2015 FES Notable Outcome
Perform experimental research on NSTX-U to resolve key spherical torus issues at magnetic field, plasma current, and pulse length beyond that achieved in NSTX, after completion of CD-4 for the project.
JRT-2015: FES 3 Facility Joint Research Target / Milestone
R(15-1): Assess H-mode energy confinement, pedestal, and scrape-off-layer characteristics with higher BT, Ip and NBI heating power
Conduct experiments and analysis to quantify the impact of broadened current and pressure profiles on tokamak plasma confinement and stability. Broadened pressure profiles generally improve global stability but can also affect transport and confinement, while broadened current profiles can have both beneficial and adverse impacts on confinement and stability. This research will examine a variety of heating and current drive techniques in order to validate theoretical models of both the actuator performance and the transport and global stability response to varied heating and current drive deposition.
Future ST devices such as ST-FNSF will operate at higher toroidal field, plasma current and heating power than NSTX. To establish the physics basis for future STs, which are generally expected to operate in lower collisionality regimes, it is important to characterize confinement, pedestal and scrape off layer trends over an expanded range of engineering parameters. H-mode studies in NSTX have shown that the global energy confinement exhibits a more favorable scaling with collisionality (BT tau_E ~ 1/nu*e) than that from ITER98y,2. This strong n*e scaling unifies disparate engineering scalings with boronization (tau_E ~ Ip^0.4 BT^1.0) and lithiumization (tau_E ~ Ip^0.8 BT^-0.15). In addition, the H-mode pedestal pressure increases with ~Ip^2, while the divertor heat flux footprint width decreases faster than linearly with Ip. With double BT, double Ip and double NBI power with beams at different tangency radii, NSTX-U provides an excellent opportunity to assess the core and boundary characteristics in regimes more relevant to future STs and to explore the accessibility to lower collisionality. Specifically, the relation between H-mode energy confinement and pedestal structure with increasing Ip, BT and PNBI will be determined and compared with previous NSTX results, including emphasis on the collisionality dependence of confinement and beta dependence of pedestal width. Coupled with low-k turbulence diagnostics and gyrokinetic simulations, the experiments will provide further evidence for the mechanisms underlying the observed confinement scaling and pedestal structure. The scaling of the divertor heat flux profile with higher IP and PNBI will also be measured to characterize the peak heat fluxes and scrape off layer widths, and this will provide the basis for eventual testing of heat flux mitigation techniques. Scrape-off layer density and temperature profile data will also be obtained for several divertor configurations, flux expansion values, and strike-point locations to validate the assumptions used in the FY2012-13 physics design of the cryopump to inform the cryo-pump engineering design to be carried out during FY2015.
R(15-2): Assess the effects of neutral beam injection parameters on the fast ion distribution function and neutral beam driven current profile
Accurate knowledge of neutral beam (NB) ion properties is of paramount importance for many areas of tokamak physics. NB ions modify the power balance, provide torque to drive plasma rotation and affect the behavior of MHD instabilities. Moreover, they determine the non-inductive NB driven current, which is crucial for future devices such as ITER, FNSF and STs with no central solenoid. On NSTX-U, three more tangentially-aimed NB sources have been added to the existing, more perpendicular ones. With this addition, NSTX-U is uniquely equipped to characterize a broad parameter space of fast ion distribution, Fnb, and NB-driven current properties, with significant overlap with conventional aspect ratio tokamaks. The two main goals of the proposed Research Milestone on NSTX-U are (i) to characterize the NB ion behavior and compare it with classical predictions, and (ii) to document the operating space of NB-driven current profile. Fnb will be characterized through the upgraded set of NSTX-U fast ion diagnostics (e.g. fast-ion D-alpha: FIDA, solid-state neutral particle analyzer: ssNPA, scintillator-based fast-lost-ion probe: sFLIP, and neutron counters) as a function of NB injection parameters (tangency radius, beam voltage) and magnetic field. Well controlled, single-source scenarios at low NB power will be initially used to compare fast ion behavior with classical models (e.g. the NUBEAM module of TRANSP) in the absence of fast ion driven instabilities. Diagnostics data will be interpreted through the “beam blip” analysis technique and other dedicated codes such as FIDASIM. Then, the NB-driven current profile will be documented for the attainable NB parameter space by comparing NUBEAM/TRANSP predictions to measurements from Motional Stark Effect, complemented by the vertical/tangential FIDA systems and ssNPA to assess modifications of the classically expected Fnb. As operational experience builds up during the first year of NSTX-U experiments, additions to the initial Fnb assessment will be considered for scenarios where deviations of Fnb from classical predictions can be expected. The latter may include scenarios with MHD instabilities, externally imposed non-axisymmetric 3D fields, and additional High-Harmonic Fast Wave (HHFW) heating.
R(15-3): Develop the physics and operational tools for obtaining high-performance discharges in NSTX-U
Steady-state, high-beta conditions are required in future ST devices, such as a FNSF/CTF facility, for increasing the neutron wall loading while minimizing the recirculating power. NSTX-U is designed to provide the physics knowledge for the achievement of such conditions by demonstrating stationary, long pulse, high non-inductive fraction operation. The ultimate toroidal field (1.0 T) and plasma current (2.0MA) capability of NSTX-U is twice that in NSTX. NSTX-U has a capability for >5 second discharges, and it has an additional beamline which doubles the available heating power and provides much greater flexibility in the beam current drive profile. The aim for studies during the first year of operation of NSTX-U is to lay the foundation for the above operational scenario goals by developing needed physics and operational tools, using toroidal fields up to ~0.8 T, plasma currents up to ~1.6 MA, improved applied 3D field capabilities from additional power supplies, a variety of plasma facing component (PFC) conditioning methods, and advanced fuelling techniques. As an example of the latter, supersonic gas injection provides higher fuelling efficiency, and will be used to develop reliable discharge formation with minimal gas loading. Differing PFC conditioning techniques, including boronization and lithium coatings, will be assessed to determine which are most favorable for longer pulse scenarios. Impurity control techniques, an example of which is ELM pacing, will be developed for the reduction of impurity accumulation in otherwise ELM-free lithium-conditioned H-modes. The higher aspect ratio, high elongation (2.8 < kappa < 3.0) plasma shapes anticipated to result in high non-inductive fraction in NSTX-U will be developed, and the vertical stability of these targets will be assessed, with mitigating actions taken if problems arise. An initial assessment of low-n error fields will be made, along with expanding the RWM control and dynamic error field correction strategies using both proportional and state-space n ≥ 1 feedback schemes, taking advantage of the spectrum flexibility provided by the 2nd SPA power supply. Resonant field amplification measurements, ideal MHD stability codes, and kinetic stability analysis will be used to evaluate the no-wall and disruptive stability limits in these higher aspect ratio and elongation scenarios. These physics and operational tools will be combined to make an initial assessment of the non-inductive current drive fraction across a range of toroidal field, plasma density, boundary shaping, and neutral beam parameters.
IR(15-1): Develop and assess the snowflake divertor configuration and edge properties in NSTX-U
The high flux expansion snowflake divertor configuration is a leading candidate for mitigation of high power exhaust in NSTX-Upgrade, where projected peak heat fluxes up to 20 MW/m^2 are anticipated in 12 MW NBI-heated discharges. In NSTX-U, an upgraded up-down symmetric set of three divertor coils will be used to develop a variety of snowflake configurations. Experiments in FY2015 will be focused on 1) development of magnetic configuration control; 2) initial studies of edge and divertor properties. In the area of magnetic control, fast numerical algorithms will be implemented and tested with the Plasma Control System to develop feedback control of inter-X-point distance, X-point orientation, and flux expansion. Divertor heat flux handling and power accountability and impurity production trends with engineering parameters that will become accessible in NSTX-U, such as Ip = 1-2 MA, PNBI = 6-12 MW, will be assessed. H-mode pedestal stability will also be assessed to determine if the snowflake configurations can be used for ELM control. Measurements will be compared to multi-fluid and kinetic model predictions. Results from initial NSTX-U snowflake divertor experiments will also be compared with the experiments that were performed in NSTX and DIII-D. This research will provide a significant step in the snowflake divertor concept development for both the ST and tokamak.