M. Viola

Exploration of Spherical Torus Physics in the NSTX Device

The National Spherical Torus Experiment (NSTX) is being built at the Princeton Plasma Physics Laboratory to test the fusion physics principles for the Spherical Torus (ST) concept at the MA level. The NSTX nominal plasma parameters are R {sub 0} = 85 cm, a = 67 cm, R/a greater than or equal to 1.26, B {sub T} = 3 kG, I {sub p} = 1 MA, q {sub 95} = 14, elongation {kappa} less than or equal to 2.2, triangularity {delta} less than or equal to 0.5, and plasma pulse length of up to 5 sec. The plasma heating/current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up, as well as a dispersive scrape-off layer for heat and particle flux handling.

Mechanical design of the NSTX Liquid Lithium Divertor

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux.

Decontamination and decommissioning of the Tokamak Fusion Test Reactor

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The decontamination and decommissioning (D&D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D&D Project where a facility is isolated and cleaned up by bulldozing all facility and hardware systems to a greenfield condition. The mission of TFTR D&D is to: a) surgically remove items which can be re-used within the DOE complex, b) remove tritium contaminated and activated systems for disposal, c) clear the test cell of hardware for future reuse, d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 and e) provide data on the D&D of a large magnetic fusion facility.

Testing of compact bolted fasteners with insulation and friction-enhanced shims for NCSX

The fastening of the National Compact Stellarator Experiments (NCSX) modular coils presented a number of engineering and manufacturing challenges due to the high magnetic forces, need to control induced currents, tight tolerances and restrictive space envelope. A fastening method using high strength studs, jack nuts, insulating spacers, bushings and alumina coated shims was developed which met the requirements. A test program was conducted to verify the design. The tests included measurements of flatness of the spacers, determination of contact area, torque vs. tension of the studs and jack nuts, friction coefficient tests on the alumina and G-10 insulators, electrical tests, and tension relaxation tests due to temperature excursions from room temperature to liquid nitrogen temperatures. This paper will describe the design and the results of the test program.

Low distortion welded joints for NCSX

The National Compact Stellarator Experiment (NCSX) required precise positioning of the field coils in order to generate suitable magnetic fields. A set of three modular field coils were assembled to form the Half Field-Period Assemblies (HPA). Final assembly of the HPA required a welded shear plate to join individual coils in the nose region due to the geometric limitations and the strength constraints. Each of the modular coil windings was wound on a stainless steel alloy (Stellalloy) casting. The alloy is similar to austenitic 316 stainless steel. During the initial welding trials, severe distortion, of approximately 1/16″, was observed in the joint caused by weld shrinkage. The distortion was well outside the requirements of the design. Solutions were attempted through several simultaneous routes. The joint design was modified, welding processes were changed, and specialized heat reduction techniques were utilized. A final joint design was selected to reduce the amount of weld material needed to be deposited, while maintaining adequate penetration and strength. Several welding processes and techniques using Miller Axcess equipment were utilized that significantly reduced heat input. The final assembly of the HPA was successful. Distortion was controlled to 0.012″, well within the acceptable design tolerance range of 0.020″ over a 3.5 foot length.

Advantages of high tolerance measurements in fusion environments applying photogrammetry

Photogrammetry, a state-of-the-art technique of metrology employing digital photographs as the vehicle for measurement, has been investigated in the fusion environment. Benefits of this high tolerance methodology include relatively easy deployment for multiple point measurements and deformation/distortion studies. Depending on the equipment used, photogrammetric systems can reach tolerances of 25 microns (0.001 in) to 100 microns (0.004 in) on a 3-meter object. During the fabrication and assembly of the National Compact Stellarator Experiment (NCSX) the primary measurement systems deployed were CAD coordinate-based computer metrology equipment and supporting algorithms such as both interferometer-aided (IFM) and absolute distance measurementbased (ADM) laser trackers, as well as portable Coordinate Measurement Machine (CMM) arms. Photogrammetry was employed at NCSX as a quick and easy tool to monitor coil distortions incurred during welding operations of the machine assembly process and as a way to reduce assembly downtime for metrology processes. This paper will explore the use of photogrammetry on NCSX during field period assembly (FPA) and the results it achieved. It will also explore other applications of this method and discuss future plans for use.

National Spherical Torus Experiment (NSTX) torus design, fabrication and assembly

The National Spherical Torus Experiment (NSTX) is a low aspect ratio spherical torus (ST) located at Princeton Plasma Physics Laboratory (PPPL), Fabrication, assembly, and initial power tests were completed in February of 1999. The majority of the design and construction efforts were constructed on the Torus system components. The Torus system includes the centerstack assembly, external poloidal and toroidal coil systems, vacuum vessel, torus support structure and plasma facing components (PFCs). NSTXs low aspect ratio required that the centerstack be made with the smallest radius possible. This, and the need to bake NSTXs carbon-carbon composite plasma facing components at 350 degrees C, was major drivers in the design of NSTX. The centerstack assembly consists of the inner legs of the toroidal field (TF) windings, the Ohmic heating (OH) solenoid and its associated tension cylinder, three inner poloidal field (PF) coils, thermal insulation, diagnostics and an Inconel casing which forms the inner wall of the vacuum vessel boundary. It took approximately nine months to complete the assembly of the centerstack. The tight radial clearances and the extreme length of the major components added complexity to the assembly of the centerstack components. The vacuum vessel was constructed of 304-stainless steel and required approximately seven months to complete and deliver to the Test Cell. Several of the issues associated with the construction of the vacuum vessel were control of dimensional stability following welding and controlling the permeability of the welds. A great deal of time and effort was devoted to defining the correct weld process and material selection to meet our design requirements. The PFCs will be baked out at 350 degrees C while the vessel is maintained at 150 degrees C. This required care in designing the supports so they can accommodate the high electromagnetic loads resulting from plasma disruptions and the resulting relative thermal expansions between the PFCs and the vacuum vessel on which supports are attached. This paper will provide a brief review of the issues associated with the design, fabrication and assembly of the NSTX Torus system including those outlined above.

A small rectangular Edge Localized Mode control coil design able to withstand a 400°C environment

Recently, an Edge Localized Mode (ELM) control coil was developed for use on the DIII-D tokamak. The coil design represented a significant challenge due primarily to the requirement for the coil insulation to withstand bakeout temperatures of 400°C for extended periods. This requirement ruled out most common organic insulating systems and necessitated a significant prototyping and development effort, leading to the selection of an advanced high temperature glass/polyimide resin system. The development included developing a heating mechanism that provided the discrete temperature ramp cycles and cure cycles required by this exotic resin. To complicate matters, the resin had a limited shelf life. Additionally the coil was small and rectangular in shape with rather small corner radii. This created a corner buildup that was not previously encountered and made dimensional control difficult. Another unique design requirement was the need to apply a sufficient internal pre-load to the wound and cured coil to insure there will be no relative motion between the coil and the Inconel case due to Lorentz forces from the 4 Tesla toroidal field on the vessel center post. This led to development of very unique leaf springs and a significant research and development effort coupled with an equally arduous finite element analysis effort. A satisfactory prototype was produced. This paper will focus primarily on the manufacturing challenges and discuss the prototyping effort.

The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.