E.E. Reis

Observation of poloidal current flow to the vacuum vessel wall during vertical instabilities in the DIII-D tokamak

An attached poloidal current, which flows in a circuit lying partly in the vacuum vessel wall and partly in the scrape-off layer of the plasma, is observed during vertical instabilities in the DIII-D tokamak. A direct measurement of the current, using Rogowski loops on several protective tiles at locations where the plasma contacts the wall, is in good agreement with the value determined from MHD equilibrium reconstructions using measured values of magnetic field and flux. This attached current, which can reach transient peaks of several hundred kiloamperes, interacts with the toroidal magnetic field to create a large vertical force on the vacuum vessel. The predicted motion of the vessel resulting from the measured currents agrees well with the observed displacement of the vacuum vessel.

ARIES-RS divertor system selection and analysis

The ARIES-RS divertor system is selected and analyzed. A radiative divertor approach using Ne as the radiator is chosen to reduce the maximum heat flux to B 6M W m 2 . A 2 mm W layer is used to withstand surface erosion allowing a design life close to 3 full-power-years. This W coating on the V-alloy structure is castellated to meet structural design limits. A detailed description of the calculated heat flux distribution, thermal-hydraulics, structural analysis, fabrication methods and vacuum system design are presented. An innovative design using adjustable bolts is utilized to support the divertor plates, withstand disruption loads and allow adjustment of alignment between plates. With the exception of the concentration of Ne at the divertor, it is found that this divertor system design can satisfy all the design criteria and most of the functional requirements specified by the project.

Tensile fracture characterization of braze joined copper-to-CFC coupon assemblies

Abstract A vacuum brazing process was used to join a broad spectrum of carbon-fiber reinforced carbon matrix composite (CFC) materials, machined into cylindrical coupons, between coupons of oxygen-free copper, the braze alloy was a copper-base alloy which contained only low activation elements (Al, Si, and Ti) relative to a titanium baseline specification. This demonstration was of particular importance for plasma facing components (PFCs) under design for use in the Tokamak Physics Experiment (TPX); the braze investigation was conducted by General Atomics for the Princeton Plasma Physics Laboratory. A tensile test of each brazed assembly was conducted. The results from the braze processing, testing, and fracture characterization studies of this reporting support the use of CFCs of varied fiber architecture and matrix processing in PFC designs for TPX. Further, the copper braze alloy investigated is now considered to be a viable candidate for a low-activation bond design. The prediction of plasma disruption-induced loads on the PFCs in TPX requires that joint strength between CFC tiles and their copper substrate be considered in design analysis and CFC selection.

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.

Engineering design of cryocondensation pumps for the DIII-D radiative divertor program

A new double-null, slotted divertor configuration will be installed for the DIII-D Radiative Divertor Program at General Atomics in late 1996. Four cryocondensation pumps, three new and one existing, will be part of this new divertor. The purpose of the pumps is to provide plasma density control and to limit the impurities entering the plasma core by providing pumping at each divertor strike point. The three new pumps are based on the design of the existing pump, installed in 1992 as part of the Advanced Divertor Program. The pump continues to operate successfully. The toroidally continuous pumps vary in lengths from approximately 7 to 12 m depending upon their locations within the vessel. Each pump is independently operated and offers on average 0.7 m/sup 2/ of liquid helium-cooled pumping surface. The tubular pumping surface is surrounded concentrically by nitrogen shields and a particle shield of larger diameters. The nitrogen-cooled shields limit the heat flux on the helium surface. The particle shield limits energetic particles from impacting the helium and nitrogen cooled surfaces, preventing the condensed gases on the pump, primarily water, from being released. The new pumps require geometry modifications to the original design. Therefore, extensive modal and dynamic analyses were performed to determine the behavior of these pumps and their helium and nitrogen feed lines during disruption events. Thermal and fluid analyses were also performed to characterize the helium two-phase flow regime in the pumps and their feedlines. A flow testing program was completed to test the change in geometry of the pump feed lines with respect to helium flow stability. The results were compared to the helium thermal and fluid analyses to verify predicted flow regimes and flow stability.

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.

