the case for ITER. Off-axis EC current drive also plays a key role in the actualization of steady-state scenarios by supporting the desired current profile. In order to carry out these applications at higher beta and higher field, an upgrade of the EC power to 15 MW is needed, and the best gyrotron frequency for the DIII-D program is 117.5 GHz.

DIII-D contributions towards the scientific basis for sustained burning plasmas

DIII-D is making significant contributions to a scientific basis for sustained burning plasma operation. These include explorations of increasingly reactor-relevant scenarios, studies of key issues for projecting performance, development of techniques for handling heat and particle efflux, and assessment of key issues for the ITER research plan. Advanced scenarios are being optimized in DIII-D via experiments to empirically determine the relationship between transport and the current profile, which in turn can provide essential input to inform improvement of the theory-based models that do not currently capture the observed behaviour. Joint DIII-D/JET ρ* scans in the hybrid regime imply Bohm-like confinement scaling. Startup and shutdown techniques were developed for the restrictive environment of future devices while retaining compatibility with advanced scenarios. Towards the goal of a fully predictive capability, the DIII-D program emphasizes validation of physics-based models, facilitated by a number of new and upgraded diagnostics. Specific areas include transport, rotation, energetic particles and the H-mode pedestal, but this approach permeates the entire research programme. Concerns for heat and particle efflux in future devices are addressed through studies of ELM control, disruption avoidance and mitigation, and hydrogenic retention in DIII-D's carbon wall. DIII-D continues to respond to specific needs for ITER. Recent studies have compared H-mode access in several different ion species, identifying not only isotopic, but density, rotation and geometrical dependences that may guide access to H-mode during ITER's non-activated early operation. DIII-D used an insertable module to simulate the magnetic perturbations introduced by one of ITER's three test blanket module sets, demonstrating that little impact on performance is seen at ITER equivalent levels of magnetic perturbation.

06/02601 Overview of the DIII-D program and constructionPetersen, P. I. et al. Fusion Engineering and Design, 2005, 74, (1–4), 171–177

This paper presents a method for the dimensioning of the low-energy District Heating (DH) piping networks operating with a control philosophy of supplying heat in low-temperature such as 55 °C in supply and 25 °C in return regularly while the supply temperature levels are being boosted in cold winter periods. The performance of the existing radiators that were formerly sized with over-dimensions was analyzed, its results being used as input data for the performance evaluation of the piping network of the low-energy DH system operating with the control philosophy in question. The optimization method was performed under different mass flow limitations that were formed with various temperature configurations. The results showed that reduction in the mass flow rate requirement of a district is possible by increasing the supply temperature in cold periods with significant reduction in heat loss from the DH network. Sensitivity analysis was carried out in order to evaluate the area of applicability of the proposed method. Hence varied values of the original capacity and the current capacity of the existing radiators were evaluated with the design temperature values that were defined by two former radiator sizing standards.

Development of Next-Step-Relevant ICRF Technology in the DIII-D Program

High-power experiments with waves in the ion cyclotron range of frequencies (ICRF) have been performed on the DIII-D tokamak since 1988. In the course of these experiments, it has been necessary to carry out some development of the relevant technology. These developments, primarily in the areas of the plasma/antenna interface, high-power transmission line components, and transmission line configurations, are reviewed in this paper. They have led to a number of recommendations for specifying systems in next-generation tokamaks.

Overview of the DIII-D Fusion Science Program

This overview of the DIII-D fusion program provides an introduction to the research program carried out on the DIII-D tokamak since its inception in 1986. It serves as the introduction and summary of this special issue of Fusion Science and Technology on the DIII-D program. Special emphasis is given to the contributions of the program to the world fusion energy program and progress toward a burning plasma.

Development of Steady-State Advanced Tokamak Research in the DIII-D Tokamak

Research into the feasibility of steady-state operation of high-fusion-gain tokamak plasmas is one of the central elements of the DIII-D program. Realization of such discharges has progressed to the point of demonstrating well-aligned noninductive current profiles for a resistive time at 90% of the total current with plasma pressure and confinement consistent with fusion gain >5 in an ITER-sized tokamak. Full current drive discharges with poorer alignment have been obtained for shorter duration. The design methodology and the path to integrating the various elements necessary for full noninductive operation on DIII-D are discussed in detail.

DIII-D integrated plasma control tools applied to next generation tokamaks

A complete software suite for integrated tokamak plasma control has been developed within the DIII-D program. The suite consists of software for real-time control of all aspects of the plasma, modeling, simulation and design tools for analysis and development of controllers, a flexible and modular architecture for implementation and testing of algorithms and many fully validated models. Many elements of the system have been applied to and implemented on NSTX and MAST. The DIII-D realtime plasma control system together with the integrated modeling and simulation suite have been selected for operational use by both the KSTAR and EAST tokamaks, and are also being used at General Atomics to investigate control issues for ITER.

Advanced tokamak profile evolution in DIII-D

Using off-axis electron cyclotron current drive (ECCD), self-consistent integrated advanced tokamak operation has been demonstrated on DIII-D [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159], combining high β (>3%) at high q (qmin>2.0) with good energy confinement (H89∼2.5) and high noninductive current fraction (fBS∼55%,fNI∼90%). Modification of the current profile by ECCD led to internal transport barrier formation even in the presence of type I edge localized modes. Improvements were observed in all transport channels, and increased peaking of profiles led to higher bootstrap current in the core. Separate experiments have shown the ability to maintain a nearly steady-state current profile for up to 1 s with qmin>1.5. Modeling indicates that this favorable current profile can be maintained indefinitely at a higher βN using tools available to the near-term DIII-D program. Modeling and simulation have become essential tools for the experimental program in interpreting the data and developing detail plans for new experiments.

Beam-driven energetic particle modes in advanced tokamak plasmas

A major goal of the DIII-D program is to study `advanced tokamak' plasmas with good confinement, large normalized β, and a large fractionof self-sustained current. Many of these plasmas have large beam pressures(≲(1/3) of the total pressure) and weak magnetic shear; Alfvén instabilities with laboratory frequencies of 100-250 kHz are oftenobserved. The instabilities correlate with reductions in the neutron ratebelow the classically expected value, complicatingdetermination of the pressure and current profiles. Quantitative analysis of onecase suggests that two types of energetic particle modes are destabilized: the resonant toroidicity-induced Alfvén eigenmode and theresonant kinetic ballooning mode. The strong dependence on neutralbeam injection parameters and the variability in mode frequency arequalitatively consistent with this identification. Further analysis andmeasurements are planned.

Tools for remote collaboration on the DIII-D National Fusion Facility

The DIII-D National Fusion Facility, a tokamak experiment funded by the US Department of Energy and operated by General Atomics (GA), is an international resource for plasma physics and fusion energy science research. This facility has a long history of collaborations with scientists from a wide variety of laboratories and universities from around the world. That collaboration has mostly been conducted by travel to and participation at the DIII-D site. Many new developments in the computing and technology fields are now facilitating collaboration from remote sites, thus reducing some of the needs to travel to the experiment. Some of these developments include higher speed wide area networks, powerful workstations connected within a distributed computing environment, network based audio/video capabilities, and the use of the world wide web. As the number of collaborators increases, the need for remote tools become important options to efficiently utilize the DIII-D facility. In the last two years a joint study by GA, Princeton Plasma Physics Laboratory (PPPL), Lawrence Livermore National Laboratory (LLNL), and Oak Ridge National Laboratory (ORNL) has introduced remote collaboration tools into the DIII-D environment and studied their effectiveness. These tools have included the use of audio/video for communication from the DIII-D control room, the broadcast of meetings, use of inter-process communication software to post events to the network during a tokamak shot, the creation of a DCE (distributed computing environment) cell for creating a common collaboratory environment, distributed use of computer cycles, remote data access, and remote display of results. This study also included sociological studies of how scientists in this environment work together as well as apart. As a result of these studies, there is now in place an automated distributed processing environment connected to the real-time experimental operations which can be joined by users at remote locations. This environment will allow further exploration of the technology and sociology of remote participation in the DIII-D program. Having the tools in place has already permitted remote participation in DIII-D experiments that would not have occurred otherwise, and thus the introduction of these tools has shown the initial feasibility of increasing and improving remote collaboration.

DIII-D Program Results and Future Plans

This paper summarizes recent research results and future plans for the DIII-D National Fusion Program. The full set of transparencies for this presentation can be found at http://fusion.gat.com/pubs-ext/presentations/Stambaugh_FPA00.pdf

Metallurgical bonding development of V–4Cr–4Ti alloy for the DIII-D radiative divertor program

General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D Radiative Divertor (RD) upgrade. The V–4Cr–4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V–4Cr–4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V–4Cr–4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid-state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high-strength, vacuum leak-tight joints by all of the methods under investigation. The solid-state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V–4Cr–4Ti alloy.

Recent DIII-D Results

This paper summarizes the recent DIII-D experimental results and the development of the relevant hardware systems. The DIII-D program focuses on divertor solutions for next generation tokamaks such as International Thermo-nuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX), and on developing configurations with enhanced confinement and stability properties that will lead to a more compact and economical fusion reactor. The DIII-D program car-ries out this research in an integrated fashion.

The long range DIII-D plan

The mission of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. The National Energy Strategy calls for the development of magnetic fusion as an energy option with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new diverter concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization. Results from experiments at DIII-D and other tokamaks indicate that plasmamore » and divertor performance can be increased transiently beyond the baseline conceptual design of ITER. A simultaneous long pulse demonstration of such improved tokamak plasma and divertor operation for steady state would establish an advanced physics foundation for the tokamak physics experiment program, provide new operating options for ITER, and open a path to an attractive DEMO. The planned DIII-D program incorporates new theory and technology developments to extend the tokamak experimental physics database toward steady state. This research program will also continue to provide increased understanding in many areas of fusion science and technology.« less

Overview of H-mode studies in DIII-D

A major portion of the DIII-D program includes studies of the L-H transition, of the VH-mode, of particle transport and control and of the power-handling capability of a divertor. Significant progress has been made in all of these areas and the aim is to summarize the major results. An increased understanding of the origin of improved confinement in H-mode and in VH-mode discharges has been obtained, good impurity control has been achieved in several operating scenarios, studies of helium transport provide encouraging results from the point of view of reactor design, an actively pumped divertor chamber has controlled the density in H-mode discharges and a radiative divertor is a promising technique for controlling the heat flux from the main plasma.

Future directions of the DIII-D tokamak program

The author discusses possible future direction for the DII-D tokamak in terms of developing the US and international tokamak research programs. DII-D is the second largest US magnetic fusion facility; an advanced tokamak and divertor research device for addressing plasma confinement and control issues, a testbed for demonstrating and developing fusion technology components, and a proving ground for applied physics theories and diagnostic instrumentation. DII-D is the largest non-circular tokamak in the US and is operated with extensive national and international collaboration, including a long-standing Japanese collaboration and major multi-disciplinary collaborations with LLNL, ORNL, and UCLA. The DIII-D program is now demonstrating advanced tokamak operating modes transiently and plans in future experiments to sustain these improved confinement, stability, and divertor modes with non-inductive rf current drive. The long range research program focuses on divertor and advanced tokamak issues. The goal is to provide an integrated demonstration of well-confined high-beta divertor plasma with non-inductive current drive. The author presents ways that DIII-D can contribute to technology issues relevant to ITER physics issues, with the requesite investment of resources.

DIII-D research program

The DIII-D tokamak is uniquely positioned to contribute to the development of Magnetic Fusion Energy over the next decade. Recent stability and confinement improvements resulting from current profile control and the discovery of the VH-mode stress the importance of non-inductive current drive. The DIII-D program plan calls for the implementation of high rf systems for such profile control; localized heating and current drive for improved to tokamak performance and advancing divertor research under current drive conditions. Funding levels will determine the potential impact of this $400 M research facility by determining the pace for implementation of rf and divertor upgrades as well as determining the amount of experimental operating time. A strong DIII-D program can effectively address the R&D issues of next generation tokamaks and allow the large number of DIII-D collaborators to explore new ideas to advance the tokamak to a commercially attractive energy option.