The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

NSTX-U theory, modeling and analysis results

ITER is incorporating two types of In Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required so that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are in close proximity to the plasma, mounted just behind the Blanket Shield Modules. This location results in a radiation and temperature environment that is severe necessitating new solutions for material selection as well as challenging analysis and design solutions. Fitting the coil systems in between the blanket shield modules and the vacuum vessel leads to difficult integration with diagnostic cabling and cooling water manifolds. The design of the IVCs is now progressing towards a final design scheduled for late CY 13. The project is a collaboration between the Princeton Plasma Physics Laboratory in Princeton NJ, the Chinese Academy of Sciences (ASIPP) in Hefei China and the ITER Organization. An extensive thermal and stress analysis to evaluate the effects of the high temperatures and electromagnetic loads on the In Vessel Coils has been undertaken. Manufacturing development is underway at ASIPP to develop the processes necessary to build ELM coil and VS Coil prototypes. This paper will outline the design and analysis issues as well as review the manufacturing development required to address these requirements and plans for prototypes.

/pdf/design-analysis-and-manufacturing-studies-for-iter-in-vessel-x6ydvg0skv.pdf

Design analysis and manufacturing studies for ITER In-Vessel Coils

Structural analysis and optimization for ITER upper ELM coil

An optimized upper divertor with divertor-coils to study enhanced divertor configurations in ASDEX Upgrade

Mechanical analysis for ITER Upper ELM coil is divided into two sections, in the 1st one a FE model is built and the critical components like the conductor and jacket are analyzed, while the 2nd part analytically checks the bolt strength with the load from the previous step. Three areas representing this effort are presented: electro-magnetic, hydraulic-thermal and mechanical analysis of upper ELM coil. Electro-magnetic analysis provides the ohmic heat and electromagnetic force for hydraulic thermal and mechanical analysis respectively. Hydraulic thermal analysis is performed to get the temperature distribution of the upper ELM coil for further mechanical analysis. Mechanical analysis aids in the design of upper ELM coil while quantifying the stresses and machine loads which can be used to verify the bolts with regard to ASME and machine design handbook.

Mechanical analysis for ITER upper ELM coil

ITER will incorporate In Vessel Coils (IVCs) as a method of stabilizing “Edge Localized Modes” (ELM) and providing “Vertical Stabilization” (VS). To meet the ELM and VS Coil requirements strong coupling with the plasma is required so that it is necessary for the coils to be installed in the vessel just behind the blanket shield modules. Due to this close proximity to the plasma the radiation and temperature environment is severe and conventional electrical insulation materials and processes cannot be used. The development of mineral insulated conductor technology has been required in the IVC design to deal with this high radiation and high temperature environment. While mineral insulated conductor technology is not new, building a large magnet with high current carrying capability and a conductor diameter larger than the mineral insulated conductor currently manufactured requires R&D and the extension of existing technologies. A 59mm Stainless Steel Jacketed Mineral Insulated Conductor (SSMIC) using MgO is being developed for this application. The IVC ELM and VS coils design includes both the development of the fabrication techniques for the SSMIC and the design and analysis of the ELM and VS Coil assemblies.

Iter In-Vessel Coil design and R&D

Abstract The engineering limits of plasma-facing components (PFCs) constrain the allowable operational space of tokamaks. Poorly managed heat fluxes that push the PFCs beyond their limits not only degrade core plasma performance via elevated impurities, but can also result in PFC failure due to thermal stresses or melting. Simple axisymmetric assumptions fail to capture the complex interaction between three-dimensional (3-D) PFC geometry and two-dimensional or 3-D plasmas. This results in fusion systems that must either operate with increased risk or reduce PFC loads, potentially through lower core plasma performance, to maintain a nominal safety factor. High-precision 3-D heat flux predictions are necessary to accurately ascertain the state of a PFC given the evolution of the magnetic equilibrium. A new code, the Heat flux Engineering Analysis Toolkit (HEAT), has been developed to provide high-precision 3-D predictions and analysis for PFCs. HEAT couples many otherwise disparate computational tools together into a single open-source python package. Magnetic equilibrium, engineering computer-aided design, finite volume solvers, scrape-off layer plasma physics, visualization, high-performance computing, and more, are connected in a single web-based user interface. Linux users may use HEAT without any software prerequisites via an appImage. This paper introduces HEAT, discusses the software architecture, presents the first HEAT results, and outlines physics modules in development.

https://www.tandfonline.com/doi/pdf/10.1080/15361055.2021.1951532?needAccess=true

A Software Package for Plasma-Facing Component Analysis and Design: The Heat Flux Engineering Analysis Toolkit (HEAT)

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Iter In-Vessel Coil design and R&D

A Software Package for Plasma-Facing Component Analysis and Design: The Heat Flux Engineering Analysis Toolkit (HEAT)