THERMAL STRESS ANALYSIS OF FIRE DIVERTOR

ABSTRACT The Fusion Engineering Research Experiment (FIRE) device is designed for high powerdensity and advanced physics operating modes. Due to the short distance of the divertor from theX–point, the connection lengths are short and the scrape off layer thickness is small. A relativelyhigh peak heat flux of 25 MW/m 2 is expected on the divertor.The FIRE divertor engineering design is based on the design approaches developed forInternational Thermonuclear Experimental Reactor (ITER). The geometry of the FIRE divertorconsists of water cooled copper fingers and a tungsten brush armor as plasma facing material.The divertor assembly consists of modular units for remote handling. A 316 stainless steel backplate is used for support and manifolding. The backing plate is joined to the copper fingers bypins. The coolant channel diameter is 8 mm at a pitch of 14 mm. The total power flow to theouter divertor is 35 MW. A water at an inlet temperature of 30°C, 1.5 MPa and a flow velocity of10 m/s is used with two channels in series. A margin of about 1.6 is obtained on the critical heatflux.A three dimensional thermal stress finite element (FE) analysis of this geometry wasperformed. Thermal hydraulic correlations derived for ITER were used to perform the thermalanalysis. Design changes were implemented to reduce the stresses and temperatures to acceptablelevels.

TPX divertor design

The TPX tokamak incorporates a double null slot divertor that may be operated in a single or double null mode. Provisions are incorporated to provide for radiative divertor operation to reduce the peak heat flux to the divertor. Particle pumping is provided through vertical ports to control the plasma density. This paper describes the conceptual design of the TPX divertor. The divertor is designed for steady state thermal operation. TPX pulse lengths will be from 1000 sec to steady-state. The materials used for the divertor are mostly titanium for the structure and water manifolds, dispersion strengthened copper for the water cooling tubes, and carbon-carbon (C-C) composite for the plasma facing surfaces. Low activation materials are used where possible in order to preserve hands on maintenance during the first two years of operation. Analysis indicates that surface heat fluxes as high as 15 MW/m/sup 2/ will heat the C-C plasma facing surface to above 1000/spl deg/C. This temperature results in an acceptable impurity release of the high conductivity C-C materials. The divertor sector (22.5/spl deg/ toroidally) has been designed into two modules which are the inner divertor module and the baffle/outer divertor module. This was done to allow for installation/removal of the divertor within the space limitations of the TPX plasma chamber. Water coolant lines, diagnostic instrumentation, and gas lines for radiative divertor gas feed are designed to be remotely connected/disconnected. Module to module alignment is critical to limit edge heating of the C-C surface and this alignment has been achieved by using mounting rings forming a common surface for aligning the modules. Disruption and halo current loads are significant and set the requirements for structural strength and attachment points.

Installation and initial operation of the DIII-D advanced divertor cryocondensation pump

Phase two of a divertor cryocondensation pump, the Advanced Divertor Program, is now installed in the DIII-D tokamak at General Atomics and complements the phase one biasable ring electrode. The installation consists of a 10 m long cryocondensation pump located in the divertor baffle chamber to study plasma density control by pumping of the divertor. The design is a toroidally electrically continuous liquid helium-cooled panel with 1 m/sup 2/ of pumping surface. The helium panel is single point grounded to the nitrogen shield to minimize eddy currents. The nitrogen shield is toroidally continuous and grounded to the vacuum vessel in 24 locations to prevent voltage potentials from building up between the pump and vacuum vessel wall. A radiation/particle shield surrounds the nitrogen-cooled surface to minimize the heat load and prevent water molecules condensed on the nitrogen surface from being released by impact of energetic particles. Large currents (>5000 A) are driven in the helium and nitrogen panels during ohmic coil ramp up and during disruptions. The pump is designed to accommodate both the thermal and mechanical loads due to these currents. A feedthrough for the cryogens allows for both radial and vertical motion of the pump with respect to the vacuum vessel. Thermal performance measured on a prototype verified the analytical model and thermal design of the pump. Characterization tests of the installed pump show the pumping speed in deuterium is 42,000 l/sec for a pressure of 5 mTorr. Induction heating of the pump (at 300 W) resulted in no degradation of pumping speed. Plasma operations with the cryopump show a 60% lower density in H-mode.