The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5–10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2–3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3–6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

The National Spherical Torus Experiment Upgrade (NSTX-U) is currently under construction at Princeton Plasma Physics Laboratory (PPPL) It is scheduled to start operations early in 2015. Upgrade designs were analyzed and qualified prior to the beginning of construction, but many issues arose during manufacture and assembly that required adjustments in design and analysis of components. Some designs relied on testing that occurred after final design when the actual material and processes were selected by vendors or in-house shops. Design of some components, like the bus bars, was deferred until field run interferences could be identified. Some components used materials that did not meet original specifications. New materials or processes had to be found and components sometimes needed requalification. PPPL responsible or "Cognizant Engineers" (COG's) and analysts worked closely to work out resolution of issues and perform redesign and reanalysis. Revisions to calculations were prepared and filed. Some significant items addressed during the construction period (or Title III in DOE parlance) are selected for more detailed discussion.

The NSTX-U requirements will double the Toroidal Field (TF), Plasma Current (Ip), Beam Injection Power, and extend pulse length. The larger centerstack requires re-aiming of the Multi Pulse Thomson Scattering (MPTS) lasers and Vacuum Vessel (VV) modifications at Bay L. The second neutral beam requirements include larger tangency radii and thus a VV modification at Bay K and Bay J. A cap design for a new weldment was developed to achieve these larger beam trajectories without losing the utility of the Bay J port for diagnostics. Analyses of loads indicated the need for reinforcements of the vessel at the midplane. NSTX has 6 picture frame type Resistive Wall Mode (RWM) coils around the exterior circumference of the vacuum vessel; each coil surrounds pairs of ports. The modifications needed for the upgrade were intended to minimize the impact to the RWM fields at the plasma. A Pro E global model segment was used to model the vacuum vessel. ANSYS was used to apply loads and investigate reinforcement configurations. A focused effort and analysis produced a design capable of achieving the desired performance of the upgrade while maintaining utility and continuity of RWM coil physics pre- and post-upgrade physics performance. The installation of the Bay J-K and Bay L Port Caps was completed and the reinforcing weldments have been partially installed.

The NSTX TF coil assembly experienced a failure in the TF center bundle during pulse number 143912 on July 20, 2011. A short circuit between turns 14 and 30 spread to turn 13 and resulted in melting of approximately 2 kg of copper and ejection of copper and solder from the cooling channels at the lower end of the TF bundle. An autopsy of the coil was immediately begun by PPPL personnel that included cutting numerous sections from the affected region of the coil and examining the chemical makeup of the insulation. In addition, careful machining to isolate various areas of turn to turn insulation uncovered suspicious spots that exhibited low resistance. This paper will present the methods used to isolate the location and identify the causes of the failure in the coil bundle and the design changes made to the NSTX Upgrade TF bundle and fabrication processes to prevent a recurrence of the fault.

The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5 s to 5–8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described.

The national spherical torus experiment (NSTX) will be upgraded to provide increased toroidal field, plasma current and pulse length. This involves the replacement of the so-called center stack, including the inner legs of the toroidal field (TF) coil, the Ohmic heating (OH) coil, and the inner poloidal field (PF) coils. In addition the increased performance of the upgrade requires qualification of remaining existing components for higher loads. Initial conceptual design efforts were based on worst-case combinations of possible currents that the power supplies could deliver. This proved to be an onerous requirement and caused many of the outer coils support structures to require costly heavy reinforcement. This has led to the planned implementation of a digital coil protection system (DCPS) to reduce design-basis loads to levels that are more realistic and manageable. As a minimum, all components must be qualified for the increase in normal operating loads with headroom. Design features and analysis efforts needed to meet the upgrade loading are discussed. Mission and features of the DCPS are presented.

The fundamental understanding of material response to a neutron and/or high heat flux environment can yield development of improved materials and operations with existing materials. A concept has been advanced to develop a facility for testing various materials under extreme heat and neutron exposure conditions at Princeton. The Extreme Environment Materials Research Facility comprises an environmentally controlled chamber (48 m boolean AND 3) capable of high vacuum conditions, with extreme flux beams and probe beams accessing a central, large volume target. The facility will have the capability to expose large surface areas (1 m boolean AND 2) to 14 MeV neutrons at a fluence in excess of 10 boolean AND 13 n/s. Depending on the operating mode. Additionally (deuterium) beam line power of 15-75 MW/m2 for durations of 1-15 seconds is planned. The facility will be housed in an existing test cell that previously held the Tokamak Fusion Test Reactor (TFTR).

The fastening of the National Compact Stellarator Experiment's (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.

The National Compact Stellarator Experiment (NCSX) was designed to test a compact, quasi-axisymmetric stellarator configuration. Flexibility and accurate realization of its complex 3D geometry were key requirements affecting the design and construction. While the project was terminated before completing construction, there were significant engineering accomplishments in design, fabrication, and assembly. The design of the stellarator core device was completed. All of the modular coils, toroidal field coils, and vacuum vessel sectors were fabricated. Critical assembly steps were demonstrated. Engineering advances were made in the application of CAD modeling, structural analysis, and accurate fabrication of complex-shaped components and sub-assemblies. The engineering accomplishments of the project are summarized.

The National Compact Stellarator Experiment (NCSX) was a collaborative effort between ORNL and PPPL. PPPL provided the assembly techniques with guidance from ORNL to meet design criteria. The individual vacuum vessel segments, modular coils, trim coils, and toroidal field coils components were delivered to the Field Period Assembly (FPA) crew who then would complete the component assemblies and then assemble the final three field period assemblies, each consisting of two sets of three modular coils assembled over a 120 vacuum vessel segment with the trim coils and toroidal field coils providing the outer layer. The requirements for positioning the modular coils were found to be most demanding. The assembly tolerances required for accurate positioning of the field coil windings in order to generate sufficiently accurate magnetic fields strained state of the art techniques in metrology and alignment and required constant monitoring of assembly steps with laser trackers, measurement arms, and photogrammetry. The FPA activities were being performed concurrently while engineering challenges were being resolved. For example, it was determined that high friction electrically isolated shims were needed between the modular coil interface joints and low distortion welding was required in the nose region of those joints. This took months of analysis and development yet the assembly was not significantly impacted because other assembly tasks could be performed in parallel with ongoing assembly tasks as well as tasks such as advance tooling setup preparation for the eventual welding tasks. The crew technicians developed unique, accurate time saving techniques and tooling which provided significant cost and schedule savings. Project management displayed extraordinary foresight and every opportunity to gain advanced knowledge and develop techniques was taken advantage of. Despite many risk concerns, the cost and schedule performance index was maintained nearly 1.0 during the assembly phase until project cancellation.In this paper, the assembly logic, the engineering challenges, solutions to those challenges and some of the unique and clever assembly techniques, will be presented.

The purpose of the NSTX Center Stack Upgrade project is to expand the NSTX operational space and thereby the physics basis for next-step ST facilities. The plasma aspect ratio (ratio of plasma major to minor radius) of the upgrade is increased to 1.5 from the original value of 1.26, which increases the cross sectional area of the center stack by a factor of ~ 3 and makes possible higher levels of performance and pulse duration.

The NCSX stellarator core design is built around a 3-period, highly shaped plasma with an assembly of four magnet systems, the TF coils (TF), the Modular Coils (MC), the PF Coils (PF), and the Trim Coils, that surrounds an all welded Vacuum Vessel (VV). The VV features approximately 100 ports for heating, pumping, diagnostics, and maintenance access. The entire system is surrounded by a cryostat to permit operation of the coils at liquid nitrogen temperature. The VV and coils are assembled in 120deg segments. The VV segments must be placed inside the MC by sliding the coils over each end of the vessel subassembly. Installation of the port extensions is done after this operation. They are slipped through access holes in the MC onto port stubs and welded on from inside. Figures 1 & 2 illustrate the assembly of the MC over a VV section and installation of TF and port extensions to complete a sector of the VV.

