P.M. Anderson

Development, installation, and initial operation of DIII-D graphite armor tiles

An upgrade of the DIII-D vacuum vessel protection system has been completed. The ceiling, floor, and inner wall have been armored to enable operation of CIT-relevant double-null diverted plasmas and to enable the use of the inner wall as a limiting surface. The all-graphite tiles replace the earlier partial coverage armor configuration which consisted of a combination of Inconel tiles and graphite brazed to Inconel tiles. A new all-graphite design concept was chosen for cost and reliability reasons. The ten minute duration between plasma discharge required the tiles to be cooled by conduction to the water-cooled vessel wall. Using two and three-dimensional analyses, the tile design was optimized to minimize thermal stresses with uniform thermal loading on the plasma-facing surface. Minimizing the stress around the tile hold-down feature and eliminating stress concentrators were emphasized in the design. The design of the tile fastener system resulted in sufficient hold-down forces for good thermal conductance to the vessel and for securing the tile against eddy current forces. The tiles are made of graphite, and a program to select a suitable grade of graphite was undertaken. Initially, graphites were compared based on published technical data. Graphite samples were then tested for thermal shock capacity in an electron beam test facility at the Sandia National Laboratory (SNLA) in Albuquerque, New Mexico, USA. The expected operational heat load to the tiles is 500 W/cm 2 for 10 s over the entire tile during limiter operation and 1000 W/cm 2 over a 2 cm wide band for 10 s for divertor operation. Comparative stress analysis was performed on the basis of 1000 W/cm 2 uniformly distributed on the front surface for 5 s with 10 min between pulses. Prototype tiles survived beam heating to the limit of the beam system which was 1000 W/cm 2 for 4 s or 1750 W/cm 2 for 1.75 s. The electron beam facility at the Sandia National Laboratory was used to deposite higher heat flux in a 2 cm wide band in order to model divertor heat flux conditions. These tests were successful with heat fluxes up to 2500 W/cm 2 for 5 s. Both the ion beam and the electron beam tests produced surface temperatures up to 3100 °C which resulted in minor surface erosion but no structural cracking. Installation of approximately 1600 armor tiles was accomplished in November/December 1987. Initial operation experience with the new armor system is summarized.

Design, fabrication, installation and testing of in-vessel control coils for DIII-D

Abstract Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. Recently, these coils have also demonstrated significant benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modelling has shown that substantial performance improvements can be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and have been tested successfully. Installation of a set of twelve internal coils and magnetic sensors in the DIII-D tokamak is to be completed in December 2002. The design requirement for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator.

Design, fabrication, installation, testing and initial results of in-vessel control coils for DIII-D

Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. In 2000, these coils also demonstrated benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements could be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and were power tested successfully after several bakes to 350/spl deg/C. A full set of twelve internal coils and related magnetic sensors are now operational in the DIII-D tokamak. The design requirements for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator. Elastic-plastic analysis was used to demonstrate acceptable thermal stresses during baking conditions. Analysis determined the optimum water cooling channel diameter. The coils were tested in high toroidal field to the limit of the power supply of 4.5 kA DC with inductance-limited current for frequencies between 300 Hz and 1000 Hz. Recent results are presented.

Development path of low aspect ratio tokamak power plants

Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% beta toroidal with the normalized beta as high as 10 and fully aligned 100% bootstrap current. Such beta values coupled with 2--3 T toroidal fields imply a pilot plant about the size of the present DIII-D tokamak could produce {approximately} 800 MW thermal, 160 MW net electric, and would have a ratio of gross electric power over recirculating power (Q{sub PLANT}) of 1.9. The high beta values in the ST mean that E x B shear stabilization of turbulence should be 10 times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated beta values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2--3 times the pilot plant size the Q{sub PLANT} rises to 4--5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He{sup 3} could be burned in a device with Q{sub PLANT} {approximately} 4.

Thermal design, analysis, and experimental verification for a DIII-D cryogenic pump

Abstract As part of the advanced divetoor program, it is planned to install a 50 m 3 /s capacity cryopump for particle removal in the DIII-D tokamak. The cryopump will be located in the outer bottom corner of the vacuum vessel. The pump will consist of a surface at liquid helium temperature (helium panel) with a surface area of about 1 m 2 , a surface at liquid nitrogen temperature (nitrogen shield) to reduce radiation heat load on the helium panel, and a secondary shield around the nitrogen shield. The cryopump design poses a number of thermal hydraulic problems such as estimation of heat loads on helium and nitrogen panels, stability of the two-phase helium flow, performance of the pump components during high temperature bakeout, and cooldown performance of the helium panel from ambient temperatures. This paper presents the thermal analysis done to resolve these issues. A prototypic experiment performed at General Atomics verified the analysis and increased the confidence in the design. The experimental results are also summarized in this paper.

