Physics advancements have been made in new stellarator experimental devices, particularly in the operation of W7-X. However the Achilles heel of the stellarator is the complexity of the winding topology and the inability to define a physics solution that also provides a credible fusion blanket maintenance scheme. A 2013-14 internal PPPL pilot plant study was undertaken to enhance the maintenance characteristics of the stellarator by straightening the modular coil (MC) outboard legs to provide greater access to plasma components. An ARIES-CS, three-period, stellarator design incorporating a higher aspect ratio (AR) plasmas fostering less complicated MC shapes was configured by a physics code (COILOPT++) upgraded to include engineering metrics and geometry constraints to enhanced maintenance access conditions. The outcome of this effort (shown in Figure 1) was a 1000 MW, 9.4-m major radius; AR 6 design developed with most modular coil outboard legs located in a near vertical plane allowing a tokamak style, vertical maintenance to remove large blanket sectors. All opportunities to bring this to a successful conclusion were considered. One solution found was to reduce the MC winding cross-section by the inclusion of TF coils in the MC optimization process. The results found that the MC winding currents could be cut in half when TF coils centers were located a sufficient distance from the MC’s. As much as blanket maintenance conditions improved in the areas of the Type-A and B MC’s there still existed complications in the movement of blankets in the area of the Type-C modular coil where shaping of the windings could not be contained within the specified region to allow radial removal of a blanket sector.
In the six years since the 2013-14 stellarator study was performed further physics, engineering and technology improvements have been made that can improve access constraints that occurred at the Type-C modular coils. The earlier physics codes used to define the plasma and MC windings (STELLOPT and COILOPT++) have been updated and combined to form a new code called STELLCOPT which has been shown to produce configurations with superior physics with enhanced shaping of the modular coil outboard region than produced in the earlier study (comparative figure shown below). Control points of a B-spline representation of MC windings on a coil surface are varied, and for each coil realization a free-boundary plasma equilibrium is constructed. Desirable physics properties of the free-boundary plasma are targeted subject to imposed engineering constraints. At each iteration of STELLCOPT the plasma shape is allowed to respond to the change in the coil geometry. Unlike the earlier two-stage reverse-engineering approach, there is no single target plasma shape - only a set of targeted physics properties. This freedom leads to a larger space of attractive plasma-coil configurations. On the left side of Figure 2 the adjacent figure the black boundary is the target fixed-boundary plasma - three field period, AR = 6.0, beta = 4.0%, QA optimized, stable to n=0 and n=1 kink families. The red boundary is the free boundary reconstruction. Kink stability has been lost due to COILOPT++ fitting errors to the B-normal target provided by STELLOPT and effective helical ripple has also degraded. The lower figure shows output from a STELLCOPT run where identical physics and engineering targets were assumed, yet the final free-boundary plasma is kink stable. The effective helical ripple of the plasma is also improved. The lower u-v curve indicates near straight outboard leg regions for the STELLCOPT MC solution while the upper u-v curve shows the COILOPT++ solution from the reverse-engineered approach.
Engineering winding designs and conductor technology also has significantly advanced over the time period of the earlier study that allows the winding current density to increase considerably over the 24 A/mm2 MC winding current density used in establishing the 2013 straight leg stellarator design. Low temperature superconductor winding designs and conductor performance has increased winding current density values to 80 A/mm2 for 10-15T peak coil field. With future technology and economic improvements in HTS conductor technology modular coil winding current density values may reach 300 A/mm2 or higher which will significantly reduce the size of the MC winding, alleviating tight bend radius and further increase access areas.
This paper provides the enhancements made to improve the 2014 straight leg stellarator design – defining an overview of physics results and improvements made in the device general arrangement to provide unrestricted access to remove all blanket sectors. The inclusion of high current density magnet systems (LT and HTS), investigation of permanent magnets and current loops added to the blanket module area of the Type-C magnets to relax blanket restrictions will be highlighted along with preliminary results investigating the impact on blanket access with lower aspect ratio.
- T. Brown, et al, “Engineering optimization of stellarator coils lead
to improvements in device maintenance”, published in 2015 IEEE 26th
Symposium on Fusion Engineering (SOFE), June 2015
- D. Gates, et al,
“Recent advancements in stellarator optimization”, Nucl. Fusion 57
- T. Brown, et al, “Three confinement Systems –
Spherical Tokamak, Standard Tokamak and Stellarator: A Comparison of
key component cost elements”. IEEE Transactions of Plasma Science
(Vol. 48, June 2018)
|Affiliation||Princeton Plasma Physics Laboratory|
|Country or International Organization||United States|