The National Compact Stellarator Experiment, NCSX, is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge national Laboratory. The goal of NCSX is to provide the understanding necessary to develop an attractive, disruption free, steady state compact stellaratorbased reactor design. This paper describes the recently revised designs of the critical interfaces between the modular coils, the construction solutions developed to meet assembly tolerances, and the recently revised trim coil system that provides the required compensation to correct for the “as built” conditions and to allow flexibility in the disposition of as-built conditions. In May, 2008, the sponsor decided to terminate the NCSX project due to growth in the project’s cost and schedule estimates. However significant technical challenges in design and construction were overcome, greatly reducing the risk in the remaining work to complete the project.

The National Compact Stellarator Experiment (NCSX) was a collaborative effort between ORNL and PPPL. PPPL provided the assembly techniques with guidance from ORNL to meet design criteria. The individual vacuum vessel segments, modular coils, trim coils, and toroidal field coils components were delivered to the Field Period Assembly (FPA) crew who then would complete the component assemblies and then assemble the final three field period assemblies, each consisting of two sets of three modular coils assembled over a 120° vacuum vessel segment with the trim coils and toroidal field coils providing the outer layer. The requirements for positioning the modular coils were found to be most demanding. The assembly tolerances required for accurate positioning of the field coil windings in order to generate sufficiently accurate magnetic fields strained state of the art techniques in metrology and alignment and required constant monitoring of assembly steps with laser trackers, measurement arms, and photogrammetry. The FPA activities were being performed concurrently while engineering challenges were being resolved. For example, it was determined that high friction electrically isolated shims were needed between the modular coil interface joints and low distortion welding was required in the nose region of those joints. This took months of analysis and development yet the assembly was not significantly impacted because other assembly tasks could be performed in parallel with ongoing assembly tasks as well as tasks such as advance tooling setup preparation for the eventual welding tasks. The crew technicians developed unique, accurate time saving techniques and tooling which provided significant cost and schedule savings. Project management displayed extraordinary foresight and every opportunity to gain advanced knowledge and develop techniques was taken advantage of. Despite many risk concerns, the cost and schedule performance index was maintained nearly 1.0 during the assembly phase until project cancellation. In this paper, the assembly logic, the engineering challenges, solutions to those challenges and some of the unique and clever assembly techniques, will be presented. Keywords-field period; assembly; coil joint; metrology; welding

The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving beta(T) and beta(N) values of 17% and 5.7 (%m T MA(-1)), respectively. A newly available motional Stark effect diagnostic has enabled validation of current-drive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the project's main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The stellarator core is designed to produce a compact 3-D plasma that combines stellarator and tokamak physics advantages. The complex geometry and tight fabrication tolerances of NCSX create some unique engineering and assembly challenges. This paper will describe a few of the challenges of the machine's Modular Coils and vacuum vessel field period assembly and how they are being solved. Coil assembly began in November 2005 and to date 3 Modular Coils have been completed. One vacuum vessel 120deg section has been delivered and field period assembly work began in May 2006. Machine sector sub-assembly, machine assembly, and testing will follow, leading to First Plasma in 2011.

Fueling NSTX plasmas with gas injected from the high field side (HFS) has produced earlier, more reliable transitions to the H-mode, longer H-mode durations, higher toroidal rotation, and higher edge T-e compared with similar discharges using the low field side (LFS) gas fueling injectors. The HFS gas fueling system consists of a Center Stack midplane injector, and an injector at the inner, upper corner of the Center Stack. The challenging design and installation constraints for the HFS gas system involved placing the control components as close as possible to the machine-vacuum interface, devising a special feed-through flange, traversing through vessel regions whose temperatures during bake-out range from 150C to 350C, adapting the gas transport tubing size and route to the small instrumentation wire channels behind the existing graphite PFC tiles on the Center Stack, and providing output orifices shielded from excessive plasma power depositions while concentrating the output flow to facilitate fast camera viewing and analysis. Design, recent performance, and future upgrades will be presented.