JRT-2017: FES Multi-Facility Joint Research Target / Milestone (Led by DIII-D)
Conduct research to examine the effect of configuration on operating space for dissipative divertors. Handling plasma power and particle exhaust in the divertor region is a critical issue for future burning plasma devices, including ITER. The very narrow edge power exhaust channel projected for tokamak devices that operate at high poloidal magnetic field is of particular concern. Increased and controlled divertor radiation, coupled with optimization of the divertor configuration, are envisioned as the leading approaches to reducing peak heat flux on the divertor targets and increasing the operating window for dissipative divertors. Data obtained from DIII-D and NSTX-U and archived from Alcator C-Mod will be used to assess the impact of edge magnetic configurations and divertor geometries on dissipative regimes, as well as their effect on the width of the power exhaust channel, thus providing essential data to test and validate leading boundary plasma models.
R(17-1): Simulation-based projection of divertor heat flux footprint for NSTX-U
Description: Understanding the underlying physical mechanisms that contribute to the scrape-off layer (SOL) power flux width, lq, is of utmost importance for future tokamaks such as ITER and FNSF. The 2016 FES Joint Theory Milestone (JTM) carried out extensive modeling to predict lq under attached divertor conditions. XGC modeling reproduced the observed trends in present day tokamaks where lq is found to be nearly inversely proportional to the midplane poloidal field, i.e. lq µ Bpol-1.19. The XGC SOL widths were influenced by both the magnetic drift of warm ions across the separatrix and by the cross-field E×B drift heat-flux width from the edge turbulence. However, only a limited NSTX dataset was included in the 2016 JTM. Here we will extend simulations of a plasma current scan for NSTX to evaluate the SOL width and edge turbulence under a wider range of conditions and to evaluate the magnitudes of the turbulent E×B and ion drift effects in XGC. Building on the NSTX results and modeling, XGC simulations will be extended to predict lq for a high-current 2 MA NSTX-U discharge using a set of assumptions to project the expected midplane profiles.
R(17-2): Advanced divertor operating scenario modeling for NSTX-U
Description: Divertor power exhaust is a critical issue for ITER and next-step tokamaks, and advanced magnetic divertor configurations are being developed and tested to provide candidate solutions for high heat flux and excessive material erosion expected in future facilities. NSTX-U will enable access to a number of advanced divertor configurations including snowflake and X-divertors thanks to a flexible set of divertor poloidal field coils. A range of scrape-off-layer (SOL) widths and high parallel heat fluxes expected in NSTX-U with IP = 1-2 MA, PNBI = 6-12 MW will enable critical tests of the underlying physics of advanced divertor configurations. To guide the experiments, modeling of advanced divertor scenarios and transport will be performed. Divertor radiation and heat fluxes as functions of current, input power, density, and seeded impurities will be studied. Predictive free-boundary codes including ISOLVER and CORSICA will be used to study the operational space of advanced divertor configurations under various solenoid and poloidal field coil current states. The recently developed GINGRED code will be utilized for numerical grid generation for divertor configurations with multiple X-points. Transport and radiation in these advanced divertor configurations will be modeled using SOLPS and UEDGE multi-fluid two-dimensional transport codes and will include studies of the effects of poloidal variation of transport coefficients. The impact of 3D fields on advanced divertor configurations will also be studied using M3D-C1 and EMC3-ERENE codes to understand how small non-axisymmetric perturbation fields may change plasma parameters inside the separatrix and in the divertor. This research will provide a significant step in advanced divertor concept development for NSTX-U and future conventional and spherical tokamaks.
R(17-3): Identify, mitigate, and develop correction strategies for intrinsic error field sources in NSTX-U
Description: A key aspect of achieving high performance in magnetic confinement fusion devices is the successful identification, mitigation, and correction of intrinsic error field sources. During the initial NSTX-U research campaign, error field correction (EFC) experiments revealed strong intrinsic error field effects. In L-mode scenarios, these included the near-universal locking of the q=2 surface and a time-dependence in the optimum phase of the applied n=1 EFC. In H-mode scenarios, error field effects may have impeded early-time H-mode access and limited flattop performance. As such, it is imperative to identify and develop correction strategies for the intrinsic error field sources in NSTX-U. Activities to support this milestone will include conducting in situ coil metrology to generate as-built coil shape models that can be used to inform numerical modeling of intrinsic error field effects. The results of these numerical modeling efforts will be compared to experimental results from the initial NSTX-U campaign in order to identify the key error field sources. Once the error field sources are identified, strategies for mitigating these sources will be developed (for example, realigning the toroidal field bundle within the center-stack casing). For error field sources that cannot be mitigated, strategies for correcting the remaining error fields during plasma operations will be developed (for example, determining optimum pre-programmed n=1 EFC phases and amplitudes). Finally, the calibration procedures for the 3D magnetic field sensors will be improved so that real-time dynamic n=1 EFC can be used during the plasma current ramp in the event that the unmitigated error field sources are difficult to correct with pre-programmed EFC. These various activities will ensure that NSTX-U is optimally positioned to access high performance plasma operations at the outset of the next research campaign.
R(17-4): Assess high-frequency Alfvén Eigenmode stability and associated transport
Description: Experiments and modeling on NSTX have indicated the potential of Compressional and Global Alfvén Eigenmodes (CAE/GAE) to induce both fast-ion redistribution/loss and enhanced electron thermal transport. More flexible NBI heating capabilities in NSTX Upgrade (NSTX-U) enable more comprehensive studies of CAE/GAE physics and support a goal of assessing CAE/GAE stability as a function of the injected NBI source mix. Initial results from the FY-2016 run campaign have already shown a clear dependence of GAE behavior on NBI from specific sources. For example, complete GAE suppression has been observed when a small fraction of NB power from the new 2nd NBI line is added to the total power. Simulations with the HYM code will be used to investigate these initial NSTX-U results and to predict expected CAE/GAE behavior on NSTX-U as the experimental heating power, plasma current, and toroidal field are increased up to their nominal maximum values. Further validation of HYM will be pursued through comparison with experiments from NSTX/NSTX-U and planned experiments from the DIII-D National Campaign. In parallel with CAE/GAE studies, an initial assessment of Ion Cyclotron Emission (ICE) observations from NSTX-U will be performed to characterize the distinctive features of ICE versus sub-ion-cyclotron frequency AEs. For example, ICE features observed on NSTX-U appear different than those observed on conventional tokamaks in that NSTX-U ICE originates near half-radius, i.e. not near the plasma edge. These observations can provide insight into theoretical models of ICE, which are crucial for the potential exploitation of ICE as a confined fast ion diagnostic for ITER. The main goals for ICE studies are: (i) to assess possible correlations with fast ion properties such as the radial profile and the energy dependence of the distribution function, and (ii) to identify which improvements to existing codes (e.g. HYM) are required to properly model ICE.
R(17-5): Analysis and modelling of current ramp-up dynamics in NSTX and NSTX-U
Description: Steady-state, high-beta, and high-confinement conditions are required in future ST devices such as an FNSF or Pilot Plant. A major research goal of NSTX-U is to generate the physics basis for the achievement of such conditions by accessing high toroidal field (0.8-1T) and plasma current (1.6-2.0MA). This milestone aims to accelerate the realization of high plasma current exceeding NSTX levels (Ip > 1.3 MA) when plasma operations resume on NSTX-U. Realizing such target plasma conditions requires operating at sufficiently high elongation in order to operate below MHD stability limits. Scenarios that have achieved suitable elongation on NSTX and NSTX-U utilized an L-H transition early in the Ip ramp-up phase in order to obtain low internal inductance (li) throughout the discharge, which is conducive to maintaining vertical stability at high elongation. The details of the ramp-up phase, encompassing the transition from a limited to diverted shape, the L-H transition, and the rapid rise in the plasma current, have a large influence on the scenario current and pressure limits. The limits to the achievable elongation realized on NSTX-U will be evaluated, with new attention to the limits during the ramp-up phase, and compared to NSTX. In particular, the growth rate of vertical instabilities will be studied by evaluating the elongation limits realized on NSTX-U during the ramp-up phase, especially at the time of diverting. The dependence of the L-H transition on plasma parameters such as density, plasma shape, and plasma current on NSTX-U will be evaluated in order to generate a threshold criterion for simulations. Further, kink and tearing stability analysis of the ramp-up phase of the experimental and simulated discharges will be initiated to improve understanding of the global MHD stability limits of the current ramp-up phase.