Modification of the NSTX-U outboard and Inboard Divertor tiles for the protection of the PF-1C coils

The National Spherical Torus Experiment (NSTX) is a low aspect ratio, spherical torus (ST) located at the Princeton Plasma Physics Laboratory (PPPL). Its Centerstack Assembly (CSA) consists of the inner legs of the Toroidal Field windings, the Ohmic Heating solenoid, the inner Poloidal Field (PF) coils, thermal insulation, diagnostics, and an Inconel casing which forms the inner wall of the vacuum vessel boundary. The outside surface of this casing is protected from the heat loads by a layer of Plasma Facing Components (PFCs), in this case, a combination of ATJ and POCO TM graphite. The CSA is electrically isolated from the outer, large major radius part of the vacuum chamber by ceramic insulators. The gaps at the top and bottom of the machine between the CSA and the outer vessel are known as the “Coaxial Helicity Injection (CHI) Gaps”. The PF-1C divertor coils are located in this region shadowed by the CHI gap, however, late in the design of the NSTX Upgrade PFCs, MHD equilibria were discovered which could direct field lines through these CHI gaps and onto the PF-1C stainless steel casings. This could result in thermal flux from the main plasma body to flow along the field lines directly onto the coil. Though the probability of such an event is low, the heat flux on the 0.125” thick stainless steel casing could damage the coil beneath, or in worst case, rupture the casing itself, resulting in an accidental vent of the vacuum vessel. By extending downward the overhanging edge of the row 1 PFCs on the Outboard Divertor (OBD) and the Inboard Divertor (IBD) and by increasing their outermost radii, effectively narrowing the CHI gap, this should provide significant protection to the PF-1C casing.