Data acquisition and protection for new DIII-D in-vessel coils

The installation of new internal magnetic coils (I-Coils) in the DIII-D tokamak at General Atomics required extensive additions to the experiment data acquisition and protection capabilities. This set of 12 coils (up to 7 kA each) is designed to allow improved feedback stabilization of resistive wall modes which limit the plasma performance. The acquisition and signal conditioning needs of the I-Coil power system presented an opportunity to try a new data acquisition approach which increased both the sampling rate and sample size per channel compared to the standard DIII-D CAMAC acquisition equipment. A 96 channel compact-PCI (cPCI) digitizer system was purchased for the I-Coil project to acquire up to approximately 380 MB of power supply and coil current data per plasma discharge. Additional instrumentation and control was provided to protect personnel, the new coils, the tokamak, the facility and improve machine availability. This paper will present discussions of technical and programmatic requirements, bases for requirements, the design selection outcome, installation experience, integration issues, commissioning experience, and lessons learned. The data acquisition system is described in detail including a conservative signal isolation scheme, signal grounding standards, anti-aliasing filters, and synchronization of acquisition. Protection interlocks are described, including high voltage isolation, water flow measurement, and the coil grounding-shorting switches.

Analysis and testing of the DIII-D ohmic heating coil lead repair clamp

DIII-D has been operating for the last year with limited volt-second capabilities due to structural failure of a conductor lead to one of the ohmic heating (OH) solenoids. The conductor failure was due to poor epoxy impregnation of the overwrap of the lead pack, resulting in copper fatigue and a water leak. A number of structural analyses were performed to assist in determining the failure scenario and to evaluate various repair options. A fatigue stress analysis of the leads with a failed epoxy overwrap indicated crack initiation after 1000 cycles at the maximum operating conditions. The failure occurred in a very inaccessible area which restricted design repair options to concepts which could be implemented remotely. Several design options were considered for repairing the lead so that it can sustain the loads for 7.5 Vs conditions at full toroidal field. A clamp, along with preloaded banding straps and shim bags, provides a system that guarantees that the stress at the crack location is always compressive and prevents further crack growth in the conductor. Due to the limited space available for the repair, it was necessary to design the clamp system to operate at the material yield stress. The primary components of the clamp system were verified by load tests prior to installation. The main body of the clamp contains a load cell and potentiometer for monitoring the load-deflection characteristics of the clamp and conductors during plasma operation. Strain gages provide redundant instrumentation. If required, the preload on the conductors can be increased remotely by a special wrench attached to the clamp assembly.

Mechanical Design and Fabrication of the Lower Divertor for DIII-D

The lower divertor of the DIII-D tokamak is being modified to provide improved density control of the tokamak plasma during operation in a high triangularity double null configuration. This divertor replaces the low triangularity advanced divertor installed in 1990. The design and analysis of the lower divertor is complete and hardware is being fabricated. Installation of the new divertor is scheduled to be completed by the end of 2005. The primary component of the lower divertor is a toroidally continuous flat plate. The plate is water cooled for heat removal. Three rows of graphite tiles are mechanically attached to the plate to shield it from plasma impingement. Owing to a concern over excessive erosion caused by plasma impingement, the through tile-face bolt holes have been eliminated from graphite in areas of high heat flux. The plate is water cooled for heat removal between shots and heated to 350degC with hot air and inductive current during vessel baking. The divertor plate is supported 100 mm from the vacuum vessel floor by two rows of 24 supports that must react the vertical loads due to halo currents. These supports are radially flexible to allow for differential radial thermal expansion between the divertor ring and the floor. The space below the plate forms a pumping plenum connecting the floor strike point to the lower cryopump. Upgraded floor tiles inboard of the plate will be installed to improve the target for the plasma strike point for outer leg pumping. The divertor plate is to be fabricated in four 90 deg sectors from type 316 stainless steel. Each sector consists of two plate halves with three machined coolant channels and is joined together by spot welds and perimeter seam TIG welds. The vacuum tight 90 deg plate sectors are welded together inside the vessel to form a toroidally continuous ring. The water cooling/air bake-out lines connecting the 4 sectors into two 180 deg cooling circuits will be welded in situ. Several plasma diagnostics will require some modification or relocation for integration into the divertor system

Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

ENGINEERING, INSTALLATION, TESTING, AND INITIAL OPERATION OF THE DIII-D ADVANCED DIVERTOR

The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The electrode is a continuous Inconel ring armoured with graphite. It is water cooled during normal operation and bakeable to 350° C for vessel conditioning. The electrode is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall, but stiff in the vertical direction to restrain the ring against large disruption forces. Similarly, the coolant and electrical feeds also are flexible in the radial direction but rigidly supported in other directions where they pass outside the toroidal field. All the feeds are supported from and maintain a 5 kV isolation to the vessel wall. Plasma facing insulators are needed in order to maintain electrical insulation. In addition, all biased surfaces behind and underneath the ring are insulated to prevent breakdown along the field lines. Tests of insulators were made in DIII-D during plasma operations prior to deciding on the final design. Other tests and analyses were performed on insulating materials. The baffle allows for future installation of approximately 50,000 l/s cryo pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx with a high probability that backstreaming particles will be reionized and redirected to the aperture. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint.