The DIII-D ECH system includes three launcher assemblies each of which can accommodate the RF beams from two gyrotrons. The launchers use four different designs for the mirrors which focus and direct the beams into the tokamak. The designs use molybdenum brazed to graphite, thin Glidcop or variable thickness Glidcop. A fourth design with laminated Glidcop/stainless steel construction has been operated, but no thermal data are available. All the mirrors operate without active cooling. This paper presents preliminary analyses and measurements of the thermal performance of the three designs for which data have been obtained.

The DIII-D ECH system includes three launcher assemblies each of which can accommodate the rf beams from two gyrotrons. The launchers use four different designs for the mirrors which focus and direct the beams into the tokamak. The designs use molybdenum brazed to graphite, thin Glidcop or variable thickness Glidcop. A fourth design with laminated Glidcop/stainless steel construction has been operated, but no thermal data are available. All the mirrors operate without active cooling. This paper presents preliminary analyses and measurements of the thermal performance of the three designs for which data have been obtained.

The DIII-D toroidal field [TF] belt bus system provides an electrical connection between adjacent TF-coil bundles to form a continuous current path for the TF-coil system. There is also a return path which is electrically isolated from the belt bus. The function of the system is to carry TF-coil current while minimizing the TF-coil error field in accordance with physics requirements. The system is currently capable of handling 5 s of operation with a peak current of 127 kA in the TF-coil. Future requirements for the system are the capability to support 10 s operation with 10 min cooldown periods in between shots. Experiments have been carried out which describe the physical parameters of the system, such as the contact resistance across the bus bar joints. Additionally, using an optical fiber-based temperature monitoring unit, the temperature response of the system to operations was determined. Based on these characterizations of the system, a 3-D thermal model was built to predict the behavior of the system for 10 s operations. The limitation of the system is the maximum allowable temperature of approximately 150/spl deg/C for the G11 insulators. The model was constructed full scale per engineering drawings using Solidworks, meshed, and then exported to Cosmos for analysis. Once good correlation was achieved with the observed responses to 5 s pulses, the behavior of the system for 10 s pulses was predicted. Various design modifications, such as water cooled bolts, were simulated in order to estimate their impact on creating a system that meets the 10 s criteria.

Each GA ECH launcher mirrors is used to transmit 800 kW of power to the plasma. Until 2000, the pulse length for use of these mirrors was limited to 2 s due to temperatures of the mirrors resulting from 1) a high ratcheted bulk temperature and 2) a large increase in temperature of the mirror during the pulse. A new design was proposed and implemented which has extended the capability of the mirror to 10 s with passive cooling. The important features of the new design are 1) increase in the passive heat transfer rate during cooling and 2) a modified shape of the mirror. The analysis shows that, the new mirrors can be used for 10 s pulses. The new mirrors have been installed in DIII-D, they have been used for up to 2 s pulses.

The DIII–D program has completed a series of density control and plasma core confinement experiments this past year. These experiments were designed to investigate the performance of baffled and open divertors with single-null plasmas and particle control in double-null plasmas. The experiments utilized all three of the DIII–D divertor assemblies located in the lower outer corner, the upper outer corner, and the upper inner corner of the vessel, which were installed last year. Each divertor consists of a liquid helium cryopump, a shielded protective ring, and a gas puff system. The divertors were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo and toroidal, induced currents. With theoretical pumping speeds varying from 15,000 to 32,000 l/s, the cryopumps, combined with the baffle structures, collect particles and prevent them from recirculating back into the plasma core. The intent of the gas puff systems is to inject neutral gases in and around the divertors to minimize the heat flux on the divertors, minimizing the impurities generated by the excessive heating of the divertor graphite tiles. This hardware permits either single- or double-null plasma experiments and enables continued research of well confined high beta divertor plasmas with noninductive current drive, which is one of the primary research goals of DIII–D.

A major upgrade of the DIII-D divertor, with the goal of enhancing impurity and density control and increasing the thermal pulse length limit of advanced tokamak (AT) plasmas has been successfully completed and commissioned. The integrated system that includes independent cryopumps at both the inner and the outer legs of the divertor, private flux region and outboard baffles, and improved graphite divertor armor, has been successfully applied to a variety of plasma conditions. Comparison of similar discharges before and after the upgrades show that with the new divertor the core plasma neutral source and carbon content are lower by as much as 50%. Calculations supported by preliminary infra-red (IR) camera measurements show that the new graphite armor design increases the limit on the discharge duration, due to temperature of the tile edges reaching sublimation point, by an order of magnitude. With the new system we have been able to control the density of high confinement H-mode plasmas to less than 1/3 of the Greenwald limit. It is observed that with divertor pumping during the current ramp phase the wall particle inventory and consequently the density rise after the H-mode transition can be significantly reduced.

