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FY2016 NSTX-U Research Milestones

FY2016 FES Notable Outcomes for NSTX-U
  • Perform experimental research on NSTX-Upgrade 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 NSTX-U project.
  • Conduct NSTX-U experiments and data analysis to support the FES joint research target on detecting and minimizing the consequences of disruptions in present and future tokamaks, including ITER.
JRT-2016: FES 3 Facility Joint Research Target / Milestone

Conduct research to detect and minimize the consequences of disruptions in present and future tokamaks, including ITER. Coordinated research will deploy a disruption prediction/warning algorithm on existing tokamaks, assess approaches to avoid disruptions, and quantify plasma and radiation asymmetries resulting from disruption mitigation measures, including both pre‐existing and resulting MHD activity, as well as the localized nature of the disruption mitigation system. The research will employ new disruption mitigation systems, control algorithms, and hardware to help avoid disruptions, along with measurements to detect disruption precursors and quantify the effects of disruptions.

NSTX-U Contribution to JRT-2016: Assess gas assimilation efficiencies for MGI, and utilize MGI and MHD control coupled to a disruption warning system

Disruptions in a reactor-scale tokamak/ST can cause unacceptable damage to plasma facing components and thus must be avoided or detected in advance and properly mitigated. Disruption studies on NSTX-U will offer new insight for these critical subjects utilizing unique MGI equipment and pioneering disruption avoidance systems and warning algorithms developed during NSTX operation. Three identical MGI systems at different poloidal locations will enable the assessment of the gas assimilation efficiency as a function of poloidal injection location and edge plasma parameters. Consequent reductions to the divertor heat loads and halo currents by mitigation as a function of the gas assimilation fraction will also be measured. The gas penetration past the SOL will be studied during the thermal quench, which will be quantitatively compared with simulation such as from the DEGAS-2 code modeling. In parallel, NSTX-U will develop and test a disruption warning system, and the quality of individual warning sensors will be improved during the course of experimental applications. With the help of physics-based MHD modeling, magnetic sensor response will be compared with state-space observers in real-time to address the correlation with RWM stability. Total and/or various combinations of these warning sensors will be investigated and improved in conjunction with mitigation and avoidance actuators. In particular, MHD sensors, such as low frequency RWM sensors, state-space observers, and resonant field amplification will be tested along with the algorithm for real-time MGI triggering and also algorithms for real-time rotation and β control to avoid global instability and disruption. These studies from NSTX-U will be used to inform the disruption mitigation system (DMS) design for ITER and will also inform the development of advanced disruption warning and avoidance systems in next-step tokamaks/STs as well as ITER.

R(16-1): Assess H-mode energy confinement, pedestal, and scrape-off-layer characteristics with higher BT, Ip and NBI heating power

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(16-2): 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(16-3): 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.