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[en] The EA evaluates the proposed action of modifying the DIII-D fusion facility and conducting related research activities at the GA San Diego site over 1995-1999 under DOE contract number DE-ACO3-89ER51114. The proposed action is need to advance magnetic fusion research for future generation fusion devices such as ITER and TPX. It was determined that the proposed action is not a major action significantly affecting the quality of the human environment according to NEPA; therefore a finding of no significant impact is made and an environmental impact statement is not required
[en] We use the guiding center code ORBIT to study the broadening of resonances and the parametric dependence of the resonance frequency broadening width ΔΩ on the nonlinear particle trapping frequency ωb of wave-particle interaction with specific examples using realistic equilibrium DIII-D shot 159243 (Collins et al. 2016 Phys. Rev. Lett. 116 095001). When the mode amplitude is small, the pendulum approximation for energetic particle dynamics near the resonance is found to be applicable and the ratio of the resonance frequency width to the deeply trapped bounce frequency ΔΩ/ωb equals 4, as predicted by theory. Lastly, it is found that as the mode amplitude increases, the coefficient a=ΔΩ/ωb becomes increasingly smaller because of the breaking down of the nonlinear pendulum approximation for the wave-particle interaction.
[en] The large size and energy content of Doublet III produces mechanical engineering problems similar to those expected in the next generation of large fusion devices. The toroidal field coil design was dictated by fatigue properties and crack growth rates in the copper, plus the shear fatigue properties of the epoxy/glass bonding between turns. Preloading of the toroidal field coil was accomplished in part by wrapping the composite center column with multiple layers of S-glass prepreg tape under high tension. The process required design and development of new state-of-the-art equipment. During construction, the machining, handling, and bonding of the large delicate components created many detailed structural problems. The assembly of the precision joints of the coil required special tooling. (orig./GG)
[en] Error field correction results in DIII-D plasmas are presented in various configurations. In both left-handed and right-handed plasma configurations, where the intrinsic error fields become different due to the opposite helical twist (handedness) of the magnetic field, the optimal error correction currents and the toroidal phases of internal(I)-coils are empirically established. Applications of the Ideal Perturbed Equilibrium Code to these results demonstrate that the field component to be minimized is not the resonant component of the external field, but the total field including ideal plasma responses. Consistency between experiment and theory has been greatly improved along with the understanding of ideal plasma responses, but non-ideal plasma responses still need to be understood to achieve the reliable predictability in tokamak error field correction.
[en] The Doublet III tokamak is to be modified wherein the original 'doublet' plasma containment vacuum vessel will be exchanged with one of a large dee-shaped cross section. The basic dimensions of the dee vessel will allow plasmas of 1.7-m major radius, 0.7-m minor radius, and a vertical elongation of 1.8. Installation of a large dee vessel in Doublet III is made possible by the demountable toroidal field coils and the large, low-ripple volume they include. Ripple at the plasma edge will be less than one percent. The plasma parameters affecting the design of the vessel will be reviewed including plasma current, power, disruption time, allowable error field, impurity control techniques, pulse length, and limiter schemes. A driving requirement for the design of the vessel is to maximize the access to the plasma for auxiliary heating (both neutral beam injection and radio frequency heating), diagnostics, developmental component and material testing, and pumping. The dee vessel is structurally designed along the same lines as the present vessel: an Inconel 625, all-welded, continuous chamber in a corrugated sandwich construction. An overview of the vessel design and its solutions to the design criteria will be presented. An overview will also be presented of the entire modification project which includes replacement of some coils, and addition of support structure, limiters and vessel armor, and power system components
[en] The current DIII-D plasma control system (PCS) has evolved through several iterations into a robust platform that has been adopted at several fusion devices around the world. Each installation, as well as each new upgrade at DIII-D, has presented new challenges. Each of these challenges has provided an additional opportunity to expand our understanding of the requirements, alternative operational methods, and differing real-time implementations for Tokamak plasma control. This paper presents a brief historical overview of PCS hardware evolutions and describes some of the design, structure, and techniques that have allowed the PCS to be a productive component at many fusion facilities. It will also discuss some of the major differences between the individual PCS installations and bring to light some of the major challenges that were overcome during integration. The lessons learned from these experiences provide general solutions and can inform control system designs for other next-generation devices. We also describe some limitations of the PCS relative to identified present and future needs at DIII-D and other devices, and discuss planned upgrades to the PCS to address these needs.
[en] Major upgrades to the DIII-D facility have been performed that significantly enhance the capability of both the DIII-D device and the entire facility. The most significant of these include the rotation of a neutral beam line, installation of a new lower divertor, and a significant set of new and enhanced diagnostics. The upgrades and initial results are presented in this paper
[en] OAK A271 COMPARISON OF SENSORS FOR RESISTIVE WALL MODE FEEDBACK CONTROL MILESTONE No.145 CONTAINING PLASMA INSTABILITIES WITH METAL WALLS. The most serious instabilities in the tokamak are those described by ideal magneto-hydrodynamic theory. These modes limit the stable operating space of the tokamak. The ideal MHD calculations predict the stable operating space of the tokamak may be approximately doubled when a perfectly conducting metal wall is placed near the plasma boundary, compared to the case with no wall (free boundary). The unstable mode distortions of the plasma column cannot bulge out through a perfectly conducting wall. However, real walls have finite conductivity and when plasmas are operated in the regime between the free boundary stability limit and the perfectly conducting wall limit, the unstable mode encountered in that case the resistive wall mode, can leak out through the metal wall, allowing the mode to keep slowly growing. The slow growth affords the possibility of feedback stabilizing this mode with external coils. DIII-D is making good progress in such feedback stabilization research and in 2002 will use an improved set of mode sensors inside the vacuum vessel and closer to the plasma surface which are expected theoretically to improve the ability to stabilize the resistive wall mode
[en] Using the plasma reluctance, the Ideal Perturbed Equilibrium Code is able to efficiently identify the structure of multi-modal magnetic plasma response measurements and the corresponding impact on plasma performance in the DIII-D tokamak. Recent experiments demonstrated that multiple kink modes of comparable amplitudes can be driven by applied nonaxisymmetric fields with toroidal mode number n = 2. This multi-modal response is in good agreement with ideal magnetohydrodynamic models, but detailed decompositions presented here show that the mode structures are not fully described by either the least stable modes or the resonant plasma response. This paper identifies the measured response fields as the first eigenmodes of the plasma reluctance, enabling clear diagnosis of the plasma modes and their impact on performance from external sensors. The reluctance shows, for example, how very stable modes compose a significant portion of the multi-modal plasma response field and that these stable modes drive significant resonant current. Finally, this work is an overview of the first experimental applications using the reluctance to interpret the measured response and relate it to multifaceted physics, aimed towards providing the foundation of understanding needed to optimize nonaxisymmetric fields for independent control of stability and transport