The design of a double-null divertor for use in JT-60SU is presented. The free-boundary equilibrium code EFIT is used to establish a symmetric highly triangular double-null plasma shape. The baffle shapes are highly contoured to match the equilibrium, with the plasma-facing surfaces intersecting the flux surfaces at steep angles in the regions of high heat flux. These contoured surfaces also provide a tightly baffled design with small-aperture pumping gaps near both the inner and outer divertor strike points. The gaps provide adequate throughput of D-2 gas for active control of impurity entrainment at reasonable pressures. The structural design is shown to be consistent with both forces from disruptions and thermal stress during vacuum vessel bakeout.

We present the results from DIII-D experiments and modeling focused on the divertor issues of an `Advanced Tokamak' (AT). Operation at high plasma pressure β with good energy confinement H requires core and divertor plasma shaping and current profile J(r) control with ECH current drive. Transport modeling indicates that the available DIII-D ECH power determines a density and temperature regime for sustained DIII-D AT experiments. We demonstrate that a high-δ, unbalanced double null divertor with cryopumping (D-2000) is a flexible AT divertor. Impurity levels in AT experiments have been reduced by careful alignment of the divertor tiles; this, in turn has changed the time evolution of the core J(r) profiles. New physics has been observed near the X-point and private flux regions, including flow reversal and recombination, that is important in understanding and controlling the flows and thereby the radiation in the divertor region, which reduces the divertor heat flux.

General Atomics recently completed a divertor design study for JAERI for the JT-60 Super Upgrade (JT-60SU) tokamak. JT-60SU is being designed as a superconducting device for an integrated R&D investigation of steady-state operation in a tokamak. A divertor design was developed to accommodate double-null operation for a 1000 s discharge duration at 8 MA of plasma current and 80 MW of auxiliary heating. The work reported here is an extension of a previous design study.(1,2) The thermal requirements are a peak heat flux of 9 MW/m(2), a maximum surface temperature of 1600 degreesC, and a poloidal cooling flow configuration for the plasma facing components. The structural requirements are determined from both the predicted stresses due to halo currents as well as the stresses due to differential thermal expansion encountered during bakeout. The halo current loads are based on a nominal halo current of 0.19 Ip with a 2.0 toroidal peaking factor. Analysis has determined that the halo current load per centimeter of circumference is P = 2856 (1+cos theta) N/cm, where 8 is the toroidal angle. The loads due to differential thermal expansion are a result of an expected 100 degreesC temperature difference between the vacuum vessel and divertor during bakeout.Based on the aforementioned criteria, a divertor design was developed for all three areas of the JT-60SU divertor: the inner baffle, the private flux baffle, and the outer baffle. In order to have highly reliable divertor components, flexible supports sized to accommodate the structural loads are utilized in the design rather than insulators or sliding interfaces. The plasma facing components are mounted on a structural mounting plate to form a removable and remotely-maintainable segment which is in turn mounted on the supports. For outer and private flux baffles, these structural mounting plates are joined together using a double shear joint design to form a structurally continuous ring to react the halo current loads. The plasma facing components are broken into 80 segmentation; however, the outer and private flux baffles have an alternating 8 degrees and 16 degrees structural segmentation which forms the double shear toroidal structural joint. The inner baffle takes advantage of its relatively short poloidal length and its proximity to the vacuum vessel to provide structural integrity. The thermal design consists of a plasma facing material of flat CFC tiles that are brazed onto a poloidally cooled copper heat sink. Adequate gaps between the baffles and wall are provided for pumping of recycled gas.

As predicted by theoretical models, under certain plasma configurations the edges of the graphite tiles lining the walls of the DIII-D vacuum vessel can exceed 2000/spl deg/C. This is the temperature at which ablation occurs for the ATJ graphite material for localized heating (1500/spl deg/C for the face of the tile) and thus represents an operational limit for the tokamak. This theoretical prediction is confirmed by observed increased carbon contamination in the plasma and wearing of the tile edges. Several approaches were investigated to minimize tile edge heating by reducing the amount of heat flux on the tiles toroidal and poloidal surfaces. This was accomplished by modifying the geometry of the top two rows of the DIII-D centerpost tiles.

The radiative divertor program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryo/helium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma